Neutronic calculations of AFPR-100 reactor based on Spherical Cermet Fuel particles

International Nuclear Information System (INIS)

Benchrif, A.; Chetaine, A.; Amsil, H.

2013-01-01

Highlights: • AFPR-100 reactor considered as a small nuclear reactor without on-site refueling originally based on TRISO micro-fuel element. • The AFPR-100 reactor was re-designed using the new Spherical Cermet fuel element. • The adoption of the Cermet fuel instead of TRISO fuel reduces the core lifetime operation by 3.1 equivalent full power years. • We discussed the new micro-fuel element candidate for small and medium sized reactors. - Abstract: The Atoms For Peace Reactor (AFPR-100), as a 100 MW(e) without the need of on-site refueling, was originally based on UO2 TRISO fuel coated particles embedded in a carbon matrix directly cooled by light water. AFPR-100 is considered as a small nuclear reactor without open-vessel refueling which is proposed by Pacific Northwest National Laboratory (PNNL). An account of significant irradiation swelling in the silicon carbide fission product barrier coating layer of TRISO fuel element, a Spherical Cermet Fuel element has been proposed. Indeed, the new fuel concept, which was developed by PNNL, consists of changing the pyro-carbon and ceramic coatings that are incompatible with low temperature by Zirconium. The latter was chosen to avoid any potential Wigner energy effect issues in the TRISO fuel element. Actually, the purpose of this study is to assess the goal of AFPR-100 concept using the Cermet fuel; undeniably, the fuel core lifetime prediction may be extended for reasonably long period without on-site refueling. In fact, we investigated some neutronic parameters of reactor core by the calculation code SRAC95. The results suggest that the core fuel lifetime beyond 12 equivalent full power years (EFPYs) is possible. Hence, the adoption of Cermet fuel concept shows a core lifetime decrease of about 3.1 EFPY

Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

International Nuclear Information System (INIS)

Atkins, D.F.; Lillie, A.F.

1983-01-01

As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

Progress in space nuclear reactor power systems technology development - The SP-100 program

Science.gov (United States)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Impingement studies at the 100-N reactor water intake

International Nuclear Information System (INIS)

Page, T.L.; Neitzel, D.A.; Gray, R.H.

1977-09-01

Fish impingement and traveling screen passage were studied at the 100-N reactor water intake structure, Columbia River mile 380, from late April to August 1977. Species and numbers of fish affected were determined and compared to those at the adjacent Hanford Generating Project (HGP). Fish protection procedures previously developed for HGP were evaluated for application at 100-N

PG-100 helium loop in the MR reactor

International Nuclear Information System (INIS)

Ponomarev-Stepnoj, N.N.; Yakovlev, V.V.; Tikhonov, N.I.

1983-01-01

Main systems and production equipment units of PG-100 helium loop in the MR reactor are described. Possible long-term synchronizing operation of loop and reactor as well as possibility of carrying out life-time tests of spherical fuel elements and materials are shown. Serviceability of spherical fuel elements under conditions similar to the ones of HTGR-50 operation as well as high serviceability of cleanup system accepted for HTGR are verified. Due to low radiation dose the loop is operated without limits, helium losses in the loop don't exceed 0.5%/24 h, taking account of experimental gas sampling

Software for the nuclear reactor dynamics study using time series processing

International Nuclear Information System (INIS)

Valero, Esbel T.; Montesino, Maria E.

1997-01-01

The parametric monitoring in Nuclear Power Plant (NPP) permits the operational surveillance of nuclear reactor. The methods employed in order to process this information such as FFT, autoregressive models and other, have some limitations when those regimens in which appear strongly non-linear behaviors are analyzed. In last years the chaos theory has offered new ways in order to explain complex dynamic behaviors. This paper describes a software (ECASET) that allow, by time series processing from NPP's acquisition system, to characterize the nuclear reactor dynamic as a complex dynamical system. Here we show using ECASET's results the possibility of classifying the different regimens appearing in nuclear reactors. The results of several temporal series processing from real systems are introduced. This type of analysis complements the results obtained with traditional methods and can constitute a new tool for monitoring nuclear reactors. (author). 13 refs., 3 figs

Software for the nuclear reactor dynamics study using time series processing; Software para el estudio de la dinamica de reactores nucleares mediante el procesamiento de series temporales

Energy Technology Data Exchange (ETDEWEB)

Valero, Esbel T.; Montesino, Maria E. [Instituto Superior de Ciencia y Tecnologia Nuclear (ISCTN), La Habana (Cuba)

1997-12-01

The parametric monitoring in Nuclear Power Plant (NPP) permits the operational surveillance of nuclear reactor. The methods employed in order to process this information such as FFT, autoregressive models and other, have some limitations when those regimens in which appear strongly non-linear behaviors are analyzed. In last years the chaos theory has offered new ways in order to explain complex dynamic behaviors. This paper describes a software (ECASET) that allow, by time series processing from NPP`s acquisition system, to characterize the nuclear reactor dynamic as a complex dynamical system. Here we show using ECASET`s results the possibility of classifying the different regimens appearing in nuclear reactors. The results of several temporal series processing from real systems are introduced. This type of analysis complements the results obtained with traditional methods and can constitute a new tool for monitoring nuclear reactors. (author). 13 refs., 3 figs.

Series lecture on advanced fusion reactors

International Nuclear Information System (INIS)

Dawson, J.M.

1983-01-01

The problems concerning fusion reactors are presented and discussed in this series lecture. At first, the D-T tokamak is explained. The breeding of tritium and the radioactive property of tritium are discussed. The hybrid reactor is explained as an example of the direct use of neutrons. Some advanced fuel reactions are proposed. It is necessary to make physics consideration for burning advanced fuel in reactors. The rate of energy production and the energy loss are important things. The bremsstrahlung radiation and impurity radiation are explained. The simple estimation of the synchrotron radiation was performed. The numerical results were compared with a more detailed calculation of Taimor, and the agreement was quite good. The calculation of ion and electron temperature was made. The idea to use the energy more efficiently is that one can take X-ray or neutrons, and pass them through a first wall of a reactor into a second region where they heat the material. A method to convert high temperature into useful energy is the third problem of this lecture. The device was invented by A. Hertzberg. The lifetime of the reactor depends on the efficiency of energy recovery. The idea of using spin polarized nuclei has come up. The spin polarization gives a chance to achieve a large multiplication factor. The advanced fuel which looks easiest to make go is D plus He-3. The idea of multipole is presented to reduce the magnetic field inside plasma, and discussed. Two other topics are explained. (Kato, T.)

The high moderating ratio reactor using 100% MOX reloads

International Nuclear Information System (INIS)

Barbrault, P.

1994-06-01

This report presents the concept of a High Moderating ratio Reactor, which should accept 100% MOX reloads. This reactor aims to be the plutonium version of the European Pressurized Reactor (EPR), which is developed jointly by French and German companies. A moderating ration of 2.5 (instead of the standard value of 2.0) is obtained by replacing several fuel rods by water holes. The core would contain 241 Fuel Assemblies. We present some advantages of over-moderation for plutonium fuel, a description of the core and assemblies, calculations of fuel reload schemes and Reactivity Shutdown Margins, and the behavior of the core during two occidental transients. (author). 2 refs., 9 figs., 2 tabs

SP-100 space reactor power system readiness

International Nuclear Information System (INIS)

Josloff, A.T.; Matteo, D.N.; Bailey, H.S.

1992-01-01

This paper discusses the SP-100 Space Reactor Power System which is being developed by GE, under contract to the U.S. Department of Energy, to provide electrical power in the range of 10's to 100's of kW. The system represents an enabling technology for a wide variety of earth orbital and interplanetary science missions, nuclear electric propulsion (NEP) stages, and lunar/Mars surface power for the Space Exploration Initiative (SEI). The technology and design is now at a state of readiness to support the definition of early flight demonstration missions. Of particular importance is that SP-100 meets the demanding U.S. safety performance, reliability and life requirements. The system is scalable and flexible and can be configured to provide 10's to 100's of kWe without repeating development work and can meet DoD goals for an early, low-power demonstration flight in the 1996-1997 time frame

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

Design of an isopropanol–acetone–hydrogen chemical heat pump with exothermic reactors in series

International Nuclear Information System (INIS)

Xu, Min; Duan, Yanjun; Xin, Fang; Huai, Xiulan; Li, Xunfeng

2014-01-01

The isopropanol–acetone–hydrogen chemical heat pump system with a series of exothermic reactors in which the reaction temperatures decrease successively is proposed. This system shows the better energy performances as compared with the traditional system with a single exothermic reactor, especially when the higher upgraded temperature is need. At the same amounts of the heat released, the work input of the compressor and the heater are both reduced notably. The results indicate that the advantages of the IAH-CHP system with exothermic reactors in series are obvious. - Highlights: • We propose the IAH-CHP system with exothermic reactors in series. • The COP and exergy efficiency of the system increase by 7.6% and 10.3% respectively. • The work input of the system is reduced notably at the same quantity of heat released

On-line analysis of reactor noise using time-series analysis

International Nuclear Information System (INIS)

McGevna, V.G.

1981-10-01

A method to allow use of time series analysis for on-line noise analysis has been developed. On-line analysis of noise in nuclear power reactors has been limited primarily to spectral analysis and related frequency domain techniques. Time series analysis has many distinct advantages over spectral analysis in the automated processing of reactor noise. However, fitting an autoregressive-moving average (ARMA) model to time series data involves non-linear least squares estimation. Unless a high speed, general purpose computer is available, the calculations become too time consuming for on-line applications. To eliminate this problem, a special purpose algorithm was developed for fitting ARMA models. While it is based on a combination of steepest descent and Taylor series linearization, properties of the ARMA model are used so that the auto- and cross-correlation functions can be used to eliminate the need for estimating derivatives. The number of calculations, per iteration varies lineegardless of the mee 0.2% yield strength displayed anisotropy, with axial and circumferential values being greater than radial. For CF8-CPF8 and CF8M-CPF8M castings to meet current ASME Code S acid fuel cells

A 50-100 kWe gas-cooled reactor for use on Mars.

Energy Technology Data Exchange (ETDEWEB)

Peters, Curtis D. (.)

2006-04-01

In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

SVBR-75/100 multi-purpose modular inherent-safety fast reactor

International Nuclear Information System (INIS)

Dragunov, Yu.G.; Stepanov, V.S.; Klimov, N.N.; Dedul, A.V.; Zrodnokov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Krushelnitsky, V.N.; Takh, S.M.

2006-01-01

In this century energy consumption, including electric power, will continue growing on a large scale especially in developing countries. Significant changes in electric power market needs are to be expected in the direction of decreasing and varying the capacity of power sources. To satisfy the expected growth of demand for electric power and to take a decision concerning the ways of further development of global power, including nuclear engineering, it is very important to continue the development of innovative concepts of nuclear power sources, which might successfully compete with alternative power technologies at the future power markets. The proposed nuclear power source (or in other words - reactor plant) of new generation is supposed: - to have small power capacity in the range of 10 - 100 MW (electric) and possibility of its multi-purpose application (independent nuclear power source for desalination installations and electricity supply, nuclear power plants (NPP) of various capacity and purpose; - to use modular principle of construction of NPP of various capacity on the basis of unified 'typical' reactor plants; - to have qualitatively new level of passive safety and possess properties of inherent safety, deterministically excluding any opportunity of severe accidents; - to have an opportunity to use different kinds of fuel and to work in various fuel cycles at various stages of development of nuclear power without change in the design. And also to have long (7-10 years, and in the long term 15-20 years) core life time and enrichment on U-235 not higher than 20 % (which is in compliance with recommendations of IAEA under non-proliferation condition); - to be completely factory-manufactured, and an opportunity of its safe transportation to and from the NPP site shall be provided. Unified multi-purpose reactor plant SVBR-75/100 (Lead-Bismuth Fast Reactor with equivalent electric power of 75 - 100 MW-e depending on the steam parameters) meets the set of the

Considerations concerning the reliability of reactor safety equipment; Considerations sur la fiabilite des ensembles de securite de reacteurs

Energy Technology Data Exchange (ETDEWEB)

Furet, J; Guyot, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

A review is made of the circumstances which favor a good collection of maintenance data at the C.E.A. The large amount of data to be treated has made necessary the use of a computer for analyzing automatically the results collected. Here, only particular aspects of the reliability from the point of view of the electronics used for nuclear reactor control will be dealt with: sale and unsafe failures; probability of survival (in the case of reactor safety); availability. The general diagrams of the safety assemblies which have been drawn up for two types of reactor (power reactor and low power experimental reactor) are given. Results are presented of reliability analysis which could be applied to the use of functional modular elements, developed industrially in France. Improvement of this reliability appears to be fairly limited by an increase in the redundancy; on the other hand it is shown how it may be very markedly improved by the use of automatic tests with different frequencies for detecting unsafe failures rates of measurements for the sub-assemblies and for the logic sub-assemblies. Finally examples are given to show the incidence of the complexity and of the use of different technologies in reactor safety equipment on the reliability. (authors) [French] On rappelle les circonstances qui favorisent au C.E.A. la collecte d'une information valable des resultats de la maintenance. L'importance des donnees a traiter a rendu necessaire l'utilisation d'une calculatrice poux l'analyse automatique des resultats recueillis. On se limitera ici aux aspects particuliers de la fiabilite du point de vue de l'electronique pour le controle et la commande de reacteurs nucleaires: pannes sures et pannes non sures; probabilite de survie dans le cas de la securite des reacteurs; facteur de disponibilite. Les schemas de principe des ensembles de securite definis pour deux types de reacteurs (reacteur de puissance et reacteur experimental de faible puissance) sont indiques. On

Thyristor-controlled reactor improves series capacitor applications

Energy Technology Data Exchange (ETDEWEB)

Renz, K.W.; Thumm, G.; Weiss, S. [Siemens AG, Erlangen (Germany)

1995-12-31

Environmental considerations make it more and more difficult to plan and erect new transmission lines. FACTS (Flexible AC Transmission Systems) technology can provide devices to improve the utility of AC transmission lines. The innovative combination of conventional fixed series capacitors and thyristor controlled reactors as a new FACTS device was introduced into a transmission system in 1992. This Advanced Series Compensation (ASC) system provides many advantages not available with conventional fixed series capacitor installations such as flexible direct and continuous control of the compensation level, direct and smooth power flow control and improved capacitor bank protection. This new technology offers enhanced system flexibility by control of transmission line overload conditions, reduction in fault currents, sub-synchronous resonance (SSR) mitigation and network power oscillation damping. The world-first three-phase installation at Kayenta Substation, USA, demonstrates that modern FACTS devices using SVC thyristor valve technology can be designed and operated successfully. 6 refs, 7 figs

Neutronic design for a 100MWth Small modular natural circulation lead or lead-alloy cooled fast reactors core

International Nuclear Information System (INIS)

Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q.

2015-01-01

Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW th natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

International Nuclear Information System (INIS)

Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

2004-01-01

As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

Analysis and evaluation of ZPPR critical experiments for a 100 kilowatt-electric space reactor

International Nuclear Information System (INIS)

McFarlane, H.F.; Collins, P.J.; Carpenter, S.G.; Olsen, D.N.; Smith, D.M.; Schaefer, R.W.; Doncals, R.A.; Andre, S.V.; Porter, C.A.; Cowan, C.L.; Stewart, S.L.; Protsik, R.

1990-01-01

ZPPR critical experiments were used for physics testing the reactor design of the SP-100, a 100-kW thermoelectric LMR that is being developed to provide electrical power for space applications. These tests validated all key physics characteristics of the design, including the ultimate safety in the event of a launch or re-entry accident. Both the experiments and the analysis required the use of techniques not previously needed for fast reactor designs. A few significant discrepancies between the experimental and calculated results leave opportunities for further reductions in the mass of the SP-100. An initial investigation has been made into application of the ZPPR-20 results, along with those of other relevant integral data, to the SP-100 design

Systems analysis of a 100-MWe modular liquid metal cooled reactor

International Nuclear Information System (INIS)

Morris, E.E.; Rhow, S.K.; Switick, D.M.

1985-01-01

The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

Neutronic design for a 100MW{sub th} Small modular natural circulation lead or lead-alloy cooled fast reactors core

Energy Technology Data Exchange (ETDEWEB)

Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q., E-mail: shchshch@ustc.edu.cn, E-mail: hlchen1@ustc.edu.cn, E-mail: kulah@mail.ustc.edu.cn, E-mail: zchen214@mail.ustc.edu.cn, E-mail: zengqin@ustc.edu.cn [Univ. of Science and Technology of China, School of Nuclear Science and Technology, Hefei, Anhui (China)

2015-07-01

Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW{sub th} natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

A Multivariate Time Series Method for Monte Carlo Reactor Analysis

International Nuclear Information System (INIS)

Taro Ueki

2008-01-01

A robust multivariate time series method has been established for the Monte Carlo calculation of neutron multiplication problems. The method is termed Coarse Mesh Projection Method (CMPM) and can be implemented using the coarse statistical bins for acquisition of nuclear fission source data. A novel aspect of CMPM is the combination of the general technical principle of projection pursuit in the signal processing discipline and the neutron multiplication eigenvalue problem in the nuclear engineering discipline. CMPM enables reactor physicists to accurately evaluate major eigenvalue separations of nuclear reactors with continuous energy Monte Carlo calculation. CMPM was incorporated in the MCNP Monte Carlo particle transport code of Los Alamos National Laboratory. The great advantage of CMPM over the traditional Fission Matrix method is demonstrated for the three space-dimensional modeling of the initial core of a pressurized water reactor

"Toore puuga kütmine on raiskamine ..." = "Att elda med sur ved är slöseri ..." : [luuletused] / Gunnar D. Hansson ; tlk. Mati Sirkel

Index Scriptorium Estoniae

Hansson, Gunnar D.

2004-01-01

Sisu: "Toore puuga kütmine on raiskamine ..." = "Att elda med sur ved är slöseri ..."; "Esimest päeva rändab tuul ..." = "Första dagen vandrar vinden ..."; "Päike tuli lähemale ..." = "Solen kom närmare ..."; "Nagu oleks kõik möödas ..." = "Som om allt vore över ..."

Control system design for a 100 MW(th) research reactor

International Nuclear Information System (INIS)

Seshadri, S.N.; Ranganath, M.V.; Singh, Manjit.

1983-01-01

This paper presents the computer simulation carried out to evolve a suitable analog controller for a 100 MW(th) heavy water moderated research reactor under construction at Trombay. The control action is based on the average neutron flux in the reactor core and the reactivity is controlled by adjusting the moderator level in the calandria. A dual control scheme controlling the inflow as well as the outflow was adopted in order to fully exploit the capabilities of control elements. For reasons of reliability, the system consists of three identical channels enabling safe operation even under one channel failure. Based on the simulation studies a suitable compensation network was incorporated to achieve satisfactory system response. (author)

The direct conversion of heat into electricity in reactors; Conversion directe de la chaleur en electricite dans les piles

Energy Technology Data Exchange (ETDEWEB)

Devin, B; Bliaux, J; Lesueur, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The direct conversion of heat into electricity by thermionic emission in an atomic reactor has been studied with the triple aim of its utilisation: as an energy source for a space device, at the head of a conventional conversion system in power installations, or finally in association with the thermoelectric conversion in very low power installations. The laboratory experiments were mainly orientated towards the electron extraction of metals and compounds and their behaviour at high temperatures. Converters furnishing up to 50 amps at 0. 4 volts with an efficiency close to 10 p. 100 have been constructed in the laboratory; the emitters were heated by electron bombardment and were composed of tungsten covered with an uranium carbide deposit or molybdenum covered with cesium. The main aspects of the coupling between the converter and the reactor have been covered from the point of view of electronics: the influence of the mismatching of the load on the temperature of the emitter and the influence of thermal flux density on the temperature of the emitter and the stability of the converter. Converters using uranium carbide as the electron emitter have been tested in reactors. Tests have been made under dynamic conditions in order to determine the dynamic characteristics. The load matching curves have been constructed and the overall performances of several cells coupled in such a way as to form a reactor rod have been deduced. This information is fundamental to the design of a control system for a thermionic conversion reactor. The problems associated with the reliability of thermionic converters connected in series in the same reactor rod have been examined theoretically. Finally, the absorption isotherms have been drawn at the ambient temperatures for krypton and xenon on activated carbon with the aim of investigating the escape of fission products in a converter. (author) [French] La conversion directe de chaleur en electricite par emission thermionique dans une

Cleanup Verification Package for the 118-H-6:2, 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils; the 118-H-6:3, 105-H Reactor Fuel Storage Basin and Underlying Soils; the 118-H-6:6 Fuel Storage Basin Deep Zone Side Slope Soils; the 100-H-9, 100-H-10, and 100-H-13 French Drains; the 100-H-11 and 100-H-12 Expansion Box French Drains; and the 100-H-14 and 100-H-31 Surface Contamination Zones

International Nuclear Information System (INIS)

Appel, M.J.

2006-01-01

This cleanup verification package documents completion of removal actions for the 105-H Reactor Ancillary Support Areas, Below-Grade Structures, and Underlying Soils (subsite 118-H-6:2); 105-H Reactor Fuel Storage Basin and Underlying Soils (118-H-6:3); and Fuel Storage Basin Deep Zone Side Slope Soils. This CVP also documents remedial actions for the following seven additional waste sties: French Drain C (100-H-9), French Drain D (100-H-10), Expansion Box French Drain E (100-H-11), Expansion Box French Drain F (100-H-12), French Drain G (100-H-13), Surface Contamination Zone H (100-H-14), and the Polychlorinated Biphenyl Surface Contamination Zone (100-H-31)

Thermoelectric converter for SP-100 space reactor power system

Science.gov (United States)

Terrill, W. R.; Haley, V. F.

1986-01-01

Conductively coupling the thermoelectric converter to the heat source and the radiator maximizes the utilization of the reactor and radiator temperatures and thereby minimizes the power system weight. This paper presents the design for the converter and the individual thermoelectric cells that are the building block modules for the converter. It also summarizes progress on the fabrication of initial cells and the results obtained from the preparation of a manufacturing plan. The design developed for the SP-100 system utilizes thermally conductive compliant pads that can absorb the displacement and distortion caused by the combinations of temperatures and thermal expansion coefficients. The converter and cell designs provided a 100 kWe system which met the system requirements. Initial cells were fabricated and tested.

Reactor G1: high power experiments; Experiences a forte puissance

Energy Technology Data Exchange (ETDEWEB)

Laage, F de; Teste du Baillet, A; Veyssiere, A; Wanner, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Retel, H [Societe Rateau, D.E.A. (France)

1957-07-01

The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

Notes on a homogeneous reactor project; Idees sur un projet de reacteur homogene

Energy Technology Data Exchange (ETDEWEB)

Benveniste, J; Bernot, J; Eidelman, D; Grenon, M; Portes, L; Raspaud, G; Tachon, J; Vendryes, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Berthod, L; Cohen de Lara, G; Delachanal, M; Fontanet, P; Halbronn, G [Societe Grenobloise d' Etudes et d' Applications Hydrauliques, 38 (France)

1958-07-01

An attempt has been made to develop certain ideas concerning homogeneous reactors. The project under consideration is based on the simultaneous use of a suspension of uranium dispersed in heavy or light water and of boiling in the reactor for heat extraction. However, the studies of suspensions and of boiling are relatively independent and can also be developed for reactors of different types using one or the other. Our aim is a minimum investment in fissile material; for this we propose to extract the steam directly from the core and to make use of a cyclone to accelerate this extraction; a cyclone-type circulation creating a field of increasing tangential velocities of the fluid towards the axis causes the droplets of vapour to accelerate towards the axial vortex in which they are collected; the steam output is then evacuated to the external heat utilisation system, for example an exchanger of the condenser-boiler type. The input speed of water into the reactor being one of the important parameters in the running of the pile, a spiral supply input chamber is used, allowing this speed to be regulated in amount and direction. (author)Fren. [French] Nous nous sommes attaches a developper certaines idees relatives aux piles homogenes. Le projet que nous etudions est base sur l'emploi simultane d'une suspension contenant de l'uranium disperse dans l'eau legere ou lourde et de l'ebullition dans le reacteur pour l'extraction de chaleur. Neanmoins, les etudes de suspensions et d'ebullition sont relativement independantes et peuvent egalement etre developpees pour des reacteurs de type different utilisant l'une ou l'autre. Le but que nous cherchons a atteindre est un investissement minimum en matiere fissile; pour cela, nous proposons d'extraire directement la vapeur dans le coeur et de recourir a un dispositif cyclone pour accelerer cette extraction; une circulation type cyclone creant un champ de vitesses tangentielles du fluide croissantes veraxe a pour effet d

Startup thaw concept for the SP-100 space reactor power system

Science.gov (United States)

Kirpich, A.; Das, A.; Choe, H.; Mcnamara, E.; Switick, D.; Bhandari, P.

1990-01-01

A thaw concept for a space reactor power system which employs lithium as a circulant for both the heat-transport and the heat-rejection fluid loops is presented. An exemplary thermal analysis for a 100-kWe (i.e., SP-100) system is performed. It is shown that the design of the thaw system requires a thorough knowledge of the various physical states of the circulant throughout the system, both spatially and temporally, and that the design has to provide adequate margins for the system to avoid a structural or thermally induced damage.

Sensor response monitoring in pressurized water reactors using time series modeling

International Nuclear Information System (INIS)

Upadhyaya, B.R.; Kerlin, T.W.

1978-01-01

Random data analysis in nuclear power reactors for purposes of process surveillance, pattern recognition and monitoring of temperature, pressure, flow and neutron sensors has gained increasing attention in view of their potential for helping to ensure safe plant operation. In this study, application of autoregressive moving-average (ARMA) time series modeling for monitoring temperature sensor response characteristrics is presented. The ARMA model is used to estimate the step and ramp response of the sensors and the related time constant and ramp delay time. The ARMA parameters are estimated by a two-stage algorithm in the spectral domain. Results of sensor testing for an operating pressurized water reactor are presented. 16 refs

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Directory of Open Access Journals (Sweden)

Pengcheng Zhao

2016-01-01

Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Turismo en Segovia sur. Análisis y propuestas de mejora

OpenAIRE

Herranz Marcos, María

2015-01-01

Este trabajo hace un análisis sobre el turismo rural y sostenible. El cual se basa en el territorio de Segovia sur. Segovia sur tiene un gran potencial turístico, que en algunas zonas es menos conocido. Por ello la creación de una serie de propuestas (senderismo, cicloturismo) que de forma sostenible potencien el desarrollo del turismo en esa zona. Grado en Turismo

Handbook for the calculation of reactor protections; Formulaire sur le calcul de la protection des reacteurs

Energy Technology Data Exchange (ETDEWEB)

NONE

1963-07-01

This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

The optimised sc dipole of SIS100 for series production

Science.gov (United States)

Roux, Christian; Mierau, Anna; Bleile, Alexander; Fischer, Egbert; Kaether, Florian; Körber, Boris; Schnizer, Pierre; Sugita, Kei; Szwangruber, Piotr

2017-02-01

At the international facility for antiproton and ion research (FAIR) in Darmstadt, Germany, an accelerator complex is developed for fundamental research in various fields of modern physics. In the SIS100 heavy-ion synchrotron, the main accelerator of FAIR, superconducting dipoles are used to bend the particle beam. The fast ramped dipoles are 3 m long super-ferric curved magnets operated at 4.5 K. The demands on field homogeneity required for sufficient beam stability are given by ΔB/B ≤ ±6 · 10-4. An intense measurement program of the First of Series (FoS) dipole showed excellent quench behavior and lower than expected AC losses yielding the main load on the SIS100 cryoplant. The FoS is capable to provide a field strength of 1.9 T. However, with sophisticated measurement systems slight distortions of the dipole field were detected. Those effects were tracked down to mechanical inaccuracies of the yoke proven by appropriate geometrical measurements and simulations. After a survey on alternative fabrication techniques a magnet with a new yoke was built with substantial changes to improve the mechanical accuracy. Its characteristics concerning cryogenic losses, cold geometry and the resulting magnetic-field quality are presented and an outlook on the series production of superconducting dipoles for SIS100 is given.

FFTF operational results: startup to 100 MWd/kg

International Nuclear Information System (INIS)

Baird, Q.L.; Harris, R.A.

1985-01-01

The Fast Flux Test Facility (FFTF) is a 400-MW(t) sodium-cooled fast reactor operating at the Hanford Engineering Development Laboratory in Richland, Washington, to conduct fuels and materials testing in support of the US liquid-metal fast breeder reactor program. Startup and initial power testing included a comprehensive series of nuclear and nonnuclear tests to verify the thermal and neutronic characteristics of the plant and to demonstrate its inherent safety features. Extensive reactor core characterization measurements were completed to provide the neutron and gamma spectra, fission rates, and other physics data needed to design and evaluate tests irradiated in the FFTF. A specially designed series of natural-circulation tests was performed to demonstrate the inherent safety features of the plant. Early in 1982 the FFTF began its first 100-d irradiation cycle. Since that time the plant has operated beyond expectations; it achieved a cycle capacity factor of 99.5% in the most recent irradiation cycle. One hundred fifty fuel assemblies (80 of which are experiments) and over 32,000 individual fuel pins have been irradiated, some in excess of 100 MWd/kg. Specialized equipment and systems unique to sodium-cooled reactor plants have performed well. There have been no sodium leaks in the 6 y of sodium system operation. Liquid-metal system maintenance techniques have been proven reliable. Plant maintenance and operating personnel radiation exposures have been very low. 15 figs

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)

Measurements of reactivity of reactor G1; Mesures de reactivite sur reacteur G1

Energy Technology Data Exchange (ETDEWEB)

Bernot, J; Koechlin, J C; Portes, L; Teste du Bailler, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [French] Nous exposons et discutons diverses methodes utilisees, lors de l'etude physique du reacteur G1, pour determiner les variations du facteur de multiplication effectif consecutives a un changement donne dans la geometrie du milieu multiplicateur. La comparaison des resultats obtenus par diverses methodes nous a permis de tester leur validite et d'en preciser les conditions d'emploi. Dans une premiere partie, nous exposons les principes utilises et leurs domaines de validite. Dans une seconde partie nous donnons les resultats experimentaux obtenus avec quelques indications sur leur comparaison avec les estimations theoriques. (auteur)

Thermoelectric converter for SP-100 space reactor power system

International Nuclear Information System (INIS)

Terrill, W.R.; Haley, V.F.

1986-01-01

Conductively coupling the thermoelectric converter to the heat source and the radiator maximizes the utilization of the reactor and radiator temperatures and thereby minimizes the power system weight. This paper presents the design for the converter and the individual thermoelectric cells that are the building block modules for the converter. It also summarizes progress on the fabrication of initial cells and the results obtained from the preparation of a manufacturing plan. The design developed for the SP-100 system utilizes thermally conductive compliant pads that can absorb the displacement and distortion caused by the combinations of temperatures and thermal expansion coefficients. The converter and cell designs provided a 100 kWe system which met the system requirements. Initial cells were fabricated and tested. The manufacturing plan showed that the chosen materials and processes are compatible with today's production techniques, that the production volume can readily be achieved and that the costs are reasonable

Reactor IPEN/MB-01 dosimetry using TLDs

Energy Technology Data Exchange (ETDEWEB)

Cavalieri, Tássio A.; Siqueira, Paulo T.D.; Yoriyaz, Hélio, E-mail: tassio.cavalieri@usp.br [Instituto de Pesquisas Energética s e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

2017-07-01

This paper is a preliminary study on the use of reactor IPEN/MB-01 as standard radiation source for mixed field dosimetry studies. As a first step on this attempt, simulations and experiments, evaluating the neutron and gamma field distributions, were performed and compared. TLDs are widely employed in dose measurements and the TLD 100 / TLD 700 pair conforms with ICRU recommendations for mixed field dosimetry. In this study, TLD irradiations were performed in the IPEN/MB-01 nuclear reactor. IPEN/MB-01 reactor is zero power reactor widely used to perform reactor physics experiments. Its neutron flux distribution is well known for a variety of reactor core configurations. However, the photon fluxes are unknown. A series of experiments with TLD 100 and TLD 700 were performed for two different core configurations (rectangular and cylindrical with a central flux trap). Simulations with MCNP5 for these two configurations were also done, and neutron and gamma fluxes distributions along the core were computed. The responses of TLD 100 and TLD 700 were compared with simulated fluxes and showing a good agreement between them. This paper presents the results of the experiments done so far given the status of the study under way in order to couple IPEN/MB-01 and TLD 100 / 700 pair into a mixed field calibration methodology. (author)

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)

History of 100-B Area

International Nuclear Information System (INIS)

Wahlen, R.K.

1989-10-01

The initial three production reactors and their support facilities were designated as the 100-B, 100-D, and 100-F areas. In subsequent years, six additional plutonium-producing reactors were constructed and operated at the Hanford Site. Among them was one dual-purpose reactor (100-N) designed to supply steam for the production of electricity as a by-product. Figure 1 pinpoints the location of each of the nine Hanford Site reactors along the Columbia River. This report documents a brief description of the 105-B reactor, support facilities, and significant events that are considered to be of historical interest. 21 figs

Inmigración alemana en el sur de Chile, siglo XIX

OpenAIRE

Winkler, Lisette

2009-01-01

Presentación (en español) Durante aproximadamente un mes de trabajo de campo en Chile, Lisette Winkler (EScoM-FMSH) ha grabado una serie de entrevistas que les hizo a los descendientes de los colonos alemanes quienes se instalaron en los alrededores del lago Llanquihue en el siglo XIX. Los documentos audiovisuales fueron recopilados en dos aldeas del sur del país, Frutillar y Puerto Octay. Recogen tanto testimonios orales de los descendientes como tomas panorámicas de los paisajes del sur, qu...

Well Completion Report for the Fiscal Year 1999 Drilling Within the Chromium Plume West of the 100-D/DR Reactors

International Nuclear Information System (INIS)

Ford, B. H.

1999-01-01

This report describes the fiscal year (FY) 1999 field activities associated with installing 12 groundwater monitoring wells in the vicinity of the 100-D Area chromium plume west of the 100-D/DR Reactors (100-HR-3 Operable Unit [OU]). The wells were installed to further investigate the extent of the hexavalent chromium hot spot west of the 100-D/DR Reactors and to support future remedial action decisions associated with the 100-HR-3 OU. These wells were designed for multi-purpose use (i.e., monitoring, extraction, and injection). In addition, one of the wells was installed to support the initial deployment of the In Situ Redox Manipulation (ISRM) technology to remediate the chromium plume

Commercial nuclear reactors and waste: the current status

International Nuclear Information System (INIS)

Platt, A.M.; Robinson, J.V.

1980-04-01

During the last five years, the declared size of the commercial light water reactor (LWR) nuclear power industry in the US has steadily decreased. As of January 1980, the total number of power plants had dropped to 191 from the 226 in December 31, 1974. At least another nine were cancelled in the last few months. This report was developed as the first of a series to track implications to waste management due to such changes in the declared size of the industry. For the presently declared size, key conclusions are: the declared reactors will peak at a capacity of 162 GWe and consume about 10 6 MTU as enrichment feed. As few as two repositories of about 100,000 MTHM capacity each would hold the waste. Predisposal storage (reactor basins and AFRs) would peak at less than 100,000 MTHM (in the year 2020) with one repository opening in the year 1997 and the other in the year 2020. Most of the 100,000 MTHM would have to be in AFR storage unless current practice regarding reactor basin size was radically changed

Data quality objectives summary report for the 100-BC-1, 100-BC-2, and 100-DR-1, and 100-DR-2 group 3 waste sites

International Nuclear Information System (INIS)

1997-03-01

The 100-BC-1, 100-BC-2, 100-DR-1, and 100-DR-2 Group 3 waste sites contain 22 past-practice liquid waste disposal sites and process effluent piping associated with four plutonium-production nuclear reactors that operated from 1944 to 1967. The 100-BC-1, 100-BC-2, 100-DR-1, and 100-DR-2 Group 3 waste sites are the third set of Hanford 100 Area sites to undergo remediation to the extent practicable. Like the sites listed in Groups 1 and 2, the Group 3 sites are considered high-priority because of the contaminants present and their proximity to the Columbia River. Remediation of the 100-BC-1, 100-HR-1 and 100-DR-1 radioactive liquid waste sites is planned to occur in two phases: The first phase, which has been completed, was a demonstration project in the 100-B/C Area to test field techniques and acquire contamination data. The second phase is full-scale remediation of all the reactor areas, starting in the 10-B/C Area, using the field experience gained in the first phase and each subsequent reactor area remediation. This document provides the DQO in support of remediation sampling and analysis at selected sites in the 100-B/C and 100-D Areas

Inmigración alemana en el sur de Chile, siglo XIX

Directory of Open Access Journals (Sweden)

Lisette Winkler

2009-01-01

Full Text Available Presentación (en español Durante aproximadamente un mes de trabajo de campo en Chile, Lisette Winkler (EScoM-FMSH ha grabado una serie de entrevistas que les hizo a los descendientes de los colonos alemanes quienes se instalaron en los alrededores del lago Llanquihue en el siglo XIX. Los documentos audiovisuales fueron recopilados en dos aldeas del sur del país, Frutillar y Puerto Octay. Recogen tanto testimonios orales de los descendientes como tomas panorámicas de los paisajes del sur, qu...

Analyses of newly digitised and reconstructed snow series over the last 100+ years in Switzerland

Science.gov (United States)

Scherrer, S. C.; Wüthrich, C.; Croci-Maspoli, M.; Appenzeller, C.

2010-09-01

Snow is an important socio-economic factor in the Swiss Alpine region (tourism, hydro-electricity, drinking water) and responsible for considerable natural hazards such as avalanches. In addition, high-quality long-term snow series can be used as an excellent indicator of climate change. The objectives of this study are threefold. First, suitable long-term snow series from different altitudes and regions in Switzerland have been selected, missing data digitized and the entire series quality checked. Second, the long-term snow series have been used for trend analyses over a time period >100 years. Third, snow depth series have been reconstructed using daily new snow, temperature and precipitation as input variables. This made it possible to analyse snow depth related variables such as days with snow pack. Results show that the snow cover is varying substantially on seasonal and decadal time scales. The analyses of the decadal new snow trends during the last 100 years shows unprecedented low new snow sums in the winter seasons (DJF) of the 1990s. The 100 year trend of days with snow pack reveals a significant decrease for stations below 800 m asl in the winter season (DJF) and for stations around 1800 m asl in spring (MAM). Similar results were found for seasonal new snow sums. The results of the trend analyses are also discussed with respect to temperature and precipitation trends. Finally we will also shortly discuss how especially "precious" snow measurements have been identified and incorporated in a National Basic Climatological Network (NBCN) as well as in the Global Climate Observing System (GCOS).

Results of Cavity Series Fabrication at Jefferson Laboratory for the Cryomodule 'R100'

International Nuclear Information System (INIS)

Marhauser, F.; Clemens, W.A.; Drury, M.A.; Forehand, D.; Henry, J.; Manning, S.; Overton, R.B.; Williams, R.S.

2011-01-01

A series production of eight superconducting RF cavities for the cryomodule R100 was conducted at JLab in 2010. The cavities underwent chemical post-processing prior to vertical high power testing and routinely exceeded the envisaged performance specifications. After cryomodule assembly, cavities were successfully high power acceptance tested. In this paper, we present the achievements paving the way for the first demonstration of 100 MV (and beyond) in a single cryomodule to be operated at CEBAF.

Application of expert systems to heat exchanger control at the 100-megawatt high-flux isotope reactor

International Nuclear Information System (INIS)

Clapp, N.E. Jr.; Clark, F.H.; Mullens, J.A.; Otaduy, P.J.; Wehe, D.K.

1985-01-01

The High-Flux Isotope Reactor (HFIR) is a 100-MW pressurized water reactor at the Oak Ridge National Laboratory. It is used to produce isotopes and as a source of high neutron flux for research. Three heat exchangers are used to remove heat from the reactor to the cooling towers. A fourth heat exchanger is available as a spare in case one of the operating heat exchangers malfunctions. It is desirable to maintain the reactor at full power while replacing the failed heat exchanger with the spare. The existing procedures used by the operators form the initial knowledge base for design of an expert system to perform the switchover. To verify performance of the expert system, a dynamic simulation of the system was developed in the MACLISP programming language. 2 refs., 3 figs

Generalities on the dynamic behaviour of rapid reactors. Preliminary studies on Rapsodie; Generalites sur le comportement dynamique des piles rapides. Etudes preliminaires de rapsodie

Energy Technology Data Exchange (ETDEWEB)

Campan, J L; Chaumont, J P; Clauzon, P P; Ghesquiere, G; Leduc, J; Schmitt, A P; Zaleski, C P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1963-07-01

The study of the dynamic behaviour of fast reactors may be divided into three section: 1. Stability studies around equilibrium power only the linear case was examining. S. Transient studies in the case of usual reactor operation (shut down, scram, etc.) with thermal shocks evaluation, for instance. 3. Explosion studies, for the maximum credible accidents. This report presents the status of the studies performed at the 'Physics Research Department' at Cadarache. Methods used are detailed and illustrated with the results obtained on a preliminary metallic core of the Rapsodie Reactor. (authors) [French] Le comportement dynamique des piles rapides, se presente tout naturellement sous trois aspects: 1. Etude de stabilite autour d'un regime d'equilibre (nous nous sommes bornes ici au cas lineaire). 2. Etude de regimes transitoires lors des operations normales de pile (arret, arret d'urgence, etc.) avec evaluation des chocs thermiques par exemple. 3. Etude des regimes transitoires de caractere explosif lors des accidents les plus graves possibles. Ce rapport presente l'etat des etudes a la date du 20 decembre 1961 a la Section d'Etudes de Piles Rapides a CADARACHE. Les methodes employees ont ete detaillees et illustrees a partir des resultats obtenus sur une premiere version 'combustible metallique' de Rapsodie. (auteurs)

Criticality-safety analyses of compacted and water-flooded. SP-100 reactors

International Nuclear Information System (INIS)

Brandon, D.I.; Sapir, J.L.

1986-01-01

Reactivity calculations were performed to determine the sensitivity of three liquid metal-cooled, fast reactor designs to various accident environments. The concepts, proposed for the SP-100 Space Nuclear Power Program, included one thermionic and two fuel-pin designs. Numerous models of each core were developed to analyze the effect of core compaction and of water-flooded lattice spreading. Results indicate that those designs incorporating in-core control are least affected by core compaction and that the thermonic concept can best withstand expansion of the flooded fuel element array

Survey of radiological contaminants in the near-shore environment at the Hanford Site 100-N Area reactor

International Nuclear Information System (INIS)

Van Verst, S.P.; Albin, C.L.; Patton, G.W.; Blanton, M.L.; Poston, T.M.; Cooper, A.T.; Antonio, E.J.

1998-09-01

Past operations at the Hanford Site 100-N Area reactor resulted in the release of radiological contaminants to the soil column, local groundwater, and ultimately to the near-shore environment of the Columbia River. In September 1997, the Washington State Department of Health (WDOH) and the Hanford Site Surface Environmental Surveillance Project (SESP) initiated a special study of the near-shore vicinity at the Hanford Site's retired 100-N Area reactor. Environmental samples were collected and analyzed for radiological contaminants ( 3 H, 90 Sr, and gamma/ emitters), with both the WDOH and SESP analyzing a portion of the samples. Samples of river water, sediment, riverbank springs, periphyton, milfoil, flying insects, clam shells, and reed canary grass were collected. External exposure rates were also measured for the near-shore environment in the vicinity of the 100-N Area. In addition, samples were collected at background locations above Vernita Bridge

Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

International Nuclear Information System (INIS)

Fletcher, C.D.

1986-01-01

The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

L'effet ENSO Sur les précipitations et les écoulements au XXème siècle - exemple de l'Equateur

Directory of Open Access Journals (Sweden)

1993-01-01

Full Text Available Sur le littoral équatorien, l'analyse statistique des données pluviométriques disponibles, qui s'étalent sur une cinquantaine d'années, a permis de mettre en évidence deux résultats principaux : 1 -les effets de ENSO ne sont pas tous aussi négatifs qu'on le pense habituellement 2 - la fréquence de retour du ENSO 1982-1983 est supérieure à 1000 ans et son impact sur le milieu morpho-dynamique va au-delà des 100 000 km2. IMPACTO DE ENSO EN LAS PRECIPITACIONES Y ESCURRIMIENTOS DURANTE EL SIGLO XX - EL CASO DEL ECUADOR. En la región litoral del Ecuador, un análisis estadístico de las series pluviométricas disponibles, alrededor de cincuenta años, acarrea dos resultados principales: 1 - los efectos de ENSO no son todos tan negativos como se piensa habitualmente 2 - el período de retorno del ENSO 1982-1983 es mayor de los mil años y su impacto sobre el medio geomorfodinámico abarca más de 100 000 km2. ENSO IMPACT ON RAINFALL AND RUNOFF IN XXth CENTURY - CASE OF ECUADOR. On the Pacific coastal region of Ecuador, a statistical analysis of the largest available pluviometric records, about fifty years, yields two main results: 1 - ENSO effects are not so negative as one usually comments 2 - the return period of ENSO 1982-1983 occurrence is over one thousand years and the geomorphodynamic impact acts upon 100 000 km2.

Study of isotopic exchange reactors (1961); Etude des reacteurs d'echange isotopique (1961)

Energy Technology Data Exchange (ETDEWEB)

Grandcollot, P; Dirian, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1961-07-01

A study is made of the general case of the theory of first-order isotopic chemical exchange between a gaseous and a liquid phase in a reactor, starting from fundamental reaction kinetics data, and without making any limiting hypothesis concerning the value of the separation factor. The cases of counter-current reactors and of co-current reactors are considered successively. The general deuterium conservation equation requires the definition of the quotient of the reactor; the performances of this reactor are characterised by its overall efficiency. The idea of the ratio is introduced because it represents a convenient intermediary in the calculations. The search for an additive value for reactors in series leads logically to the defining of an exchange capacity, and a total efficiency, or number of theoretical reactors. This method of expressing the performances of a reactor is more general than the efficiency due to Murphee which only has a physical significance in the particular case of homogeneous liquid reactors. The relationships between these various quantities are established, and the representation due to Mc Cabe and Thiele is generalized. The reactor performances are linked to the first - order reaction kinetics by the transfer number. The relationships are given for a certain number of concrete cases. Finally the application of these calculations is given, together with the approximations necessary in the case where, because of the presence of several components in each phase, the exchange reaction no longer obeys a single kinetic law. (authors) [French] On examine dans le cas general la theorie d'un reacteur quelconque pour l'echange chimique isotopique du premier ordre entre une phase gazeuse et une phase liquide, a partir des donnees fondamentales sur la cinetique de la reaction, sans faire aucune hypothese limitative sur le cas des reacteurs a contre ourant, puis celui des reacteurs a co-courant. L'equation generale de conservation du deuterium

Completamiento de series de precipitación en la región sur de Ecuador y caracterización de su pluviometría y aridez

Directory of Open Access Journals (Sweden)

Orlando H. Álvarez Hernández

2017-03-01

Full Text Available El objetivo del presente estudio fue determinar el comportamiento de la variable climática "Precipitación" en la Región Sur del Ecuador para su aplicación a diferentes proyectos relacionados con aspectos económicos y sociales en la misma y es una continuación del trabajo realizado anteriormente con variables como temperaturas (media, máxima y mínima, radiación solar global, velocidad del viento y evapotranspiración real en el Área de la Energía, las Industrias y los Recursos Naturales No Renovables y en el Área Agropecuaria y de los Recursos Naturales Renovables de la Universidad Nacional de Loja cuando se modelaron series de radiación solar global, velocidad del viento, temperaturas y evapotranspiración. Las series de información meteorológica de precipitación disponibles fueron procesadas y completadas y posteriormente se obtuvieron mapas de este elemento climático para la Región Sur de Ecuador, así como se calcularon las diferencias mensuales y anuales entre la precipitación y la evapotranspiración real. Este estudio se realizó de agosto de 2016 hasta enero de 2017. Se incluyeron 15 estaciones meteorológicas del instituto nacional de meteorología e hidrología de ecuador (INAMHI para un período de 21 años (1991-2011, aunque se extrapolaron hasta el año 2015. Se utilizaron métodos estadísticos para el completamiento de las series y métodos de geomática para la confección de mapas temáticos.

Systematic staging design applied to the fixed-bed reactor series for methanol and one-step methanol/dimethyl ether synthesis

International Nuclear Information System (INIS)

Manenti, Flavio; Leon-Garzon, Andres R.; Ravaghi-Ardebili, Zohreh; Pirola, Carlo

2014-01-01

This work investigates possible design advances in the series of fixed-bed reactors for methanol and dimethyl ether synthesis. Specifically, the systematic staging design proposed by Hillestad [1] is applied to the water-cooled and gas-cooled series of reactors of Lurgi's technology. The procedure leads to new design and operating conditions with respect to the current best industrial practice, with relevant benefits in terms of process yield, energy saving, and net income. The overall mathematical model for the process simulation and optimization is reported in the work together with dedicated sensitivity analysis studies. - Highlights: • Systematic staging design is applied to methanol and methanol/DME synthesis. • New configurations for the synthesis reactor network are proposed and assessed. • Comparison with the industrial best practice is provided. • Energy-process optimization is performed to improve the overall yield of the process

Le plateau Cyrénien : promontoire africain sur la marge Ionienne The Cyrenian Plateau: African Promontory on the Ionian Margin

Directory of Open Access Journals (Sweden)

Groupe Escarmed

2006-11-01

Full Text Available Le Comité d'Études Pétrolières Marines (CEPM et le Centre National pour l'Exploration des Océans (CNEXO ont mené une série d'études sous-marines des escarpements Ioniens pour en définir l'origine, l'âge et l'évolution. Les résultats présentés ici concernent le plateau Cyrénien, au sud du bassin Ionien profond. Ils comprennent : les données bathymétriques sous forme d'une carte à 1/100 000e hors texte, l'analyse morphologique des reliefs, les déterminations stratigraphiques effectuées sur les échantillons récoltés en dragages ou en plongées, la description sommaire des cinq plongées réalisées sur les escarpements. Les conclusions font ressortir : - l'existence de séries carbonatées de plate-forme d'âge Crétacé inférieur, affectées d'une subsidence rapide; - un changement paléogéographique majeur, post-Albien, qui correspondrait sans doute à la première individualisation du plateau sous-marin; - la distinction d'une série litée, probablement lithologiquement diversifiée, d'âge Crétacé supérieur; - la distinction d'une série supérieure massive, non datée, probablement entièrement carbonatée; - une évolution récente marquée par la formation de figures de dissolution sur les parois rocheuses, la diagenèse d'encroûtements carbonatés sur les pentes envasées. The Comité d'Études Pétrolières Marines (CEPM and the Centre National pour l'Exploitation des Océans (CNEXO have conducted a series of subsea surveys of the Ionian escarpments to determine their origin, age and evolution. The results presented here concern the Cyrenian Shelf, south of the deep Ionian basin. They include the bathymetric data in the form of a map at 1:100,000 (inset, the morphological analysis of the reliefs, the stratigraphic determinations carried out on samples taken by dredging and diving, and the brief description of the five dives carried out on the escarpments. The conclusions reveal the following:(a the existence

Neutron-physical characteristics of the TVRM-100 reactor with ten ring fuel channels

International Nuclear Information System (INIS)

Mikhajlov, V.M.; Myrtsymova, L.A.

1988-01-01

Three-dimensional heterogeneous calculations of TVRM-100 reactor which is a research reactor using enriched fuel with heavy-water moderator, coolant and reflector, are conducted. Achievable burnup depths depending on the number of removable FAs are presented. The maximum non-perturbed thermal neutron flux in the reflector is (2-1.8)x10 15 cm -2 c -1 ; mean flux on the fuel is 2.9x10 14 cm -2 c -1 . Energy release radial non-uniformity is 0.67, maximum bending by FA is ∼3.7. Reactivity temperature effect is negative and is equal to - 0.9x10 -4 grad -1 without accounting for experimental channels. Control rod efficiency in the radial reflector is high, but their location dose to experimental devices in the high neutron flux area is undesirable. 4 refs.; 5 figs

100-B area technical baseline report

International Nuclear Information System (INIS)

Carpenter, R.W.

1994-01-01

This document supports the environmental remediation effort of the 100-B Area by providing remediation planners with key data that characterize the 100-B and 100-C Reactor sites. It provides operational histories of the 100-B and 100-C Reactors and each of their associated liquid and solid waste sites

100-B area technical baseline report

Energy Technology Data Exchange (ETDEWEB)

Carpenter, R.W.

1994-09-01

This document supports the environmental remediation effort of the 100-B Area by providing remediation planners with key data that characterize the 100-B and 100-C Reactor sites. It provides operational histories of the 100-B and 100-C Reactors and each of their associated liquid and solid waste sites.

Description of work for the drilling within the chromium plume west of 100-D/DR Reactors

International Nuclear Information System (INIS)

Peterson, R.E.; Walker, L.D.

1997-07-01

This document describes the work scope associated with installing four new monitoring wells in the 100-D/DR Area (100-HR-3 Operable Unit). The strategy relies on estimates for flow paths that might have existed during operation of the 100-D Reactor and on experience gained during the recent installation of well 199-D4-1. A Data Quality Objectives (DQO) workshop was held to evaluate data collection needs during well installation. The workshop included input from key project team members and the lead regulatory agency. Decisions concerning data resulting from the DQO process have been incorporated into this document

Advanced CANDU reactors

International Nuclear Information System (INIS)

Dunn, J.T.; Finlay, R.B.; Olmstead, R.A.

1988-12-01

AECL has undertaken the design and development of a series of advanced CANDU reactors in the 700-1150 MW(e) size range. These advanced reactor designs are the product of ongoing generic research and development programs on CANDU technology and design studies for advanced CANDU reactors. The prime objective is to create a series of advanced CANDU reactors which are cost competitive with coal-fired plants in the market for large electricity generating stations. Specific plant designs in the advanced CANDU series will be ready for project commitment in the early 1990s and will be capable of further development to remain competitive well into the next century

Application of distantiometric protections to 500 kV lines with AL 100% series compensation; Aplicacion de protecciones distanciometricas a lineas de 500 kV con compensacion serie AL 100%

Energy Technology Data Exchange (ETDEWEB)

Gaudio, Juan Carlos; Pesce, Jose Juan O.; Faure, Carlos Hugo [TRANSENER S.A., Buenos Aires (Argentina)

2001-07-01

The TRANSENER S.A. viewing the rising of the transmission capacity of the Northeastern Argentine Generators, assumed the installation costs of compensation by series capacitors. In the middle of the year 2000 was put into service a series capacitor bank on the extremity Recreo of the 500 kV to the Malvinas Argentina of the TRANSENER S.A.. The reactance compensation of the line in about 100% was responsible by the necessity of replacing the existent analogic electronic protection by new numerical distantiometric protections on the El Bracho and Malvinas Argentinas transformer station. This paper presents the imposed conditions for the selected line protection, the prior studies posterior the mentioned selection for validating the application, and the difficulties necessary to the task performing as well.

LWR [Light Water Reactor] power plant simulations using the AD10 and AD100 systems

International Nuclear Information System (INIS)

Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.

1989-01-01

Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

Science.gov (United States)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

SP-100/Brayton power system concepts

International Nuclear Information System (INIS)

Owen, D.F.

1989-01-01

Use of closed Brayton cycle (CBC) power conversion technology has been investigated for use with SP-100 reactors for space power systems. The CBC power conversion technology is being developed by Rockwell International under the Dynamic Isotype Power System (DIPS) and Space Station Freedom solar dynamic power system programs to provide highly efficient power conversion with radioisotype and solar collector heat sources. Characteristics including mass, radiator area, thermal power, and operating temperatures for systems utilizing SP-100 reactor and CBC power conversion technology were determined for systems in the 10-to 100-kWe power range. Possible SP-100 reactor/CBC power system configurations are presented. Advantages of CBC power conversion technology with regard to reactor thermal power, operating temperature, and development status are discussed

Seismic behaviour of gas cooled reactor components

International Nuclear Information System (INIS)

1990-08-01

On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

Development Plan and R and D Status of China Lead-based Reactor

Energy Technology Data Exchange (ETDEWEB)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS [Institute of Nuclear Energy Safety Technology, Beijing (Switzerland)

2013-07-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Development Plan and R and D Status of China Lead-based Reactor

International Nuclear Information System (INIS)

Wu, Yican; Bai, Yunqing; Song, Yong; Li, Yazhou; Team, FDS

2013-01-01

Chinese Academy of Sciences (CAS) launched an engineering project to develop ADS system and lead-based reactors named China LEAd-based Reactor (CLEAR) series. The Institute of Nuclear Energy Safety Technology (INEST) will be responsible for the CLEAR design and R and D. In this project, CAS plans to develop the lead-based reactors through 3 phases which are 10MWth lead based research reactor (CLEAR-I), 100MWth lead-based experimental reactor (CLEAR-II), 1000MWth lead-based demonstration reactor (CLEAR-III). As a pre-testing facility, a lead-based zero-power reactor (CLEAR-0) is required to be built before CLEAR-I construction and operation. The new conceptual design of lead-based reactors, including hydrogen production, tritium production for fusion energy and thorium utilization, is also on-going. Lead-lithium cooled fusion reactor blanket design and lead-lithium experimental loops have been developed more than 10 years. CLEAR series reactor conceptual design has been finished and detailed engineering design for CLEAR-I is underway. The R and D activities for CLEAR reactor including design and safety software, key components, structural materials, lead-based experimental loops and neutronics experimental platform are developing. Series of liquid lead-based experimental loops named DRAGON (Lead-Lithium) and KYLIN (Lead-Bismuth) have already been built or on constructing to performed experiments investigating the structure material corrosion issues and the thermal-hydraulic properties of lead-based coolant. The Highly Intensified D-T Neutron Generator HINEG for neutron experiment and software validation will be constructed. Series advanced reactor design software and nuclear library have been developed for lead-alloy cooled reactor, including CAD based Multi-Functional 4D Neutronics Simulation System (Visual Bus), Monte Carlo Automatic Modeling Program for Radiation Transport Simulation (MCAM), Super Monte Carlo Simulation Program (SuperMC), Nuclear Radiation

Pebble red modular reactor - South Africa

International Nuclear Information System (INIS)

Fox, M.; Mulder, E.

1996-01-01

In 1995 the South African Electricity Utility, ESKOM, was convinced of the economical advantages of high temperature gas-cooled reactors as viable supply side option. Subsequently planning of a techno/economic study for the year 1996 was initiated. Continuation to the construction phase of a prototype plant will depend entirely on the outcome of this study. A reactor plant of pebble bed design coupled with a direct helium cycle is perceived. The electrical output is limited to about 100 MW for reasons of safety, economics and flexibility. Design of the reactor will be based on internationally proven, available technology. An extended research and development program is not anticipated. New licensing rules and regulations will be required. Safety classification of components will be based on the merit of HTGR technology rather than attempting to adhere to traditional LWR rules. A medium term time schedule for the design and construction of a prototype plant, commissioning and performance testing is proposed during the years 2002 and 2003. Pending the performance outcome of this plant and the current power demand, series production of 100 MWe units is foreseen. (author)

Radio-chemical dosage of {sup 90}Sr in large volumes of drinking water; Dosage radiochimique du {sup 90}Sr sur des volumes importants d'eaux potables

Energy Technology Data Exchange (ETDEWEB)

Jeanmaire, L; Patti, F; Bullier, D [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1965-07-01

I. Principle of the method: 1. Fixing on a resin of all the cations present in the water. 2. Elution using 5 N nitric acid and precipitation of strontium as the carbonate. 3. Concentration of the strontium using the fuming nitric acid method. 4. Purification of the strontium on a resin by selective elution with ammonium citrate. 5. The strontium-90 is measured by separation at the {sup 90}Y equilibrium in the form of the oxalate which is then counted. II. Advantages of the method The concentration of the radio-activity starting from large volumes (100 l) is generally tedious but this method which makes use of a fixation on a cationic resin makes it very simple. The rest of the method consists of a series of simple chemical operations using ion-exchange on resins and coprecipitation. Finally, it is possible to dose stable strontium. (authors) [French] I. Principe du dosage 1. Fixation sur resine de tous les cations presents dans l'eau, 2. Elution par l'acide nitrique 5 N et precipitation du strontium sous forme de carbonate. 3. Concentration du strontium par la methode a l'acide nitrique fumant. 4. Purification du strontium sur resine par elution selective au citrate d'ammonium. 5. Le strontium-90 est dose par separation a l'equilibre du {sup 90}Y sous forme d'oxalate qui est compte. II. Interet de la methode La concentration de la radioactivite a partir de volumes importants (100 l) est generalement fastidieuse, la technique proposee rend cette phase tres simple en utilisant une fixation sur resine cationique. Le reste de la technique est une suite d'operations chimiques simples a realiser, faisant appel a l'echange d'ions sur resine et a la coprecipitation. Enfin, il est possible de realiser le dosage du strontium stable. (auteurs)

The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

Energy Technology Data Exchange (ETDEWEB)

Barral, J C [Electricite de France (EDF), 75 - Paris (France); Zaetta, A [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G [CEA/Saclay, DEN, 91 - Gif sur Yvette (France); and others

2000-07-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

Analyse de l'effet de la suréducation sur l'efficacité technique des ...

African Journals Online (AJOL)

de tire-au-flanc, et l'état de santé (Vroom 1964 ; Sheppard & Herrick 1972 ; ... impact négatif significatif de la suréducation sur la satisfaction au travail ; son ..... présence syndicale sur la performance financière de 250 entreprises américaines.

SP-100 Program: space reactor system and subsystem investigations

International Nuclear Information System (INIS)

Harty, R.B.

1983-01-01

For a space reactor power system, a comprehensive safety program will be required to assure that no undue risk is present. This report summarizes the nuclear safety review/approval process that will be required for a space reactor system. The documentation requirements are presented along with a summary of the required contents of key documents. Finally, the aerospace safety program conducted for the SNAP-10A reactor system is summarized. The results of this program are presented to show the type of program that can be expected and to provide information that could be usable in future programs

La cooperación Sur-Sur agrícola argentina con África Subsahariana: una historia que comienza

OpenAIRE

Morasso, Carla

2015-01-01

[es] En la última década Argentina ha sido un actor dinámico de la cooperación Sur-Sur. Sus acciones se han dirigido principalmente hacia América Latina, pero también se han promovido los vínculos con Asia y África. El artículo analiza particularmente la cooperación Sur-Sur entre Argentina y países de África Subsahariana en materia de desarrollo agrícola en el período 2003-2013, donde se destacan los roles del Fondo Argentino de Cooperación Sur-Sur y Triangula...

Can a nuclear reactor operate for 100 years?

International Nuclear Information System (INIS)

Hertel, O.

2010-01-01

The TWR (Travelling Wave Reactor) concept was invented in the fifties, then forgotten and it reappeared in 2001 but it was considered too immature to be selected for the fourth generation of nuclear reactors, now an American company 'Terrapower' proposes one whose design is given in the article. This TWR operates with depleted uranium, only the lower part of the fuel rod involves uranium fuel with a civil enrichment ratio (less that 20%). The lower part of the fuel will ignite the fission reaction and enrich the part of fuel just above through neutron absorption. The burning part of the fuel will move up progressively. The main advantage of this reactor is that it can operate for decades without maintenance nor fuel loading. The principle is right on the paper but requires huge technological work to select materials and systems that will be able to withstand decades of operation time in harsh conditions. (A.C.)

Calibration of RB reactor power

International Nuclear Information System (INIS)

Sotic, O.; Markovic, H.; Ninkovic, M.; Strugar, P.; Dimitrijevic, Z.; Takac, S.; Stefanovic, D.; Kocic, A.; Vranic, S.

1976-09-01

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8√2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation

Adsorption of Hydrophobically Modified Polyelectrolytes on Hydrophobic Substrates Adsorption de polyélectrolytes modifiés hydrophobiquement sur les substrats hydrophobes

Directory of Open Access Journals (Sweden)

Mays J. W.

2006-12-01

Full Text Available A series of diblock copolymers, poly (tert-butyl styrene-sodium poly (styrene sulfonate with different molecular weight and percentage of sulfonation have been used to study the effect of polymer structure on its adsorption behavior onto hydrophobically modified silicon wafers. The percentage of the hydrophobic block varies from 3. 6-8. 9%. Previous studies show that salt concentration is very important for the adsorption of such polyelectrolytes onto silica surfaces. Octadecyltriethoxysilane (OTE has been used to modify the silicon wafer which changes the water contact angle from 50° on unmodified silica to 100° to 120°. On this hydrophobic surface, we found that the adsorption of these slightly hydrophobically modified polyelectrolytes is close to the 4/23rd power of salt concentration predicted by a recent model. The grafting density is also consistent with a dependence on the length of the hydrophobic block to the -12/23rd power, and the length of the polyelectrolyte block to the -6/23rd power, predicted by this model. Une série de copolymères à diblocs poly (tert-butyle styrène-sodium (sulfonate de polystyrène de masses moléculaires et pourcentages de sulfonation différents ont été utilisés pour étudier les effets de la structure du polymère sur son pouvoir d'adsorption sur des surfaces de silicium modifiées hydrophobiquement. Le pourcentage du bloc hydrophobe varie de 3,6 à 8,9%. Les études antérieures montrent que la concentration saline est très importante pour l'adsorption de ces polyélectrolytes sur les surfaces de silice. Nous avons utilisé l'octadecyltriéthoxysilane (OTE pour modifier la surface de silicium qui change l'angle de contact de l'eau de 50° sur la silice non modifiée à une valeur comprise entre 100° et 120° sur la silice modifiée. Sur cette surface hydrophobe, nous constatons que l'adsorption de ces polyélectrolytes légèrement modifiés hydrophobiquement est proche de la loi puissance 4

A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

International Nuclear Information System (INIS)

Minatsuki, Isao; Otani, Tomomi; Shimizu, Katsusuke; Mizokami, Yorikata; Oyama, Sunao; Tsukamoto, Hiroki

2014-01-01

A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R and D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies

A development strategy for the business plan of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is)

Energy Technology Data Exchange (ETDEWEB)

Minatsuki, Isao, E-mail: isao_minatsuki@mhi.co.jp [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Otani, Tomomi; Shimizu, Katsusuke [Mitsubishi Heavy Industries, Ltd., 16-5, Konan 2-Chome, Minato-ku, Tokyo (Japan); Mizokami, Yorikata; Oyama, Sunao; Tsukamoto, Hiroki [Mitsubishi Heavy Industries, Ltd., 1-1 Wadasaki-cho 1-Chome, Hyogo-ku, Kobe (Japan)

2014-05-01

A business plan and a new concept of Mitsubishi Small-sized High Temperature Gas-cooled Modular Reactor (MHR-50/100is) has been investigated toward a commercialization in near future by Mitsubishi Heavy Industries cooperated with Japan Atomic Energy Agency (JAEA) in Japan. The potential market of small sized reactor is expected to increase from the points of view of smaller investment, industrial use of the nuclear heat and IPP (Independent Power Producer). Especially minimization of construction unit cost including R and D and plant construction period are important issues in order to realize a business plan for them. The study includes four pertinent subject areas of (1) a market analysis, (2) a conceptual design, (3) improvement of safety design and (4) plant dynamics. In summary, the MHR-50/100 is designed to target a short construction period, competitive cost, and an inherent safety feature while applying only the verified technology of the High Temperature Engineering Test Reactor (HTTR) of JAEA or conventional technologies.

Unión del magreb, un caso de regionalismo Sur-Sur

Directory of Open Access Journals (Sweden)

SofianE Bouhdiba

2009-01-01

Full Text Available Se exploran brevemente las fuerzas y limitaciones de la Unión del Magreb (UMA como una experiencia de regionalismo Sur-Sur, en el marco de mayor extensión. Se plantea la necesidad de que la UMA redireccione sus intereses, estrategias y alianzas hacia el sur, puesto que se ha vuelto un instrumento de intercambios comerciales con la Unión Europea y un eje estratégico de lucha contra el terrorismo para Estados Unidos. Además se invita a la misma UMA a que considere los alcances de instituciones de integración en América Latina y Asia a fin de fortalecerse como bloque.

Fast wave current drive in reactor scale tokamaks

International Nuclear Information System (INIS)

Moreau, D.

1992-01-01

The IAEA Technical Committee Meeting on Fast Wave Current Drive in Reactor Scale Tokamaks, hosted by the Commissariat a l'Energie Atomique (CEA), Departement de Recherches sur la Fusion Controlee (Centres d'Etudes de Cadarache, under the Euratom-CEA Association for fusion) aimed at discussing the physics and the efficiency of non-inductive current drive by fast waves. Relevance to reactor size tokamaks and comparison between theory and experiment were emphasized. The following topics are described in the summary report: (i) theory and modelling of radiofrequency current drive (theory, full wave modelling, ray tracing and Fokker-Planck calculations, helicity injection and ponderomotive effects, and alternative radio-frequency current drive effects), (ii) present experiments, (iii) reactor applications (reactor scenarios including fast wave current drive; and fast wave current drive antennas); (iv) discussion and summary. 32 refs

Vertigo in childhood: a retrospective series of 100 children.

Science.gov (United States)

Batu, Ezgi Deniz; Anlar, Banu; Topçu, Meral; Turanlı, Güzide; Aysun, Sabiha

2015-03-01

Evaluation and management of vertigo in children vary between institutions and medical specialties. The aim of this study is to describe the characteristics of vertigo in children presenting to a pediatric neurology referral center and to investigate the relationship between vertigo and migraine. Patients vertigo to Hacettepe University Ihsan Dogramaci Children's Hospital Neurology Unit between January 1996-January 2012 were included (n = 100). Data were obtained from patient files and phone interviews. Mean age was 7.5 years. The most common etiological groups were benign paroxysmal vertigo of childhood (BPVC) (39%), psychogenic vertigo (21%), epileptic vertigo (15%), and migraine-associated vertigo (MAV) (11%). BPVC was the most common diagnosis in children ≤5 years of age while psychogenic vertigo prevailed in children >5 years. Staring episodes characterized epileptic vertigo patients (p = 0.021) while headache was more often described by MAV patients (p Vertigo attacks >5 min were uncommon in BPVC patients compared to others (p = 0.013). Twenty percent of BPVC patients contacted through phone interviews were experiencing migraine type headaches that started at a median age of 7.5 years. An algorithm for evaluation of children with vertigo was formed based on data obtained from this study and the literature. When this algorithm was applied to 100 cases of this series, 88 (88%) were correctly diagnosed. While most vertigo cases in children can be diagnosed accurately by a detailed medical history, physical and neurological examination, a standard algorithm can help with the correct classification. Copyright © 2014 European Paediatric Neurology Society. Published by Elsevier Ltd. All rights reserved.

The Use of Multi-Reactor Cascade Plasma Electrolysis for Linear Alkylbenzene Sulfonate Degradation

Science.gov (United States)

Saksono, Nelson; Ibrahim; Zainah; Budikania, Trisutanti

2018-03-01

Plasma electrolysis is a method that can produce large amounts of hydroxyl radicals to degrade organic waste. The purpose of this study is to improve the effectiveness of Linear alkylbenzene sulfonate (LAS) degradation by using multi-reactor cascade plasma electrolysis. The reactor which operated in circulation system, using 3 reactors series flow and 6 L of LAS with initial concentration of 100 ppm. The results show that the LAS degradation can be improved multi-reactor cascade plasma electrolysis. The greatest LAS degradation is achieved up to 81.91% with energy consumption of 2227.34 kJ/mmol that is obtained during 120 minutes by using 600 Volt, 0.03 M of KOH, and 0.5 cm of the anode depth.

Más allá de los mitos: análisis de la Cooperación Sur-Sur y Norte-Sur en el Ecuador. Sector Educación, período 2003- 2008

OpenAIRE

Escobar Sánchez, Wladimir Alexander

2012-01-01

La presente tesis de maestría analiza a la cooperación Norte-Sur y Sur-Sur como los dos principales modelos de funcionamiento de la cooperación hoy en día, a través de dos programas de ayuda internacional en el sector educación, en el período 2003-2008. El uno referente a la cooperación Norte-Sur, y el otro a la cooperación Sur-Sur. El propósito que presenta este trabajo académico es explicar a la cooperación internacional al desarrollo tomando en consideración a la teoría real...

Étude sur l'entrepreneuriat dalit : Regard sur 25 % de la population ...

International Development Research Centre (IDRC) Digital Library (Canada)

Étude sur l'entrepreneuriat dalit : Regard sur 25 % de la population indienne. Cette étude, pilotée par Udit Raj, député du parlement indien, aide à mieux comprendre les contraintes que rencontrent les entrepreneurs dalits. Elle permettra de sonder 12 000 entrepreneurs dalits dans 14 états et de recueillir des données ...

Irradiations under magnetic field. Measurement of resistivity sample irradiations between 100 and 500 deg C in a swimming-pool reactor

International Nuclear Information System (INIS)

Pauleve, J.; Marchand, A.; Blaise, A.

1964-01-01

An oven is described which enables the irradiation of small samples in the maximum neutron flux of a swimming-pool reactor of 15 MW (Siloe), at temperatures of between 100 and 500 deg.C defined to ± 0,5 deg.C, The oven is very simple from the technological point of view, and has a diameter of only 27 mm, This permits resistivity measurements to be carried out under irradiation in the reactor, or as another example, it enables irradiations in a magnetic field of 5000 oersteds, created by an immersed solenoid. (authors) [fr

A 100 MWe Advanced Sodium-cooled Fast Reactor (AFR-100)

International Nuclear Information System (INIS)

Grandy, C.; Kim, T.K.; Jin, E.

2013-01-01

• AFR-100 Design development is continuing in the U.S.; • Various innovations are included in the design to understand their feasibility; • Engineering and safety analyses have been performed that demonstrate the inherent safety characteristics of the AFR-100 design during severe accidents; • R&D is being performed on a number of the innovations such as advanced materials, compact fuel handing system, advanced energy conversion system, advanced core design, etc

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Pressurized water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project

Training and teaching with SILOETTE reactor and associated simulators at the Nuclear Research Centre of Grenoble

International Nuclear Information System (INIS)

Destot, M.

1983-10-01

Thanks to its three reactors SILOE (35 MW), MELUSINE (8 MW) and SILOETTE (100 KW), the Reactor Department of the Nuclear Research Centre of Grenoble has gained a considerable experience in the operation and utilization of research and material testing reactors. Inside of this general framework, the Reactor Department of Grenoble has built up a training and teaching centre that has been permanently active since 1975, with the aim of satisfying the considerable needs arising from the development of electro-nuclear power stations. The course is mainly intended for engineers and technicians who will be responsible for running power stations. A thorough series of practical exercices, carried out in the SILOETTE training reactor and in a PWR or in a Gas Cooled Reactor Simulator, desmonstrates the application of the theorical courses and familiarises the trainees with the behaviour of reactors and power stations

SurR9C84A protects and recovers human cardiomyocytes from hypoxia induced apoptosis

Energy Technology Data Exchange (ETDEWEB)

Ashok, Ajay [Nanomedicine-Laboratory of Immunology and Molecular Biomedical Research (NLIMBR), School of Medicine (SoM), Faculty of Health, Centre for Molecular and Medical Research - C-MMR, Deakin University, Waurn Ponds, Victoria 3216 (Australia); Department of Pathology, Case Western Reserve University, 2103 Cornell Rd. WRB 5128, Cleveland, OH 44106-7288 (United States); Kanwar, Jagat Rakesh [Nanomedicine-Laboratory of Immunology and Molecular Biomedical Research (NLIMBR), School of Medicine (SoM), Faculty of Health, Centre for Molecular and Medical Research - C-MMR, Deakin University, Waurn Ponds, Victoria 3216 (Australia); Krishnan, Uma Maheswari [Centre for Nanotechnology & Advanced Biomaterials (CeNTAB), School of Chemical & Biotechnology (SCBT), SASTRA University, Thanjavur 613401 (India); Kanwar, Rupinder Kaur, E-mail: rupinder.kanwar@deakin.edu.au [Nanomedicine-Laboratory of Immunology and Molecular Biomedical Research (NLIMBR), School of Medicine (SoM), Faculty of Health, Centre for Molecular and Medical Research - C-MMR, Deakin University, Waurn Ponds, Victoria 3216 (Australia)

2017-01-01

Survivin, as an anti-apoptotic protein and a cell cycle regulator, is recently gaining importance for its regenerative potential in salvaging injured hypoxic cells of vital organs such as heart. Different strategies are being employed to upregulate survivin expression in dying hypoxic cardiomyocytes. We investigated the cardioprotective potential of a cell permeable survivin mutant protein SurR9C84A, for the management of hypoxia mediated cardiomyocyte apoptosis, in a novel and clinically relevant model employing primary human cardiomyocytes (HCM). The aim of this research work was to study the efficacy and mechanism of SurR9C84A facilitated cardioprotection and regeneration in hypoxic HCM. To mimic hypoxic microenvironment in vitro, well characterized HCM were treated with 100 µm (48 h) cobalt chloride to induce hypoxia. Hypoxia induced (HI) HCM were further treated with SurR9C84A (1 µg/mL) in order to analyse its cardioprotective efficacy. Confocal microscopy showed rapid internalization of SurR9C84A and scanning electron microscopy revealed the reinstatement of cytoskeleton projections in HI HCM. SurR9C84A treatment increased cell viability, reduced cell death via, apoptosis (Annexin-V assay), and downregulated free cardiac troponin T and MMP-9 expression. SurR9C84A also upregulated the expression of proliferation markers (PCNA and Ki-67) and downregulated mitochondrial depolarization and ROS levels thereby, impeding cell death. Human Apoptosis Array further revealed that SurR9C84A downregulated expression of pro-apoptotic markers and augmented expression of HSPs and HTRA2/Omi. SurR9C84A treatment led to enhanced levels of survivin, VEGF, PI3K and pAkt. SurR9C84A proved non-toxic to normoxic HCM, as validated through unaltered cell proliferation and other marker levels. Its pre-treatment exhibited lesser susceptibility to hypoxia/damage. SurR9C84A holds a promising clinical potential for human cardiomyocyte survival and proliferation following hypoxic injury

Structural qualification of the multifunctional instrument tree for installation in double-shell and 100-series single-shell tanks

International Nuclear Information System (INIS)

Strohlow, J.P.

1995-12-01

This document provides the technical basis and methodology for qualifying the multifunctional instrument tree (MIT) structure for installation in double-shell and 100-series single-shell tanks. Structural qualification for MIT installations in specific tanks are also contained in this document

WWER-1000 reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series 12, 'Reactor Simulator Development' (2001). Course material for workshops using a pressurized water reactor (PWR) Simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003) and Training Course Series No. 23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation. N. V. Tikhonov and S. B. Vygovsky of the Moscow Engineering and Physics Institute prepared this report for the IAEA

Boiling water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

Development of toroid-type HTS DC reactor series for HVDC system

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Development of toroid-type HTS DC reactor series for HVDC system

International Nuclear Information System (INIS)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-01-01

Highlights: • The authors developed the 400 mH, 400 A class toroid-type HTS DC reactor system. • The target temperature, inductance and operating current are under 20 K at magnet, 400 mH and 400 A, respectively. All target performances of the HTS DC reactor were achieved. • The HTS DC reactor was conducted through the interconnection operation with a LCC type HVDC system. • Now, the authors are studying the 400 mH, 1500 A class toroid-type HTS DC reactor for the next phase HTS DC reactor. - Abstract: This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

A Novel Dual-Stage Hydrothermal Flow Reactor

DEFF Research Database (Denmark)

Hellstern, Henrik Christian; Becker, Jacob; Hald, Peter

2015-01-01

The dual-stage reactor is a novel continuous flow reactor with two reactors connected in series. It is designed for hydrothermal flow synthesis of nanocomposites, in which a single particle consists of multiple materials. The secondary material may protect the core nanoparticle from oxidation....... The dual-stage reactor combines the ability to produce advanced materials with an upscaled capacity in excess of 10 g/hour (dry mass). TiO2 was synthesized in the primary reactor and reproduced previous results. The dual-stage capability was succesfully demonstrated with a series of nanocomposites incl. Ti...

Construction of the core of the 'heavy water-gas' reactor EL 4; Structures du coeur du reacteur 'eau- lourde-gaz EL 4'

Energy Technology Data Exchange (ETDEWEB)

Bernard, J L; Foulquier, H; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

problem of thermal insulation around a zirconium alloy liner tube. The neutron absorption equivalent is about 1, 1 mm of Al, and the mean loss around 2 p. 100 of the thermal power of the reactor. The methods proposed have proved practicable as a result of important research and developments on automatic remote control for all the operations which make up the sequences of mounting, demounting and repairing of the construction components. In particular the possibilities opened up by the new techniques of welding tubes from the inside have been extended to other problems connected with the assembling of a reactor. (authors) [French] Le coeur de ce reacteur est constitue par une cuve contenant l'eau lourde, cuve traversee d'une serie de tubes de force dans lesquels circule le gaz caloporteur sous pression de 60 at. Les specifications de depart qui ont joue un role important dans la conception de ces structures concernent des aspects de securite de fonctionnement (chargement du combustible par les deux faces du reacteur, remplacement des structures sur les deux faces du reacteur), des necessites neutroniques (absorption des structures minimum, pas du reseau, diametre des tubes de force) et des considerations thermiques (temperature de sortie 500 C). Ces specifications ont entraine une disposition horizontale des tubes de force et des problemes d'encombrement tres delicats qui ont elimine (pour les dimensions d'EL 4) toute possibilite de recourir a des compensateurs de dilatation sur les tubes de force. II s'ensuit un dessin de cuve semi-rigide dans lequel les tubes de force contribuent pour une part importante a la resistance mecanique de l'ensemble en jouant le role de tirant, d'ou des contraintes elevees sur les jonctions et tubes de force (et le choix des alliages de zirconium). Les structures comprennent le tube de force, les jonctions, l'isolement thermique et le tube de guidage. On expose brievement les moyens d'essais mis en oeuvre et les performances de ces diverses

Étude du Partenariat de recherche sur l'influenza aviaire en Asie sur ...

International Development Research Centre (IDRC) Digital Library (Canada)

Étude du Partenariat de recherche sur l'influenza aviaire en Asie sur l'efficacité des mesures de lutte. L'influenza aviaire hautement pathogène (H5N1) est une grave maladie qui touche la volaille et contamine les êtres humains dans une faible proportion. L'Organisation mondiale de la santé croit que le virus pourrait subir ...

Development of toroid-type HTS DC reactor series for HVDC system

Science.gov (United States)

Kim, Kwangmin; Go, Byeong-Soo; Park, Hea-chul; Kim, Sung-kyu; Kim, Seokho; Lee, Sangjin; Oh, Yunsang; Park, Minwon; Yu, In-Keun

2015-11-01

This paper describes design specifications and performance of a toroid-type high-temperature superconducting (HTS) DC reactor. The first phase operation targets of the HTS DC reactor were 400 mH and 400 A. The authors have developed a real HTS DC reactor system during the last three years. The HTS DC reactor was designed using 2G GdBCO HTS wires. The HTS coils of the toroid-type DC reactor magnet were made in the form of a D-shape. The electromagnetic performance of the toroid-type HTS DC reactor magnet was analyzed using the finite element method program. A conduction cooling method was adopted for reactor magnet cooling. The total system has been successfully developed and tested in connection with LCC type HVDC system. Now, the authors are studying a 400 mH, kA class toroid-type HTS DC reactor for the next phase research. The 1500 A class DC reactor system was designed using layered 13 mm GdBCO 2G HTS wire. The expected operating temperature is under 30 K. These fundamental data obtained through both works will usefully be applied to design a real toroid-type HTS DC reactor for grid application.

Innovative Approach to Validation of Ultraviolet (UV) Reactors for Disinfection in Drinking Water Systems

Science.gov (United States)

Slide presentation at Conference: ASCE 7th Civil Engineering Conference in the Asian Region. USEPA in partnership with the Cadmus Group, Carollo Engineers, and other State & Industry collaborators, are evaluating new approaches for validating UV reactors to meet groundwater & sur...

Sur la plurifonctionnalité du discours direct

OpenAIRE

Cigada Sara

2012-01-01

La comparaison entre les résultats de nombreux travaux sur le dialogue dans le texte littéraire, sur l’attestation linguistique de la subjectivité, sur la fonction argumentative des émotions dans le discours et sur les effets de polyphonie, suggère que la structure sémiotique et linguistique du discours direct (DD) se trouve au croisement stratégique de plusieurs axes de la construction discursive. Nous étudions donc l’insertion du DD dans le discours (cf. Rosier 2008: Le discours rapporté en...

The effect of fractionating 100 R hard X-ray doses on the hemogram and proteinogram of the rabbit; Action du fractionnement de 100 R de rayons X sur l'hemogramme et le proteinogramme du lapin

Energy Technology Data Exchange (ETDEWEB)

Alix, D [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1965-07-01

Eighty one rabbits 'Fauve de Bourgogne' received total exposure of 100 R X-rays, by fractional doses of 5 - 10 - 25 - 50 and 100 R per irradiation. Blood count and protein dosage were practised before, during and after the irradiation session. Seventeen controls were subjected to the same investigations with the same periodicity. Statistical analysis of experimental data permitted to infer from the exposures of X-rays, that irradiation induces drop in leucocytes count, resulting from a drop in mononuclear cells. On the other hand, it is more difficult to clear up its action on the protein profiles, because the observed changes were likewise observed on the controls, although at a lesser degree. Nevertheless, the authors think that it may be concluded to an increase of albumin concentration and to some instability of the gamma-globulins concentration. Then they discuss the possible mechanisms of X-rays effects: either perturbations of protean metabolism following loss of mitotic activity or histolysis. (author) [French] 81 lapins de race fauve de Bourgogne ont ete irradies in toto aux rayons X a la dose totale de 100 R aux fractionnements suivants: 5 - 10 - 25 - 50 et 100 R par seance. Des examens hematologiques et des dosages de proteines ont ete pratiques avant et apres irradiation et au cours de la periode d'irradiation; 17 temoins non irradies ont ete soumis aux memes examens avec la meme frequence. L'irradiation provoque une chute des leucocytes totaux, due surtout a la chute des elements mononuclees. Par contre, on ne peut pas mettre en evidenre une action propre a l'irradiation sur la formule proteique car les variations rencontrees s'observent egalement dans le lot temoin non irradie. (auteur)

Magnet design with 100-kA HTS STARS conductors for the helical fusion reactor

Science.gov (United States)

Yanagi, N.; Terazaki, Y.; Ito, S.; Tamura, H.; Hamaguchi, S.; Mito, T.; Hashizume, H.; Sagara, A.

2016-12-01

The high-temperature superconducting (HTS) option is employed for the conceptual design of the LHD-type helical fusion reactor FFHR-d1. The 100-kA-class STARS (Stacked Tapes Assembled in Rigid Structure) conductor is used for the magnet system including the continuously wound helical coils. Protection of the magnet system in case of a quench is a crucial issue and the hot-spot temperature during an emergency discharge is estimated based on the zero-dimensional and one-dimensional analyses. The number of division of the coil winding package is examined to limit the voltage generation. For cooling the HTS magnet, helium gas flow is considered and its feasibility is examined by simple analysis as a first step.

Kir6.2 activation by sulfonylurea receptors: a different mechanism of action for SUR1 and SUR2A subunits via the same residues

Science.gov (United States)

Principalli, Maria A; Dupuis, Julien P; Moreau, Christophe J; Vivaudou, Michel; Revilloud, Jean

2015-01-01

ATP-sensitive potassium channels (K-ATP channels) play a key role in adjusting the membrane potential to the metabolic state of cells. They result from the unique combination of two proteins: the sulfonylurea receptor (SUR), an ATP-binding cassette (ABC) protein, and the inward rectifier K+ channel Kir6.2. Both subunits associate to form a heterooctamer (4 SUR/4 Kir6.2). SUR modulates channel gating in response to the binding of nucleotides or drugs and Kir6.2 conducts potassium ions. The activity of K-ATP channels varies with their localization. In pancreatic β-cells, SUR1/Kir6.2 channels are partly active at rest while in cardiomyocytes SUR2A/Kir6.2 channels are mostly closed. This divergence of function could be related to differences in the interaction of SUR1 and SUR2A with Kir6.2. Three residues (E1305, I1310, L1313) located in the linker region between transmembrane domain 2 and nucleotide-binding domain 2 of SUR2A were previously found to be involved in the activation pathway linking binding of openers onto SUR2A and channel opening. To determine the role of the equivalent residues in the SUR1 isoform, we designed chimeras between SUR1 and the ABC transporter multidrug resistance-associated protein 1 (MRP1), and used patch clamp recordings on Xenopus oocytes to assess the functionality of SUR1/MRP1 chimeric K-ATP channels. Our results reveal that the same residues in SUR1 and SUR2A are involved in the functional association with Kir6.2, but they display unexpected side-chain specificities which could account for the contrasted properties of pancreatic and cardiac K-ATP channels. PMID:26416970

Investigation of structural and electronic properties of epitaxial graphene on 3C–SiC(100/Si(100 substrates

Directory of Open Access Journals (Sweden)

Gogneau N

2014-09-01

Full Text Available Noelle Gogneau,1 Amira Ben Gouider Trabelsi,2 Mathieu G Silly,3 Mohamed Ridene,1 Marc Portail,4 Adrien Michon,4 Mehrezi Oueslati,2 Rachid Belkhou,3 Fausto Sirotti,3 Abdelkarim Ouerghi1 1Laboratoire de Photonique et de Nanostructures, Centre National de la Recherche Scientifique, Marcoussis, France; 2Unité des Nanomatériaux et Photonique, Faculté des Sciences de Tunis, Université de Tunis El Manar Campus Universitaire, Tunis, Tunisia; 3Synchrotron-SOLEIL, Saint-Aubin, BP48, F91192 Gif sur Yvette Cedex, France; 4Centre de Recherche sur l'HétéroEpitaxie et Ses Application, Centre National de la Recherche Scientifique, Valbonne, France Abstract: Graphene has been intensively studied in recent years in order to take advantage of its unique properties. Its synthesis on SiC substrates by solid-state graphitization appears a suitable option for graphene-based electronics. However, before developing devices based on epitaxial graphene, it is desirable to understand and finely control the synthesis of material with the most promising properties. To achieve these prerequisites, many studies are being conducted on various SiC substrates. Here, we review 3C–SiC(100 epilayers grown by chemical vapor deposition on Si(100 substrates for producing graphene by solid state graphitization under ultrahigh-vacuum conditions. Based on various characterization techniques, the structural and electrical properties of epitaxial graphene layer grown on 3C–SiC(100/Si(100 are discussed. We establish that epitaxial graphene presents properties similar to those obtained using hexagonal SiC substrates, with the advantage of being compatible with current Si-processing technology. Keywords: epitaxial graphene, electronic properties, structural properties, silicon carbide 

Demonstration of a 100-kWth high-temperature solar thermochemical reactor pilot plant for ZnO dissociation

Science.gov (United States)

Koepf, E.; Villasmil, W.; Meier, A.

2016-05-01

Solar thermochemical H2O and CO2 splitting is a viable pathway towards sustainable and large-scale production of synthetic fuels. A reactor pilot plant for the solar-driven thermal dissociation of ZnO into metallic Zn has been successfully developed at the Paul Scherrer Institute (PSI). Promising experimental results from the 100-kWth ZnO pilot plant were obtained in 2014 during two prolonged experimental campaigns in a high flux solar simulator at PSI and a 1-MW solar furnace in Odeillo, France. Between March and June the pilot plant was mounted in the solar simulator and in-situ flow-visualization experiments were conducted in order to prevent particle-laden fluid flows near the window from attenuating transparency by blocking incoming radiation. Window flow patterns were successfully characterized, and it was demonstrated that particle transport could be controlled and suppressed completely. These results enabled the successful operation of the reactor between August and October when on-sun experiments were conducted in the solar furnace in order to demonstrate the pilot plant technology and characterize its performance. The reactor was operated for over 97 hours at temperatures as high as 2064 K; over 28 kg of ZnO was dissociated at reaction rates as high as 28 g/min.

Solvent refined coal reactor quench system

Science.gov (United States)

Thorogood, Robert M.

1983-01-01

There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

Conceptual site models for groundwater contamination at 100-BC-5, 100-KR-4, 100-HR-3, and 100-FR-3 operable units

International Nuclear Information System (INIS)

Peterson, R.E.; Raidl, R.F.; Denslow, C.W.

1996-09-01

This document presents technical information on groundwater contamination in the 100-BC-5, 100-KR-4, 100-HR-3, and 100-FR-3 Operable Units on the Hanford Site in Richland, Washington. These operable units are defined for groundwater that underlies the retired plutonium production reactors and their associated support facilities. This technical information supports conceptual site models (CSM) for each operable unit. The goal in maintaining a CSM is to ensure that a reasonable understanding of contamination issues in each groundwater operable unit is available for selecting a final remediation alternative and for developing a record of decision. CSMs are developed for hazardous waste sites to help evaluate potential risks to human health and the environment from exposure to contamination

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Design and application of the HTR-100 industrial nuclear power plant

International Nuclear Information System (INIS)

Brandes, S.; Kohl, W.

1988-01-01

The small HTR-100 high temperature reactor combines the reactor concept of the AVR reactor, which has been proven for 20 years, with the latest component technology of the THTR power plant which has been in operation since 1985. The nuclear heat supply system is conceived so as to be applicable for the generation of electric power, district heat and process steam according to the customer's demand. The HTR-100 reactor has a thermal power of 258 MW and offers steam parameters of 190 bar/530 0 C. To cover a higher power demand HTR-100 reactors can be combined forming a larger power plant. Economic analyses have shown competitiveness with fossil power plants. (orig.)

Lunar electric power systems utilizing the SP-100 reactor coupled to dynamic conversion systems. Final report

International Nuclear Information System (INIS)

Harty, R.B.; Durand, R.E.

1993-03-01

An integration study was performed by Rocketdyne under contract to NASA-LeRC. The study was concerned with coupling an SP-0100 reactor to either a Brayton or Stirling power conversion system. The application was for a surface power system to supply power requirements to a lunar base. A power level of 550 kWe was selected based on the NASA Space Exploration Initiative 90-day study. Reliability studies were initially performed to determine optimum power conversion redundancy. This study resulted in selecting three operating engines and one stand-by unit. Integration design studies indicated that either the Brayton or Stirling power conversion systems could be integrated with the PS-100 reactor. The Stirling system had an integration advantage because of smaller piping size and fewer components. The Stirling engine, however, is more complex and heavier than the Brayton rotating unit, which tends to off-set the Stirling integration advantage. From a performance consideration, the Brayton had a 9 percent mass advantage, and the Stirling had a 50 percent radiator advantage

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

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

Power plant system assessment. Final report. SP-100 Program

International Nuclear Information System (INIS)

Anderson, R.V.; Atkins, D.F.; Bost, D.S.

1983-01-01

The purpose of this assessment was to provide system-level insights into 100-kWe-class space reactor electric systems. Using these insights, Rockwell was to select and perform conceptual design studies on a ''most attractive'' system that met the preliminary design goals and requirements of the SP-100 Program. About 4 of the 6 months were used in the selection process. The remaining 2 months were used for the system conceptual design studies. Rockwell completed these studies at the end of FY 1983. This report summarizes the results of the power plant system assessment and describes our choice for the most attractive system - the Rockwell SR-100G System (Space Reactor, 100 kWe, Growth) - a lithium-cooled UN-fueled fast reactor/Brayton turboelectric converter system

Calibration of RB reactor power; Kalibrisanje snage reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Sotic, O; Markovic, H; Ninkovic, M; Strugar, P; Dimitrijevic, Z; Takac, S; Stefanovic, D; Kocic, A; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

1976-09-15

The first and only calibration of RB reactor power was done in 1962, and the obtained calibration ratio was used irrespective of the lattice pitch and core configuration. Since the RB reactor is being prepared for operation at higher power levels it was indispensable to reexamine the calibration ratio, estimate its dependence on the lattice pitch, critical level of heavy water and thickness of the side reflector. It was necessary to verify the reliability of control and dosimetry instruments, and establish neutron and gamma dose dependence on reactor power. Two series of experiments were done in June 1976. First series was devoted to tests of control and dosimetry instrumentation and measurements of radiation in the RB reactor building dependent on reactor power. Second series covered measurement of thermal and epithermal neuron fluxes in the reactor core and calculation of reactor power. Four different reactor cores were chosen for these experiments. Reactor pitches were 8, 8{radical}2, and 16 cm with 40, 52 and 82 fuel channels containing 2% enriched fuel. Obtained results and analysis of these results are presented in this document with conclusions related to reactor safe operation.

Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

International Nuclear Information System (INIS)

Su'ud, Zaki

2008-01-01

In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

Steady State Simulation of Two-Gas Phase Fluidized Bed Reactors in Series for Producing Linear Low Density Polyethylene

Directory of Open Access Journals (Sweden)

Ali Farhangiyan Kashani

2012-12-01

Full Text Available A linear low density polyethylene (LLDPE production process, including two- fuidized bed reactors in series (FBRS and other process equipment, was completely simulated by Aspen Polymer Plus software. Fluidized bed reactors were considered as continuous stirred tank reactors (CSTR consisted of polymer and gas phases. POLY-SRK and NRTL-RK equations of state were used to describe polymer and non-polymer streams, respectively. In this simulation, a kinetic model, based on a double active site heterogeneous Ziegler-Natta catalyst was used for simulation of LLDPE process consisting of two FBRS. Simulator using this model has the capability to  predict a number of  principal characteristics of LLDPE such as melt fow index (MFI, density, polydispersity index, numerical and weight average molecular weights (Mn,Mw and copolymer molar fraction (SFRAC. The results of the simulation were compared with industrial plant data and a good agreement was observed between the predicted model and plant data. The simulation results show the relative error of about 0.59% for prediction of polymer mass fow and 2.67% and 0.04% for prediction of product MFI and density, respectively.

Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function

Research on anhydrous hydrazine synthesis; Recherche sur la synthese de l'hydrazine anhydre

Energy Technology Data Exchange (ETDEWEB)

Gaussens, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-03-15

The first part of this work concerns the radiolysis of pure liquid ammonia. The fundamental importance of the dose rate and of the dose on the yield of radiolytic products has been demonstrated. By using a capture solute at concentrations of between 10{sup -3} and 1.2 mole s/litre, it has been possible to determine the yields of radicals and of molecules in the irradiated pure ammonia. During later work, it was possible to determine, by systematically varying the physico-chemical parameters, the most favorable conditions for carrying out the radiosynthesis; the maximum radiochemical yield of the hydrazine obtained has a value: G (N{sub 2}H{sub 4}) = 2.2/100 eV. An analysis of the molecular yields in the presence of deuterated solutes makes it possible to explain partially the role of the capture species. A project is also described for an installation producing hydrazine continuously; it is followed by an economic study of the process. From this work it appears that the yields of hydrazine obtained justify an industrial application, especially if strong radiation sources are available, for example nuclear reactors. (author) [French] La premiere partie de l'etude a porte sur la radiolyse de l'ammoniac liquide pur. Le role fondamental du debit de dose et de la dose sur les rendements des produits de radiolyse a ete mise en evidence. L'emploi de solute capteur dont les concentrations sont comprises entre 10{sup -3} et 1,2 mole/litre, a permis de determiner la valeur des rendements radicalaires et moleculaires dans l'ammoniac pur irradie. Au cours d'une etape ulterieure, une variation systematique des parametres physico-chimiques a permis de determiner les conditions les plus favorables a la radiosynthese le rendement radiochimique maximum de l'hydrazine obtenu a pour valeur: G(N{sub 2}H{sub 4}) 2,2/100 eV. L'analyse des rendements moleculaires en presence de solutes deuteres nous permet de rendre compte partiellement du role des capteurs. Un projet d'une installation

Principle of natural and artificial radioactive series equivalency

International Nuclear Information System (INIS)

Vasilyeva, A.N.; Starkov, O.V.

2001-01-01

In the present paper one approach used under development of radioactive waste management conception is under consideration. This approach is based on the principle of natural and artificial radioactive series radiotoxic equivalency. The radioactivity of natural and artificial radioactive series has been calculated for 10 9 - years period. The toxicity evaluation for natural and artificial series has also been made. The correlation between natural radioactive series and their predecessors - actinides produced in thermal and fast reactors - has been considered. It has been shown that systematized reactor series data had great scientific significance and the principle of differential calculation of radiotoxicity was necessary to realize long-lived radioactive waste and uranium and thorium ore radiotoxicity equivalency conception. The calculations show that the execution of equivalency principle is possible for uranium series (4n+2, 4n+1). It is a problem for thorium. series. This principle is impracticable for neptunium series. (author)

Low power modular power generating reactors or Small Modular Reactors (SMR)

International Nuclear Information System (INIS)

Chenais, Jacques

2016-01-01

Electronuclear reactors were small reactors at the beginning, and then tend to be always bigger and more powerful, but since some recent times, several countries specialized in reactor design and fabrication (USA, Russia, China, and South Korea) have been developing Small Modular Reactors (SMR) of less than 300 MW. As France has already produced feasibility studies and is about to launch a SMR development programme, the author comments some specific aspects of this new architecture of reactors, characterises the targeted markets, gives an overview of the various more or less advanced existing concepts: a floating barge in Russia, the SMART 100 MW project in South Korea, several concepts in the USA (the mPower 125 MW, the NuScale 45 MW, the Westinghouse 225 MW, and the HI-SMUR 160 MW projects), the ACP 100 MW in China, the CAREM 27 MW in Argentina. French projects developed by the CEA, EDF, Areva and DCNS are then presented

Agir sur la langue pour agir sur le monde : Micropolitiques linguistiques autogérées du genre dans les brochures libertaires

OpenAIRE

Abbou, Julie

2015-01-01

La dimension agissante de la langue sur le monde entraîne avec elle, comme un appel d'air, l'action sur la langue elle-même, et l'action sur la langue n'est jamais loin de se faire action politique. Les politiques linguistiques de féminisation en sont un exemple, où l'action politique se concentre sur la catégorisation du genre. Dans de nombreux cas, la politique linguistique se fait planification linguistique, cherchant de nouvelles normes. Dans de nombreux cas, mais pas toujours. Il existe ...

Materials for fusion reactors

International Nuclear Information System (INIS)

Ehrlich, K.; Kaletta, D.

1978-03-01

The following report describes five papers which were given during the IMF seminar series summer 1977. The purpose of this series was to discuss especially the irradiation behaviour of materials intended for the first wall of future fusion reactors. The first paper deals with the basic understanding of plasma physics relating to the fusion reactor and presents the current state of art of fusion technology. The next two talks discuss the metals intended for the first wall and structural components of a fusion reactor. Since 14 MeV neutrons play an important part in the process of irradiation damage their role is discussed in detail. The question which machines are presently available to simulate irradiation damage under conditions similar to the ones found in a fusion reactor are investigated in the fourth talk which also presents the limitations of the different methods of simulation. In this context also discussed is the importance future intensive neutron sources and materials test reactors will have for this problem area. The closing paper has as a theme the review of the present status of research of metallic and non-metallic materials in view of the quite different requirements for different fusion systems; a closing topic is the world supply on rare materials required for fusion reactors. (orig) [de

CARACTERIZACIÓN DEL CONSUMO DE HORTALIZAS EN LAS FAMILIAS DEL SUR-SUR DE COSTA RICA

Directory of Open Access Journals (Sweden)

Alexis Villalobos-Monge

2013-01-01

Full Text Available El objetivo de este trabajo fue determinar los diferentes aspectos cuantitativos y cualitativos que explican la cultura de consumo actual de hortalizas en las familias residentes en la zona sur-sur de Costa Rica. Este trabajo expone resultados relacionados a la caracterización realizada en el 2011 en familias residentes en la denominada zona sur-sur de Costa Rica sobre la cultura de consumo de productos hortícolas. Se aplicaron cuestionarios a una muestra estratificada de familias; lo que permitió establecer la valoración de los precios de mercado, por parte de las familias, para consumir estos alimentos, donde se determinó un valor máximo de US$1,74 millones por semana. Los principales rubros de consumo de acuerdo al valor pagado fueron el tomate, la papa, el plátano, la cebolla y el brócoli, para citar los cinco principales. También fue posible establecer valoraciones cualitativas sobre el consumo de estos productos; por ejemplo se determinó que para el 71,2% de las familias, la frescura representa la característica de mayor valor, además, el principal sitio donde los núcleos familiares realizan las compras de estos alimentos corresponde a supermercados (38,4% de los casos.

Reactor technology

International Nuclear Information System (INIS)

Erdoes, P.

1977-01-01

This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

Pilot program: NRC severe reactor accident incident response training manual: Severe reactor accident overview

International Nuclear Information System (INIS)

McKenna, T.J.; Martin, J.A.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Severe Reactor Accident Overview is the second in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assesment. Each volume serves, respectively, as the text for a course of instruction in a series of courses. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

Nuclear reactor safety systems

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1980-01-01

A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

The reactor Cabri; La pile cabri

Energy Technology Data Exchange (ETDEWEB)

Ailloud, J; Millot, J P [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [French] II devenait necessaire de construire en France une pile qui permette d'etudier les conditions de fonctionnement des installations futures, de choisir, tester et mettre au point les dispositifs de securite a adopter. On a choisi une pile a eau, type de pile qui correspond aux constructions les plus nouvelles du CEA en matiere de piles laboratoire ou d'universite; il importe en effet de pouvoir evaluer les risques presentes et d'etudier les possibilites d'augmentation de puissance constamment demandees par les utilisateurs: il est particulierement interessant d'eclaircir les phenomenes d'oscillation de puissance et les risques de calefaction (burn out). Les programmes de travaux sur CABRI seront harmonises avec les travaux effectues sur les Spert americains de meme type; lors de sa construction des contacts fructueux ont ete etablis avec les specialistes americains qui ont defini les premiers de ces reacteurs. La communication donne une description sommaire de la pile et decrit le

Oradour-sur-Glane: On the emergence of a glocal site of memory in France

Directory of Open Access Journals (Sweden)

Léger, Eva

2014-12-01

Full Text Available Oradour-sur-Glane (France is the memorial site of a massacre perpetrated by the Second SS Panzer Division Das Reich on June 10th, 1944. It preserves the memory of the 642 people slaughtered there, including 18 Spanish refugees. In 1945, the French State, led by General de Gaulle, decided to preserve the ruins of Oradour-sur- Glane. Since then, a series of commemorative processes have ensued at the site, corresponding to different temporalities. Over time, this site of national memory has been linked both with European memory discourse and with the private memory of exiled Spaniards and, consequently, with the memory of the Spanish Civil War and Francoism. In this article, I analyze the different appropriations and interpretations of the site, focusing in particular on the memory of the exiles. To do so, I will look into the initiatives undertaken between 2008 and 2014, considering both the institutions and the associations related to Oradour-sur-Glane.Oradour-sur-Glane (Francia es el lugar de memoria de una masacre perpetrada por la segunda división blindada S.S. Das Reich el 10 de junio de 1944 que conserva el recuerdo de sus 642 supliciados, entre los cuales 18 españoles refugiados. En 1945, el Estado francés encabezado por el General de Gaulle decidió preservar las ruinas de Oradour-sur-Glane. Desde entonces, diferentes procesos conmemorativos, correspondientes a temporalidades destacadas, se han sucedido en el sitio. Con el tiempo, este lugar de memoria nacional, se ha vinculado tanto con un discurso de memoria europeo como con la memoria particular de los españoles exiliados y a través de ello con la memoria de la Guerra Civil española y del franquismo. En este estudio analizaré las diferentes apropiaciones e interpretaciones de este lugar, enfocándome especialmente en la memoria de los exiliados. Para esto, contemplaré las iniciativas que se han desarrollado entre los años 2008 y 2014, atendiendo tanto a las instituciones como a

Status of advanced technologies for CANDU reactors

International Nuclear Information System (INIS)

Lipsett, J.J.

1989-01-01

The future development of the CANDU reactor is a continuation of a successful series of reactors, the most recent of which are nine CANDU 6 Mk 1* units and four Darlington units. There are three projects underway that continue the development of the CANDU reactor. These new design projects flow from the original reactor designs and are a natural progression of the CANDU 6 Mk 1, two units of which are operating successfully in Canada, one each in Argentina and Korea, with five more being built in Rumania. These new design projects are known as: CANDU 6 Mk 2, an improved version of CANDU 6 Mk 1; CANDU 3, a small, advanced version of the CANDU 6 Mk 1; CANDU 6 Mk 3, a series of advanced CANDU reactors. A short description of modified versions of CANDU reactors is given in this paper. 5 figs

Chromosomes and irradiation: in vitro study of the action of X-rays on human lymphocytes; Chromosomes et radiations: etude in vitro de l'action des rayons X sur les lymphocytes humains

Energy Technology Data Exchange (ETDEWEB)

Mouriquand, C; Patet, J; Gilly, C; Wolff, C

1966-07-01

Radioinduced chromosomal aberrations were studied in vitro on leukocytes of human peripheral blood after x irradiation at 25, 50, 100, 200, and 300 R. The numeric and structural anomalies were examined on 600 karyotypes. The relationship between these disorders and the dose delivered to the blood are discussed. An explanation on their mechanism of formation is tentatively given. (authors) [French] L'etude in vitro des anomalies chromosomiques radioinduites a ete pratiquee sur des leucocytes de sang peripherique preleve chez 4 sujets et irradie aux doses de 25, 50, 100, 200, 300 R. Les aberrations numeriques et structurales ont ete examinees sur 600 caryotypes. Les rapports entre ces anomalies et les doses appliquees sont etudies. Une hypothese sur leur mecanisme de formation est avancee. (auteurs)

Fusion reactor materials

International Nuclear Information System (INIS)

Rowcliffe, A.F.; Burn, G.L.; Knee', S.S.; Dowker, C.L.

1994-02-01

This is the fifteenth in a series of semiannual technical progress reports on fusion reactor materials. This report combines research and development activities which were previously reported separately in the following progress reports: Alloy Development for Irradiation Performance; Damage Analysis and Fundamental Studies; Special purpose Materials. These activities are concerned principally with the effects of the neutronic and chemical environment on the properties and performance of reactor materials; together they form one element of the overall materials programs being conducted in support of the Magnetic Fusion Energy Program of the U.S. Department of Energy. The Fusion Reactor Materials Program is a national effort involving several national laboratories, universities, and industries. The purpose of this series of reports is to provide a working technical record for the use of the program participants, and to provide a means of communicating the efforts of materials scientists to the rest of the fusion community, both nationally and worldwide

Research on anhydrous hydrazine synthesis; Recherche sur la synthese de l'hydrazine anhydre

Energy Technology Data Exchange (ETDEWEB)

Gaussens, G. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-03-15

The first part of this work concerns the radiolysis of pure liquid ammonia. The fundamental importance of the dose rate and of the dose on the yield of radiolytic products has been demonstrated. By using a capture solute at concentrations of between 10{sup -3} and 1.2 mole s/litre, it has been possible to determine the yields of radicals and of molecules in the irradiated pure ammonia. During later work, it was possible to determine, by systematically varying the physico-chemical parameters, the most favorable conditions for carrying out the radiosynthesis; the maximum radiochemical yield of the hydrazine obtained has a value: G (N{sub 2}H{sub 4}) = 2.2/100 eV. An analysis of the molecular yields in the presence of deuterated solutes makes it possible to explain partially the role of the capture species. A project is also described for an installation producing hydrazine continuously; it is followed by an economic study of the process. From this work it appears that the yields of hydrazine obtained justify an industrial application, especially if strong radiation sources are available, for example nuclear reactors. (author) [French] La premiere partie de l'etude a porte sur la radiolyse de l'ammoniac liquide pur. Le role fondamental du debit de dose et de la dose sur les rendements des produits de radiolyse a ete mise en evidence. L'emploi de solute capteur dont les concentrations sont comprises entre 10{sup -3} et 1,2 mole/litre, a permis de determiner la valeur des rendements radicalaires et moleculaires dans l'ammoniac pur irradie. Au cours d'une etape ulterieure, une variation systematique des parametres physico-chimiques a permis de determiner les conditions les plus favorables a la radiosynthese le rendement radiochimique maximum de l'hydrazine obtenu a pour valeur: G(N{sub 2}H{sub 4}) 2,2/100 eV. L'analyse des rendements moleculaires en presence de solutes deuteres nous permet de rendre compte partiellement du role des

Testing and analyses of a high temperature duct for gas-cooled reactors

International Nuclear Information System (INIS)

Black, W.E.; Roberge, A.; Felten, P.; Bastien, R.

1979-01-01

A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. Pre and post test metallurgical analyses were conducted on the Hastelloy-X structures and reference specimens. The results gave evidence of aging in the form of noticeable changes in room temperature tensile and reduction in area parameters. The Hastelloy-X welds exhibited greater changes in properties due to thermal aging. The antifriction coating (Cr 3 C 2 ) performed well without spallation or excessive wear. (orig.)

Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

2003-01-01

The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

Thermal and flow design of helium-cooled reactors

International Nuclear Information System (INIS)

Melese, G.; Katz, R.

1984-01-01

This book continues the American Nuclear Society's series of monographs on nuclear science and technology. Chapters of the book include information on the first-generation gas-cooled reactors; HTGR reactor developments; reactor core heat transfer; mechanical problems related to the primary coolant circuit; HTGR design bases; core thermal design; gas turbines; process heat HTGR reactors; GCFR reactor thermal hydraulics; and gas cooling of fusion reactors

Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

International Nuclear Information System (INIS)

Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

1990-11-01

The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs

Review of selected 100-N waste sites related to N-Springs remediation projects

International Nuclear Information System (INIS)

DeFord, D.H.; Carpenter, R.W.

1996-01-01

This document has been prepared in support of the environmental restoration program at the US Department of Energy's Hanford Site near Richland, Washington, by the Bechtel Hanford, Inc. Facility and Waste Site Research Office. It provides historical information that documents and characterizes selected waste sites that are related to the N-Springs remediation projects. The N-Springs are a series of small, inconspicuous groundwater seepage springs located along the Columbia River shoreline near the 100-N Reactor. The spring site is hydrologically down-gradient from several 100-N Area liquid waste sites that are believed to have been the source(s) of the effluents being discharged by the springs. This report documents and characterizes these waste sites, including the 116-N-1 Crib and Trench, 116-N-3 Crib and Trench, unplanned releases, septic tariks, and a backwash pond

SP-100 converter multicouple thermoelectric cell

International Nuclear Information System (INIS)

Kull, R.A.; Terrill, W.R.

1990-01-01

The General Electric Company is under contract to DOE to design, fabricate, and test an SP-100 Ground Engineering System. This paper provides a description of the SP-100 space reactor power system configuration, and a more detailed description of the power conversion subsystem (PCSS) and the key building block of the power converter, the thermoelectric cell. The functions of the various elements of the PCSS and the cells are also presented. These cells convert the thermal energy from the reactor into electrical power at the desired voltage while being conductively coupled to the hot and cold side heat exchangers to maximize the power output and system specific power

ACADEMIC TRAINING LECTURE SERIES: Introduction to General Relativity and Black Holes

CERN Multimedia

2003-01-01

10, 11, 12, 13, 14 February ACADEMIC TRAINING LECTURE SERIES from 11.00 to 12.00 hrs - Auditorium, bldg. 500 Introduction to General Relativity and Black Holes by T.DAMOUR, IHES, Bures-sur-Yvette, F - Physical motivation behind Einstein's theory. - Mathematical formalism of General Relativity. - Experimental confirmations of Einstein's theory. - Introduction to Black Holes physics.

Applications of Research Reactors

International Nuclear Information System (INIS)

2014-01-01

One of the IAEA's statutory objectives is to 'seek to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world.' One way this objective is achieved is through the publication of a range of technical series. Two of these are the IAEA Nuclear Energy Series and the IAEA Safety Standards Series. According to Article III.A.6 of the IAEA Statute, the safety standards establish 'standards of safety for protection of health and minimization of danger to life and property'. The safety standards include the Safety Fundamentals, Safety Requirements and Safety Guides. These standards are written primarily in a regulatory style, and are binding on the IAEA for its own programmes. The principal users are the regulatory bodies in Member States and other national authorities. The IAEA Nuclear Energy Series comprises reports designed to encourage and assist R and D on, and application of, nuclear energy for peaceful uses. This includes practical examples to be used by owners and operators of utilities in Member States, implementing organizations, academia, and government officials, among others. This information is presented in guides, reports on technology status and advances, and best practices for peaceful uses of nuclear energy based on inputs from international experts. The IAEA Nuclear Energy Series complements the IAEA Safety Standards Series. The purpose of the earlier publication, The Application of Research Reactors, IAEA-TECDOC-1234, was to present descriptions of the typical forms of research reactor use. The necessary criteria to enable an application to be performed were outlined for each one, and, in many cases, the minimum as well as the desirable requirements were given. This revision of the publication over a decade later maintains the original purpose and now specifically takes into account the changes in service requirements demanded by the relevant stakeholders. In particular, the significant improvements in

El (los Sur buscando al Sur. Una construcción entre estereotipos y realidad.

Directory of Open Access Journals (Sweden)

Mabel Franzone

2010-11-01

Full Text Available Todas las lecturas que podamos hacer del Sur, de sus propias representaciones, de aquellas vehiculizadas por el Norte, de sus creaciones, de sus reacciones, de su búsqueda constante de identidad, son lecturas de múltiples situaciones, que llaman a cruces insoslayables entre pueblos, entre distintas etnias, entre disciplinas, entre sentimientos que van desde el rechazo hasta la atracción casi magnética entre los dos hemisferios. Nuestra intención primera era la de reunir de manera aproximativa las problemáticas actuales de los Sur, ligadas a lo imaginario.

Bioconversion reactor

Science.gov (United States)

McCarty, Perry L.; Bachmann, Andre

1992-01-01

A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

Summary report for 1990 inservice inspection (ISI) of SRS 100-L reactor tank

International Nuclear Information System (INIS)

Morrison, J.M.; Loibl, M.W.

1991-01-01

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in about 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. The primary objective of this inspection was to determine if the accessible welds and selected portions of base metal in the L Reactor tank wall contain any indications of IGSCC. This inspection included areas in and beyond the weld HAZ, extending out as far as two to three inches from the centerline of the welds, plus selected areas of base metal at the intersection of the main tank vertical and mid-girth welds. No evidence of such degradation was found in any of the areas examined. Further, additional inspections were conducted of areas that had been damaged and repaired during original fabrication, and on a sample of areas containing linear indications observed during the 1986 visual inspection of the tank. No evidence of IGSCC or other service induced degradation was detected in these areas, either. The inspection was initially planned to cover a minimum of 60% of the accessible welds, plus repair areas and a sample of the indications from the 1986 visual inspection. Direction was received from DOE while the inspection was in progress to expand the scope to cover 100% of the accessible weld areas, and the plan was adjusted accordingly. Initial setup of the tank, which prior to inspection contained Mark 60B target assemblies and nearly a full charge of Mark 22 fuel assemblies, began on October 15, 1990. The inspection was completed on April 12, 1991

Summation of series

CERN Document Server

Jolley, LB W

2004-01-01

Over 1,100 common series, all grouped for easy reference. Arranged by category, these series include arithmetical and geometrical progressions, powers and products of natural numbers, figurate and polygonal numbers, inverse natural numbers, exponential and logarithmic series, binomials, simple inverse products, factorials, trigonometrical and hyperbolic expansions, and additional series. 1961 edition.

Characterization of a continuous agitated cell reactor for oxygen dependent biocatalysis.

Science.gov (United States)

Toftgaard Pedersen, Asbjørn; de Carvalho, Teresa Melo; Sutherland, Euan; Rehn, Gustav; Ashe, Robert; Woodley, John M

2017-06-01

Biocatalytic oxidation reactions employing molecular oxygen as the electron acceptor are difficult to conduct in a continuous flow reactor because of the requirement for high oxygen transfer rates. In this paper, the oxidation of glucose to glucono-1,5-lactone by glucose oxidase was used as a model reaction to study a novel continuous agitated cell reactor (ACR). The ACR consists of ten cells interconnected by small channels. An agitator is placed in each cell, which mixes the content of the cell when the reactor body is shaken by lateral movement. Based on tracer experiments, a hydrodynamic model for the ACR was developed. The model consisted of ten tanks-in-series with back-mixing occurring within and between each cell. The back-mixing was a necessary addition to the model in order to explain the observed phenomenon that the ACR behaved as two continuous stirred tank reactors (CSTRs) at low flow rates, while it at high flow rates behaved as the expected ten CSTRs in series. The performance of the ACR was evaluated by comparing the steady state conversion at varying residence times with the conversion observed in a stirred batch reactor of comparable size. It was found that the ACR could more than double the overall reaction rate, which was solely due to an increased oxygen transfer rate in the ACR caused by the intense mixing as a result of the spring agitators. The volumetric oxygen transfer coefficient, k L a, was estimated to be 344 h -1 in the 100 mL ACR, opposed to only 104 h -1 in a batch reactor of comparable working volume. Interestingly, the large deviation from plug flow behavior seen in the tracer experiments was found to have little influence on the conversion in the ACR, since both a plug flow reactor (PFR) model and the backflow cell model described the data sufficiently well. Biotechnol. Bioeng. 2017;114: 1222-1230. © 2017 Wiley Periodicals, Inc. © 2017 Wiley Periodicals, Inc.

Purification by molecular sieve of helium used as inert cover gas in nuclear reactors; Epuration de l'helium de couverture des reacteurs nucleaires par adsorption sur tamis moleculaire

Energy Technology Data Exchange (ETDEWEB)

Rozenberg, J; Kahan, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1967-07-01

A method carried out at fairly low temperatures (between -50 and -80 deg. C) has been studied for the purification of the helium used as cover gas for heavy water in reactors. The use of the 5A molecular sieve has been adopted because of its superiority over other adsorbents in this temperature range. The particular problems connected with adsorption under dynamic conditions have been dealt with separately. The nitrogen adsorption isotherms have been plotted and the heat of adsorption calculated. (authors) [French] Une methode d'epuration, a temperature moderement basse (comprise entre -50 et -80 deg. C) de l'helium servant de couverture inerte a l'eau lourde des reacteurs a ete etudiee. L'emploi au tamis moleculaire 5A a ete retenu pour la superiorite de celui-ci sur d'autres adsorbants dans ce domaine de temperatures. Les problemes particuliers a l'adsorption en regime dynamique ont ete separement traites. Les isothermes d'adsorption d'azote ont ete tracees et la chaleur d'adsorp. tion calculee. (auteurs)

SOLUCIÓN ANALÍTICA PARA OBTENER EL VOLUMEN ÓPTIMO DE UNA SERIE DE REACTORES DE AGITACIÓN CONTINUA DONDE SE EFECTÚA UNA REACCIÓN DE PRIMER ORDEN

Directory of Open Access Journals (Sweden)

Ignacio Elizalde

2013-01-01

Full Text Available An analytical procedure for determining the optimum size of CSTR in series operating under isothermal and isobaric conditions sustaining first order reaction at constant density has been developed. The procedure requires the concentration of reactant at the entrance of the first reactor and at the outlet of the last reactor; it is also required the continuity of reaction rate as function of conversion, due to the later changes from one reactor to another. The optimization method involves the calculation of intermediate concentrations instead of their estimation, as it is done by graphical solution reported previously. Also, the procedure reported in this contribution is valid for any reactor number. Under these circumstances the method predicts that all reactors must have the same size in order to minimize the total volume of the system.

Preliminary studies of vanadium-base alloys intended for use in fabrication of cans for fast reactors; Etudes preliminaires sur les alliages a base de vanadium envisages pour la fabrication de gaines de reacteurs rapides

Energy Technology Data Exchange (ETDEWEB)

Conte, M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

1967-03-15

Preliminary research has been carried out on a series of vanadium-based alloys: V, 0.5 per cent Si; V, 5 per cent Ca; V, 5 per cent Mo; V, 5 per cent Nb; V, 2 per cent Zr; V, 20 per cent Ti; V, 10 per cent Al; V, 10 per cent Sn and v, 10 per cent Ti liable to be used as canning material in fast reactors. The transformation by forging at about 1000 deg. C and rolling between 200 deg. C and room temperature is satisfactory for all types of alloys except V with 10 per cent Sn and V with 10 per cent Al. The mechanical properties deduced from tensile strength tests carried out on alloy samples annealed 1 hour at 1050 deg. C in a vacuum show that, generally speaking, the addition elements lead to an improvement in these properties as compared to those of pure vanadium. After undergoing corrosion tests in a liquid sodium loop purified by a cold trap, the alloys become brittle at room temperature. Only the vanadium containing 20 per cent Ti keeps its plastic properties. These alloys are covered by a layer of vanadium carbide VC. After undergoing treatment in a liquid sodium loop purified by a hot trap, all the alloys keep their good mechanical characteristics. The surface layer with which they are covered is composed of two vanadium carbides VC and {sub {gamma}}VC, and a vanadium sub-oxide VO{sub 0.9}. (author) [French] Des etudes preliminaires ont ete faites sur une serie d'alliages a base de vanadium: V-0,5 pour cent Si, V-5 pour cent Ca, V-5 pour cent Mo, V-5 pour cent Nb, V-2 pour cent Zr, V-20 pour cent Ti, V-10 pour cent Al, V-10 pour cent Sn et V-10 pour cent Ti susceptibles d'etre utilises comme materiau de gainage pour les reacteurs rapides. La transformation par forgeage a 1000 deg. C environ et laminage entre 200 deg. C et la temperature ambiante est satisfaisante pour toutes les nuances d'alliage sauf le V-10 pour cent Sn et le V-10 pour cent Al. Les proprietes mecaniques deduites des essais de traction realises sur des eprouvettes d'alliages recuits 1 heure a

Lead-based Fast Reactor Development Plan and R&D Status in China

International Nuclear Information System (INIS)

Wu Yican

2013-01-01

• Lead-based fast reactors have good potential for waste transmutation, fuel breeding and energy production, which has been selected by CAS as the advanced reactor development emphasis with the support of ADS program and MFE program. Sharing of technologies R&D is possible among GIF/ADS/Fusion. • The concepts and test strategy of series China lead-based fast reactors (CLEAR) have been developed. The preliminary engineering design and safety analysis of CLEAR-I are underway. • Technology R&D on CLEAR with series lead alloy loops and accelerator-based neutron generator have been constructed or under construction. • CLEAR series reactor design and construction have big challenges, widely international cooperation on reactor design and technology R&D is welcome

Hanford Site 100-N Area In Situ Bioremediation of UPR-100-N-17, Deep Petroleum Unplanned Release - 13245

International Nuclear Information System (INIS)

Saueressig, Daniel G.

2013-01-01

In 1965 and 1966, approximately 303 m 3 of Number 2 diesel fuel leaked from a pipeline used to support reactor operations at the Hanford Site's N Reactor. N Reactor was Hanford's longest operating reactor and served as the world's first dual purpose reactor for military and power production needs. The Interim Action Record of Decision for the 100-N Area identified in situ bioremediation as the preferred alternative to remediate the deep vadose zone contaminated by this release. A pilot project supplied oxygen into the vadose zone to stimulate microbial activity in the soil. The project monitored respiration rates as an indicator of active biodegradation. Based on pilot study results, a full-scale system is being constructed and installed to remediate the vadose zone contamination. (authors)

RB reactor noise analysis; Analiza sumova reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Petrovic, M; Velickovic, Lj; Markovic, V; Jovanovic, S [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)

1964-07-01

Statistical fluctuations of reactivity represent reactor noise. Analysis of reactor noise enables determining a series of reactor kinetic parameters. Fluctuations of power was measured by ionization chamber placed next to the tank of the RB reactor. The signal was digitized by an analog-digital converter. After calculation of the mean power, 3000 data obtained by sampling were analysed.

The state of the PIK reactor construction

International Nuclear Information System (INIS)

Konoplev, K.A.

1995-01-01

Principle concepts of the PIK reactor project were stated late in the 60's but its construction was started in 1976. By the year 1986 the initial project was realised by approximately 70% but then, after Chernobyl accident the construction was essentially frozen to adjust the project to the revised nuclear safety regulations. The revised project was approved only in 1990 when the country was on the threshold of serious economic problems. The PIK reactor is a source of neutrons placed in the heavy water reflector. The fuel is uranium-235 (90% enrichment) of total weight 27 kg. Light water is used as moderator and coolant. Design parameters: thermal power is 100 W; thermal neutron flux in the reflector is 1.2x10 15 n/cm 2 s; in the central vertical beam tube is 5x10 15 n/cm 2 s; number of horizontal beam tubes is 10; diameter of beam tubes is 10 cm, with the possibility of replacement with beam tubes up to 25 cm in diameter. The reactor will be equipped with sources of hot, cold, and ultracold neutrons to obtain beams in different intervals of energy spectrum. The low temperature circuit will enable to irradiate samples at helium temperatures. The reactor has three series cooling circuits. Emergency core cooling systems in LOCA are double and in emergency power supply system is triple. The PIK reactor has no single common containment but four separate systems: for pipelines and units of the first circuit, for heavy water reflector, for operating hall, and for experimental beam tubes hall

Thermionic reactors for space nuclear power

Science.gov (United States)

Homeyer, W. G.; Merrill, M. H.; Holland, J. W.; Fisher, C. R.; Allen, D. T.

1985-01-01

Thermionic reactor designs for a variety of space power applications spanning the range from 5 kWe to 3 MWe are described. In all of these reactors, nuclear heat is converted directly to electrical energy in thermionic fuel elements (TFEs). A circulating reactor coolant carries heat from the core of TFEs directly to a heat rejection radiator system. The recent design of a thermionic reactor to meet the SP-100 requirements is emphasized. Design studies of reactors at other power levels show that the same TFE can be used over a broad range in power, and that design modifications can extend the range to many megawatts. The design of the SP-100 TFE is similar to that of TFEs operated successfully in test reactors, but with design improvements to extend the operating lifetime to seven years.

Variabilidad de la abundancia de zooplancton en Bahía Magdalena Baja California Sur, México (1997-2001 Zooplankton abundance variability in Magdalena Bay, Baja California Sur, Mexico (1997-2001

Directory of Open Access Journals (Sweden)

Sergio Hernández-Trujillo

2010-01-01

Full Text Available Se analizaron muestras de zooplancton de 16 campañas oceanógraficas, efectuadas en Bahía Magdalena, Baja California Sur, México, entre agosto de 1997 y marzo de 2001. Se identificó un total de 26 grupos taxonómicos, de los cuales los más abundantes y frecuentes fueron copépodos y quetognatos; en 2000-2001 se observó una tendencia a disminuir entre 10 y 20 el número de grupos de zooplancton. La biomasa zooplanctónica y abundancia de copépodos disminuyeron en el periodo de estudio, en contraste con los quetognatos que tuvieron un ligero aumento. Las fluctuaciones de abundancia de zooplancton no estuvieron relacionadas con la concentración de clorofila-α, a diferencia de los máximos de abundancia de zooplancton, que estuvieron asociados a los cambios de la temperatura superficial del mar. El ciclo estacional de la abundancia del zooplancton en Bahía Magdalena, indicó que en invierno el promedió fue mayor de 65.000 ind 100 m-3 , valor que aumentó en primavera a más de 99.000 ind 100 m-3 , se mantuvo en verano alrededor de 100.000 ind 100 m-3 y en otoño descendió rápidamente a casi 40.000 ind 100 m-3.Zooplankton were studied from 16 oceanographic surveys carried out in Magdalena Bay, Baja California Sur, Mexico, between August 1997 and March 2001. Twenty-six taxonomic groups were identified, the most abundant and frequent of which were copepods and chaetognaths. In 2000-2001, the number of zooplankton groups tended to decrease by 10 to 20. Both zooplankton biomass and copepod abundance declined, unlike chaetognaths, which increased slightly. Fluctuations in zooplankton abundance were independent of the chlorophyll-a concentration, whereas the maximum zooplankton abundances were associated with changes in the sea surface temperature. The seasonal zooplankton abundance cycle in Magdalena Bay indicated that, in winter, the averaged was than 65,000 ind 100 m-3 , a value that increased to more than 99,000 ind 100 m-3 in spring

RMB. The new Brazilian multipurpose research reactor

International Nuclear Information System (INIS)

Perrotta, Jose Augusto; Soares, Adalberto Jose

2015-01-01

Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also presents the

RMB. The new Brazilian multipurpose research reactor

Energy Technology Data Exchange (ETDEWEB)

Perrotta, Jose Augusto; Soares, Adalberto Jose [Comissao Nacional de Energia Nuclear (CNEN) (Brazil)

2015-01-15

Brazil has four research reactors (RR) in operation: IEA-R1, a 5 MW pool type RR; IPR-R1, a 100 kW TRIGA type RR; ARGONAUTA, a 500 W Argonaut type RR, and IPEN/MB-01, a 100 W critical facility. The first three were constructed in the 50's and 60's, for teaching, training, and nuclear research, and for many years they were the basic infrastructure for the Brazilian nuclear developing program. The last, IPEN/MB-01, is the result of a national project developed specifically for qualification of reactor physics codes. Considering the relative low power of Brazilian research reactors, with exception of IEAR1, none of the other reactors are feasible for radioisotope production, and even IEA-R1 has a limited capacity. As a consequence, since long ago, 100% of the Mo-99 needed to attend Brazilian nuclear medicine services has been imported. Because of the high dependence on external supply, the international Moly-99 supply crisis that occurred in 2008/2009 affected significantly Brazilian nuclear medicine services, and as presented in previous IAEA events, in 2010 Brazilian government formalized the decision to build a new research reactor. The new reactor named RMB (Brazilian Multipurpose Reactor) will be a 30 MW open pool type reactor, using low enriched uranium fuel. The facility will be part of a new nuclear research centre, to be built about 100 kilometres from Sao Paulo city, in the southern part of Brazil. The new nuclear research centre will have several facilities, to use thermal and cold neutron beams; to produce radioisotopes; to perform neutron activation analysis; and to perform irradiations tests of materials and fuels of interest for the Brazilian nuclear program. An additional facility will be used to store, for at least 100 years, all the fuel used in the reactor. The paper describes the main characteristics of the new centre, emphasising the research reactor and giving a brief description of the laboratories that will be constructed, It also

Sexualidad, migraciones y fronteras en contextos de integración sur-sur

Directory of Open Access Journals (Sweden)

Martha Cecilia Ruiz

Full Text Available Resumen Este artículo se pregunta sobre los procesos de diferenciación, jerarquización e inclusión/exclusión de poblaciones migrantes en contextos migratorios sur-sur, y sobre el rol que juega la sexualidad en estos procesos. Partiendo de un estudio etnográfico sobre las migrantes peruanas y colombianas en el sector del comercio sexual de Ecuador, se analiza la manera en que la sexualidad se convierte en un sitio privilegiado para re-imaginar las diferencias y jerarquías nacionales en un mundo globalizado e integrado regionalmente, y se explica cómo los regímenes de control sobre las migraciones y la sexualidad femenina se articulan entre sí para restablecer el orden idealizado de la nación.

Effets des différents doses d'agar du milieu de culture sur l'induction ...

African Journals Online (AJOL)

SARAH

31 janv. 2014 ... récipients utilisés étaient des bacs en plastic munis du filtre d'aération sur le couvercle. Il a été garni de 100 ml de milieu de culture ... The containers used were plastic containers marked with the breather filter on the lid. ..... 1991; Brown et Atanassov, 1985 ; Reisch et Bingham,. 1980; Bingham et al., 1975).

The reactor Cabri; La pile cabri

Energy Technology Data Exchange (ETDEWEB)

Ailloud, J.; Millot, J.P. [Commissariat a l' Energie Atomique, Cadarache (France). Centre d' Etudes Nucleaires

1964-07-01

exceptional circumstances... - experimental investigations on power excursions linked with precise initial conditions: the aim of this work is to define the basis for theoretical research, and the limits beyond which the risks of explosion cease to be negligible. The research work will be done so as to enable checking with outside reactor experiments and to continue them in the explosion field. - studies of the behaviour of the reactor control-instrumentation. - experimental investigations related with transient operation with initial short life (study of boiling, temperature measurements, vacuum pressure and fraction...) with the aim of defining the hypotheses of a theory on swimming-pool reactor kinetics related to heat transfer phenomena, - investigations of the behaviour of fuels in reactors (these experiments are planned to be carried out in loops) Preliminary experimental results. CABRI went critical on the 21 December 1963. The first transient experiments are expected for March 1964. (authors) [French] II devenait necessaire de construire en France une pile qui permette d'etudier les conditions de fonctionnement des installations futures, de choisir, tester et mettre au point les dispositifs de securite a adopter. On a choisi une pile a eau, type de pile qui correspond aux constructions les plus nouvelles du CEA en matiere de piles laboratoire ou d'universite; il importe en effet de pouvoir evaluer les risques presentes et d'etudier les possibilites d'augmentation de puissance constamment demandees par les utilisateurs: il est particulierement interessant d'eclaircir les phenomenes d'oscillation de puissance et les risques de calefaction (burn out). Les programmes de travaux sur CABRI seront harmonises avec les travaux effectues sur les Spert americains de meme type; lors de sa construction des contacts fructueux ont ete etablis avec les specialistes americains qui ont defini les premiers de ces reacteurs. La communication donne une

Removal of Total Coliforms, Thermotolerant Coliforms, and Helminth Eggs in Swine Production Wastewater Treated in Anaerobic and Aerobic Reactors

Science.gov (United States)

Zacarias Sylvestre, Silvia Helena; Lux Hoppe, Estevam Guilherme; de Oliveira, Roberto Alves

2014-01-01

The present work evaluated the performance of two treatment systems in reducing indicators of biological contamination in swine production wastewater. System I consisted of two upflow anaerobic sludge blanket (UASB) reactors, with 510 and 209 L in volume, being serially arranged. System II consisted of a UASB reactor, anaerobic filter, trickling filter, and decanter, being also organized in series, with volumes of 300, 190, 250, and 150 L, respectively. Hydraulic retention times (HRT) applied in the first UASB reactors were 40, 30, 20, and 11 h in systems I and II. The average removal efficiencies of total and thermotolerant coliforms in system I were 92.92% to 99.50% and 94.29% to 99.56%, respectively, and increased in system II to 99.45% to 99.91% and 99.52% to 99.93%, respectively. Average removal rates of helminth eggs in system I were 96.44% to 99.11%, reaching 100% as in system II. In reactor sludge, the counts of total and thermotolerant coliforms ranged between 105 and 109 MPN (100 mL)−1, while helminth eggs ranged from 0.86 to 9.27 eggs g−1 TS. PMID:24812560

SP-100 space nuclear power system

International Nuclear Information System (INIS)

Given, R.W.; Morgan, R.E.; Chi, J.W.H.; Westinghouse Electric Corp., Madison, PA)

1984-01-01

A baseline design concept for a 100 kWe nuclear reactor space power system is described. The concept was developed under contract from JPL as part of a joint program of the DOE, DOD, and NASA. The major technical and safety constraints influencing the selection of reactor operating parameters are discussed. A lithium-cooled compact fast reactor was selected as the best candidate system. The material selected for the thermoelectric conversion system was silicon germanium (SiGe) with gallium phosphide doping. Attention is given to the improved safety of the seven in-core control rod configuration

Effet de la substitution du maïs par le manioc dans l'aliment SUR les ...

African Journals Online (AJOL)

%, (R2) 66% et (R3) 100% de farine de manioc en remplacent du maïs ont été testées. Les données sur la consommation alimentaire, l'évolution du poids vifs et les caractéristiques de la carcasse des mâles à l'âge de 20 semaines ont été ...

Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon

Dose rate in the reactor room and environment during maintenance in fusion reactors

International Nuclear Information System (INIS)

Maki, Koichi; Satoh, Satoshi; Takatsu, Hideyuki; Seki, Yasushi

1995-01-01

According to the International Thermonuclear Experimental Reactor (ITER) conceptual design activity, after reactor shutdown, damaged segments are pulled up from the reactor and hung from the reactor room ceiling by a remote handling device. The dose rate in the reactor room and the environment is estimated for this situation, and the following results are obtained. First, the dose rate in the room is > 10 8 μSv/h. Since this dose rate is 10 7 times greater than the biological radiation shielding design limit of 25 μSv/h, workers cannot enter the room. Second, lenses and optical fiber composed of glass that is radiation resistant up to 10 6 Gy would be damaged after <100 h near the segment, and devices using semiconductors could not work after several hours or so in the aforementioned dose-rate conditions. Third, during suspension of one blanket segment from the ceiling, the dose rate in the site boundary can be reduced by one order by a 23-cm-thicker reactor building roof. To reduce dose rate in public exposure to a value that is less than one-tenth of the public exposure radiation shielding design limit of 100 μSv/yr, the distance of the site boundary from the reactor must be greater than 200 m for a reactor building with a 160-cm-thick concrete roof. 9 refs., 6 figs., 2 tabs

European Research Reactor Conference (RRFM) 2015: Conference Proceedings

International Nuclear Information System (INIS)

2015-01-01

In 2015 the European Research Reactor Conference, RRFM, took place in Bucharest, Romania. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors

European Research Reactor Conference (RRFM) 2016: Conference Proceedings

International Nuclear Information System (INIS)

2016-01-01

The 2016 European Research Reactor Conference, RRFM, took place in Berlin, Germany. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors.

Small propulsion reactor design based on particle bed reactor concept

International Nuclear Information System (INIS)

Ludewig, H.; Lazareth, O.; Mughabghab, S.; Perkins, K.; Powell, J.R.

1989-01-01

In this paper Particle Bed Reactor (PBR) designs are discussed which use 233 U and /sup 242m/Am as fissile materials. A constant total power of 100MW is assumed for all reactors in this study. Three broad aspects of these reactors is discussed. First, possible reactor designs are developed, second physics calculations are outlined and discussed and third mass estimates of the various candidates reactors are made. It is concluded that reactors with a specific mass of 1 kg/MW can be envisioned of 233 U is used and approximately a quarter of this value can be achieved if /sup 242m/Am is used. If this power level is increased by increasing the power density lower specific mass values are achievable. The limit will be determined by uncertainties in the thermal-hydraulic analysis. 5 refs., 5 figs., 6 tabs

Hanford Site 100-N Area In Situ Bioremediation of UPR-100-N-17, Deep Petroleum Unplanned Release - 13245

Energy Technology Data Exchange (ETDEWEB)

Saueressig, Daniel G. [Washington Closure Hanford, 2620 Fermi, Richland, Washington, 99354 (United States)

2013-07-01

In 1965 and 1966, approximately 303 m{sup 3} of Number 2 diesel fuel leaked from a pipeline used to support reactor operations at the Hanford Site's N Reactor. N Reactor was Hanford's longest operating reactor and served as the world's first dual purpose reactor for military and power production needs. The Interim Action Record of Decision for the 100-N Area identified in situ bioremediation as the preferred alternative to remediate the deep vadose zone contaminated by this release. A pilot project supplied oxygen into the vadose zone to stimulate microbial activity in the soil. The project monitored respiration rates as an indicator of active biodegradation. Based on pilot study results, a full-scale system is being constructed and installed to remediate the vadose zone contamination. (authors)

Effet de la composition de différents substrats culturaux sur ...

African Journals Online (AJOL)

SARAH

31 janv. 2014 ... Effet de la substrats culturaux sur paramètres de croissance ... des graines, mais ce substrat a donné des performances acceptables sur la croissance longitudinale des ..... élevés en conteneurs sur différents types de.

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

International Nuclear Information System (INIS)

2005-01-01

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

Space reactor fuels performance and development issues

International Nuclear Information System (INIS)

Wewerka, E.M.

1984-01-01

Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

SP-100 Test Site

International Nuclear Information System (INIS)

Cox, C.M.; Mahaffey, M.K.; Miller, W.C.

1988-01-01

Preparatory activities are well under way at Hanford to convert the 309 Containment Building and its associated service wing to a 2.5 MWt nuclear test facility for the SP-100 Ground Engineering System (GES) test. Preliminary design is complete, encompassing facility modifications, a secondary heat transport system, a large vacuum system to enclose the high temperature reactor, a test assembly cell and handling system, control and data processing systems, and safety and auxiliary systems. The design makes extensive use of existing equipment to minimize technical risk and cost. Refurbishment of this equipment is 75% complete. The facility has been cleared of obstructing equipment from its earlier reactor test. Current activities are focusing on definitive design and preparation of the Preliminary Safety Analysis Report (PSAR) aimed at procurement and construction approvals and schedules to achieve reactor criticality by January 1992. 6 refs

Heavy water moderated gas-cooled reactors; Filiere eau lourde - gaz

Energy Technology Data Exchange (ETDEWEB)

Bailly du Bois, B; Bernard, J L; Naudet, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

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

Fast flux measurements by means of threshold detectors on the reactor 'Melusine'; Mesures de flux rapides a l'aide de detecteurs a seuil sur le reacteur 'Melusine'

Energy Technology Data Exchange (ETDEWEB)

Leger, P; Sautiez, B [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

Using existing data on the (n,p) and (n,{alpha}) threshold reactions we have carried out fast flux measurements on the swimming pool type reactor 'Melusine'. Four common elements: P, S, Mg, Al were chosen because from the point of view of fast spectrum analysis they represent a fairly good energy range from 2.4 MeV to 8 MeV. The fission flux value found in the central element at a power of 1 MW is 1.4 x 10{sup 13} n/cm{sup 2}/s {+-} 0.14. (author) [French] A l'aide des donnees actuelles sur les reactions a seuil (n,p) et (n,{alpha}) nous avons realise des mesures de flux rapide dans le reacteur du type piscine 'Melusine'. Quatre corps courants: P, S, Mg, Al, ont ete choisis parce qu'ils constituent au point de vue de l'analyse du spectre rapide un assez bon etalement en energie de 2,4 MeV A 8 MeV. La valeur du flux de fission trouve dans l'element central a une puissance de 1 MW est de 1,4.10{sup 13} n/cm{sup 2}/s {+-} 0,14. (auteur)

SUR. Breve informe de la Expedición Yelcho al Polo Sur (1909-1910

Directory of Open Access Journals (Sweden)

Ursula K. Le Guin

2013-12-01

Full Text Available Ursula K. Le Guin Escritora norteamericana, famosa por sus obras de ciencia fi cción y fantasía, en las cuales el tema de género ocupa una posición central. Ganadora de los premios Hugo y Nébula que la catapultaron a la fama. “SUR. A summary report of the Yelcho expedition to the Antarctic, 1909-1910”. Apáreció publicado por primera vez en la revista New Yorker, el 1° de febrero de 1982. La palabra SUR aparece en español en el original. (http://www.newyorker.com/ar Traducción de Susana E. Matallana Peláez

Sur terre comme sur mer: organisations spatiales en mer du Nord et mer de Norvège

Directory of Open Access Journals (Sweden)

Maryvonne LE BERRE

1998-09-01

Full Text Available L'article montre, au moyen de représentations cartographiques et graphiques, les structures spatiales développées en mer pour l'exploitation des hydrocarbures. Malgré les différences de milieu, l'appropriation et l'occupation territoriales du domaine maritime s'effectuent selon les mêmes principes que sur le continent, à ceci près que ce qui s'étale en surface sur le continent se développe à la verticale en mer.

REACTOR SHIELD

Science.gov (United States)

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Numerical reactor evaluation

International Nuclear Information System (INIS)

Venter, A.M.

1973-08-01

A short discussion is given of the physics of a nuclear reactor and the parameters which are used in the study of neutron transport. The mathematical formulation and detailed derivation is given of the neutron diffusion and transport equations. A description is given of the computer programmes, FIRE-5 and PELSN, developed at Pelindaba for the evaluation of both thermal and fast reactor systems. It is indicated how these computer programmes have been applied in the study of the PELINDUNA-O and other known critical facilities. The application of Lie-series to the solution of the neutron diffusion equation is discussed in detail. The time dependence of the variables is removed by means of a Laplacetransformation and the semi-analytical solution is written in terms of a transfer matrix. A complete set of recursion formulae, applicable to both homogeneous and heterogeneous reactor systems, is derived. The method used in the evaluation of the effective multiplication factor, k-eff, and the alpha-eigen-value is described. A computer programme was written to solve the neutron diffusion equation in terms of the Lie-series. The results are compared with the TIMOC and PELSN computer programmes. A method is suggested in which the Lie-series are used to solve the neutron transport equation. The transfer matrix for this case, is derived. A complete discussion is given of the solution to the space and time dependent diffusion equation in the presence of a delta source [af

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Influence of the irradiation by the {gamma} rays on various factors of immunity; Influence de l'irradiation par les rayons {gamma} sur divers facteurs d'immunite

Energy Technology Data Exchange (ETDEWEB)

Gaude, G; Coursaget, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

Exposed animals often present, in the days or weeks following the irradiation, heavy infectious lesions. The goal of this survey was to search the influence of the total irradiation by the gamma radiation from cobalt-60, on the organism, on the phagocytic power of the leukocytes and on the biogenesis of antibodies. The experiences were done on hundred animals, rabbits and guinea pigs. All irradiations have been achieved with eight sources of Cobalt 60 of 10 curies each, disposed on a sphere of 1 m, the animal living the region neighboring of the center. The intensity of the radiation, in these conditions, is about 100 roentgens per hour. (authors) [French] Les animaux irradies presentent souvent, dans les jours ou les semaines qui suivent l'irradiation, de graves lesions infectieuses. Le but de cette etude etait de rechercher l'influence de l'irradiation totale de l'organisme par le rayonnement gamma du cobalt 60 sur le pouvoir phagocytaire des leucocytes et sur la biogenese des anticorps. Les experiences portent sur une centaine d'animaux, lapins et cobayes. Toutes les irradiations ont ete realisees avec huit sources de Cobalt 60 de 10 curies chacune, disposees sur une sphere de 1 m de rayon, les animaux ocoupant la region voisine du centre. L'intensite du rayonnement est, dans ces conditions, d'environ 100 roentgens par heure. (auteurs)

Hausse des taxes sur le tabac et politiques de lutte antitabac au ...

International Development Research Centre (IDRC) Digital Library (Canada)

Par ailleurs, les études existantes sur les effets des taxes sur le tabac sont fondées sur l'" élasticité de la demande en fonction du prix ", une mesure qui indique l'évolution de la demande globale de cigarettes en réaction à une modification des prix, mais qui ne tient pas compte des effets d'une augmentation des taxes sur ...

Australian research reactor studies

International Nuclear Information System (INIS)

McCulloch, D.B.

1978-01-01

The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)

Reactor Physics Development for Advanced Gas-Cooled Reactors; Recherches en Physique des Reacteurs, pour des Reacteurs Perfectionnes Refroidis par un Gaz; Razrabotka metodov v oblasti reaktornoj fiziki dlya usovershenstvovannogo reaktora s gazovym okhlazhdeniem; Progresos de la Fisica de los Reactores de Tipo Avanzado Refrigerados por Gas

Energy Technology Data Exchange (ETDEWEB)

Moore, J. [United Kingdom Atomic Energy Authority (United Kingdom)

1964-04-15

verifier les methodes theoriques elaborees pour etudier des coeurs de reacteurs heterogenes. Ces methodes theoriques, utilisees jusqu'a ce jour, sont connues sous les noms dr 'hetrecontrol' et de 'FTD2'. Les experiences avaient pour but de verifier dans le detail les caracteristiques de ces methodes; on a analyse les mesures faites sur plusieurs coeurs de 'reacteur' de differentes dimensions dans les installations APEX et HERO pour determiner une serie coherente de constantes de reseau concordant avec les resultats des experiences. A ces constantes purement empiriques, on a applique ensuite les methodes <> et 'FTD2' pour preparer la mise en service sans accord d'AGR et le choix du regime de chargement de ce reacteur. Le memoire enumere les techniques experimentales qui ont ete essayees et celles qui ont ete elaborees pour resoudre certains problemes qui se presentaient. Particulierement interessantes sont les methodes ayant pour but de mesurer les effets sur la reactivite dans les installations APEX, HERO et AGR, et de determiner les donnees relatives a la structure fine ainsi que la repartition de la puissance dans les assemblages complexes. Les recherches theoriques actuelles et futures sont axees principalement sur la mise au point d'une methode capable de remplacer 'hetrecontrol' et 'FTD2' pour les etudes sur des coeurs de reacteur apres qu'une bonne partie du combustible a 'brule'. Le programme d'experiences avec l'installation HERO a pour but de verifier ces methodes au moyen de coeurs complexes contenant du plutonium. On compte obtenir des renseignements supplementaires sur l'effet du plutonium au cours du fonctionnement d'AGR et a la suite de mesures de physique sur le combustible irradie. (author) [Spanish] La memoria describe los trabajos experimentales y teoricos que se han ejecutado durante el diseno, el desarrollo y la puesta en marcha del reactor AGR de Windscale y para facilitar el desarrollo de nuevos tipos de reactores refrigerados por gas

Activated Sludge and Aerobic Biofilm Reactors

OpenAIRE

Von Sperling, Marcos

2007-01-01

"Activated Sludge and Aerobic Biofilm Reactors is the fifth volume in the series Biological Wastewater Treatment. The first part of the book is devoted to the activated sludge process, covering the removal of organic matter, nitrogen and phosphorus.A detailed analysis of the biological reactor (aeration tank) and the final sedimentation tanks is provided. The second part of the book covers aerobic biofilm reactors, especially trickling filters, rotating biological contractors and submerged ae...

Cleanup Verification Package for the 100-K-55:1 and 100-K-56:1 Pipelines and the 116-KW-4 and 116-KE-5 Heat Recovery Stations

International Nuclear Information System (INIS)

Capron, J.M.

2005-01-01

This cleanup verification package documents completion of remedial action for the 100-K-55:1 and 100-K-56:1 reactor cooling effluent underground pipelines and for the 116-KW-4 and 116-KE-5 heat recovery stations. The 100-K-55 and 100-K-56 sites consisted of those process effluent pipelines that serviced the 105-KW and 105-KE Reactors. This cleanup verification package documents completion of remedial action for the 100-K-55: 1 and 100-K-56: 1 reactor cooling effluent underground pipelines, referred to herein as the 100-K-55:1 and 100-K-56:l sites, as well as for the 116-KW-4 and 116-KE-5 heat recovery stations, referred to herein as the 116-KW-4 and 116-KE-5 sites. The 116-KW-4 and 116-KE-5 heat recovery stations were co-located and remediated with the 100-K-55:1 and 100-K-56:1 pipelines, respectively. These sites are located in the 100-KR-2 Operable Unit in the 100-K Area of the Hanford Site in southeastern Washington State. The 100-K-55 and 100-K-56 sites consisted of those process effluent pipelines that serviced the 105-KW and 105-KE Reactors, respectively. Both of these sites have been administratively divided into subunits based on the current extent of remediation. Portions of the pipelines remaining within the reactor security fencing and in proximity to active utility features have been delineated as the 100-K-55:2 and 100-K-56:2 pipelines, with the portions of the pipelines excluded from these boundaries identified as the 100-K-55:1 and 100-K-56:1 pipelines. This cleanup verification package addresses only the 100-K-55:1 and 100-K-56:I subunits; the 100-K-55:2 and 100-K-56:2 subunits will be addressed within a separate cleanup verification package. Site excavation and waste disposal are complete, and the exposed surfaces have been sampled and analyzed to verify attainment of the remedial action goals. Results of the sampling, laboratory analyses, and data evaluations for the 100-K-55:1, 100-K-56:1, 116-KW-4, and 116-KE-5 sites indicate that all remedial

Radiation conditions at the training IRT-2000 and IR-100 reactors

International Nuclear Information System (INIS)

Fedorin, Eh.V.; Bronshtejn, I.Eh; Martynov, Yu.N.; Chistyakov, N.I.

1978-01-01

The experience is reviewed of radiation hygiene surveys and radiation safety provision during instructional processes on two training and research nuclear reactors of the IRT-2000 type (No. 1 and No. 2) and on an IR-200 reactor. From an analysis of individual dosimetry data the conclusion is made that the trainees and personnel are exposed mainly to external gamma-radiation and also, to a minor degree, to thermal neutrons and beta-radiation. It has been found that a high level of radiation safety is ensured on the training and research so that research and instruction activities are conducted at annual levels of exposure substantially lower than 0.5 rem in the case of trainees and 5 rem in the case of personnel

10 CFR 100.11 - Determination of exclusion area, low population zone, and population center distance.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Determination of exclusion area, low population zone, and population center distance. 100.11 Section 100.11 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) REACTOR... and for Testing Reactors § 100.11 Determination of exclusion area, low population zone, and population...

Modeling Taylor series approximations for prompt neutron kinetics with lab view simulations

International Nuclear Information System (INIS)

Adzri, E. P.

2012-09-01

The reactor point kinetics equations have been subjected to intense research in an effort to find simple yet accurate numerical solutions methods. The equations are very stiff numerically, meaning that there is a wide variation in the decay constants, so that using a particular time step in the numerical solution may provide sufficient accuracy for the group, but not for another. Several solutions techniques have been presented on the point kinetics equations with varying degrees of complexity. These include Power Series Solutions, CORE, PCA, Genapol and Taylor series methods. In this research, algorithms were developed based on the first and second order Taylor series expansion and simulated in LabVIEW to solve the Reactor Point Kinetics equations using block diagram nodes implemented within stacked sequences. The algorithms developed were fast,accurate and simple to code. Several reactivity insertions were used to simulate the change in neutron population with time. The LabVIEW- Taylor series solutions were compared with other solution techniques such as Power Series Solutions, CORE, PCA, Genapol and McMahon and Pierson's Taylor series approximation. The results of LabVIEW-Taylor series technique used by McMahon and Pearson The LabVIEW-implemented techniques were found to agree very well with these other methods. At 1x10 -8 s the neutron population was 1.000220 neutrons / cm 3 , at 1 x 10 -2 s it was 2.007681 neutrons / cm 3 and at 1x10 -1 s it was 2.075317 neutrons / cm 3 ; same results reported by Genapol for a fast reactor, it produced good and accurate results and compared very favorably with other methods found in the literature. Using much smaller time steps to the order or 10 -8 s commensurate with fast reactor parameters also produced very satisfactory results, indicating that the LabVIEW-based Taylor series technique is suitable for simulating the kinetics of fast reactors as well as thermal reactors. Algorithms developed that included second order terms

Effets de six composts sur les réponses physiologiques ...

African Journals Online (AJOL)

Les effets de ces six composts sur le potentiel hydrique, la fuite d'électrolyte, la teneur en amidon des feuilles et sur quelques paramètres agronomiques ont été évalués sur les plantes stressées en comparaison avec les plantes témoins. Les résultats indiquent des valeurs du potentiel hydrique foliaire très faibles et une ...

Pulsed fusion reactors

International Nuclear Information System (INIS)

1975-01-01

This summer school specialized in examining specific fusion center systems. Papers on scientific feasibility are first presented: confinement of high-beta plasma, liners, plasma focus, compression and heating and the use of high power electron beams for thermonuclear reactors. As for technological feasibility, lectures were on the theta-pinch toroidal reactors, toroidal diffuse pinch, electrical engineering problems in pulsed magnetically confined reactors, neutral gas layer for heat removal, the conceptual design of a series of laser fusion power plants with ''Saturn'', implosion experiments and the problem of the targets, the high brightness lasers for plasma generation, and topping and bottoming cycles. Some problems common to pulsed reactors were examined: energy storage and transfer, thermomechanical and erosion effects in the first wall and blanket, the problems of tritium production, radiation damage and neutron activation in blankets, and the magnetic and inertial confinement

Characteristic time series and operation region of the system of two tank reactors (CSTR) with variable division of recirculation stream

International Nuclear Information System (INIS)

Merta, Henryk

2006-01-01

The paper deals with a system of a cascade of two tank reactors, being characterized by the variable stream of recirculating fluid at each stage. The assumed mathematical model enables one to determine the system's dynamics for the case when there is no time delay and for the opposite case. The time series of the conversion degree and of the dimensionless fluid temperature, characteristic for the system considered as well as the operation regions-the latter-basing on Feingenbaum diagrams with respect to the division ratio of the recirculating stream are presented

Sur un deplacement de valeurs: traire et tirer

Directory of Open Access Journals (Sweden)

André Burger

1972-12-01

Full Text Available L'étymologie de firer est inconnue. Celie que le regretté W. von Wart burg a ern pouvoir avancer est inacceptable: firer serait sorti de marty rier par la grâce d'une fausse coupure mar tirier. Il est plus qu'improbable que cette «étymologie populaire» eût pu se produire si la langue ne possédait pas au préalable un verbe tirier. Au surplus aucun fait n'appuie cette hypothèse. Le FEW, t. VI, 1, p. 396, donne bien pour firer le sens de «torturer sur un treteam>, au XIIIe siècle, et p. 403, «démembrer en faisant tirer les quatre membles par des chevaux» et «torturer (qn en l'étendant sur un tréteau” au XIVe siècle; ces sens sont évidemment trop tardifs pour permettre des conclusions sur l'origine de firer, d'autant plus qu'ils s'expliquent sans peine par la valeur normale du mot, attesté dès la Chanson de Roland, sans aucun rapport avec martirie, mot savant de clerc, qui, dans le même texte s'applique deux fois sur quatre exemples aux Sarrasins, vv. 501 et 1467, où il ne signifie pas «martyre» mais <sur son étymologie et non sur les textes, que tirer aurait à l'origine »einen starken affektiven unterton« (FEW XIII, 2, p. 185. L'étude des exemples de la Chanson de Roland, confrontés avec ceux de traire, nons amènera à une conclusion toute différente.

Effets de l'audit clinique basé sur des critères sur de la qualité de la ...

African Journals Online (AJOL)

Effets de l'audit clinique basé sur des critères sur de la qualité de la prise en charge de la prééclampsie sévère dans le Département de Gynécologie Obstétrique du Centre Hospitalier Universitaire Yalgado Ouédraogo, Ouagadougou (Burkina Faso). Sibraogo Kiemtoré, Adama Dembélé, Adama Ouattara, Hyacinthe1,2 ...

Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

International Nuclear Information System (INIS)

Wright, A.E.; Dutt, D.S.; Harrison, L.J.

1990-01-01

Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

Conférence sur l'efficacité

CERN Document Server

Jullien, François

2005-01-01

Philosophe et sinologue, François Jullien présente ici une conférence qu'il a prononcée auprès de chefs d'entreprise et dans le milieu du management. D'un côté, la conception européenne de l'efficacité est liée à la modélisation comme à la finalité et revendique l'action jusqu'à l'héroïsme ; de l'autre, la pensée chinoise de l'efficience, indirecte et discrète, s'appuie sur le potentiel de situation et induit des " transformations silencieuses ", sans éclat ni même événement. Par-delà cet écart, il s'agira d'interroger la nature de l'effectivité ; ou comment l'intervention humaine réussit à se brancher sur la propension des choses et s'y laisse intégrer. Ce propos se garde donc de séparer tant soit peu l'art d'opérer sur des situations et l'exercice de la philosophie ; en résultent des effets de lecture portant sur l'histoire du XXe siècle ainsi que la géopolitique - et géoéthique - à venir.

Standard deviation of local tallies in global Monte Carlo calculation of nuclear reactor core

International Nuclear Information System (INIS)

Ueki, Taro

2010-01-01

Time series methodology has been studied to assess the feasibility of statistical error estimation in the continuous space and energy Monte Carlo calculation of the three-dimensional whole reactor core. The noise propagation was examined and the fluctuation of track length tallies for local fission rate and power has been formally shown to be represented by the autoregressive moving average process of orders p and p-1 [ARMA(p,p-1)], where p is an integer larger than or equal to two. Therefore, ARMA(p,p-1) fitting was applied to the real standard deviation estimation of the power of fuel assemblies at particular heights. Numerical results indicate that straightforward ARMA(3,2) fitting is promising, but a stability issue must be resolved toward the incorporation in the distributed version of production Monte Carlo codes. The same numerical results reveal that the average performance of ARMA(3,2) fitting is equivalent to that of the batch method with a batch size larger than 100 and smaller than 200 cycles for a 1,100 MWe pressurized water reactor. (author)

Ground test facility for nuclear testing of space reactor subsystems

International Nuclear Information System (INIS)

Quapp, W.J.; Watts, K.D.

1985-01-01

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

Use of cadmium in solution in the EL 4 reactor moderator irreversible fixing of cadmium on the metallic surfaces; Utilisation du cadmium en solution dans le moderateur du reacteur EL 4 - fixation irreversible du cadmium sur les surfaces metalliques

Energy Technology Data Exchange (ETDEWEB)

Croix, O; Paoli, O; Lecomte, J; Dolle, L; Gallic, Y [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

In the framework of research into the poisoning of the EL-4 reactor by cadmium sulphate, measurements have been made by two different methods of the residual amounts of cadmium liable to be fixed irreversibly on the surfaces in contact with the heavy water. A marked influence of the pH has been noticed. The mechanism of the irreversible fixing is compatible with the hypothesis of an ion-exchange in the surface oxide layer. In a sufficiently wide range of pH the cadmium thus fixed causes very little residual poisoning. The stability of the cadmium sulphate solutions is however rather low in the conditions of poisoning. (authors) [French] Dans le cadre des etudes sur l'empoisonnement du reacteur EL-4 par le sulfate de cadmium, les quantites residuelles de cadmium susceptibles de se fixer irreversiblement sur les parois que mouillerait l'eau lourde, ont ete mesurees experimentalement par deux methodes differentes. On observe une influence nette du pH. Le mecanisme de la fixation irreversible est compatible avec l'hypothese d'un echange d'ions dans la pellicule d'oxyde superficielle. Dans des limites suffisamment larges de pH, la cadmium ainsi fixe n'occasionne pas d'empoisonnement residuel important. La stabilite des solutions de sulfate de cadmium dans les conditions de l'empoisonnement est cependant mediocre. (auteurs)

SP-100 multimegawatt scaleup to meet electric propulsion mission requirements

International Nuclear Information System (INIS)

Newkirk, D.W.; Salamah, S.A.; Stewart, S.L.; Pluta, P.R.

1991-01-01

The SP-100 space power nuclear reactor nuclear heat source technology, utilizing uranium nitride fuel clad in PWC-11 in a fast reactor with lithium coolant circulated by an electromagnetic pump, is shown in this paper to be directly extrapolatable to thermal power levels that meet NASA nuclear electric propulsion requirements using different power conversion techniques. The SP-100 nuclear technology can be applied for missions with NEP requirements as low as 10's of kWe to 10's of MWe

Taxes sur les cigarettes en Tanzanie | IDRC - International ...

International Development Research Centre (IDRC) Digital Library (Canada)

Taxes sur les cigarettes en Tanzanie. De 2002 à 2007, le tabagisme a connu une hausse de 20 % en Tanzanie, une augmentation qui devrait atteindre 46 % en 2016 si la tendance se maintient. On s'attend à ce que cette hausse du tabagisme ait d'importantes répercussions sur la santé publique et le développement ...

Fundamentals of pressurized water reactors

International Nuclear Information System (INIS)

Murray, L.

1982-01-01

In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

Development task of compact reactor

International Nuclear Information System (INIS)

Kurushima, Morihiro

1982-01-01

In the Ministry of International Trade and Industry, studies proceed on the usage of compact medium and small LWRs. As such, the reactors from 100 to 200 MW may meet varieties of demands in scale and kind in view of the saving of petroleum and the economy of nuclear power. In this case, the technology of light water reactors with already established safety will be suitable for the development of compact reactors. The concept of ''nuclear power community'' using the compact reactors in local society and industrial zones was investigated. The following matters are described: need for the introduction of compact reactors, the survey on the compact reactor systems, and the present status and future problems for compact reactor usage. (J.P.N.)

Asesinas en series: violencia femenina en la televisión mexicana

OpenAIRE

Michael, Joachim

2014-01-01

Mujeres que matan o que toman el control de un cártel avanzaron hacia el centro de interés de los productores televisivos. Con series como Mujeres asesinas o La reina del sur la televisión pretende mostrar que ya superó la tradición melodramática que cualifica la mujer o como víctima abnegada o como ‘devoradora’. Estas series proponen que el género femenino no sólo sufre la agresión del hombre o lo seduce con sus ‘propias armas’ sino que le responde a éste con las ‘mismas armas’. Lo que está ...

New models of radical polymerization with branching and scission predicting molecular weight distribution in tubular and series of continuous stirred tank reactors allowing for multiradicals and gelation

NARCIS (Netherlands)

Yaghini, N.; Iedema, P.D.

2015-01-01

Modeling of the mol. wt. distribution (MWD) of low-​d. Polyethylene (ldPE) has been carried out for a tubular reactor under realistic non-​isothermal conditions and for a series of CSTR's. The model allows for the existence of multiradicals and the occurrence of gelation. The deterministic model is

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

This book concentrates on the different types of noise present in power reactors and how the analysis of this noise can be used as a tool for reactor monitoring and diagnostics. Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions thus preventing further complications by alerting operators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussion of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

Selection of catalysts and reactors for hydroprocessing

Energy Technology Data Exchange (ETDEWEB)

Furimsky, E. [Imaf Group, Ottawa, ON (Canada)

1998-07-13

The performance of hydroprocessing units can be influenced by the selection of the catalysts and the type of reactor to suit a particular feed. The catalysts and reactors selected for light feeds differ markedly from those selected for heavy feeds. Fixed-bed reactors have been traditionally used for light feeds. High asphaltene and high metal content feeds are successfully processed using moving-bed and/or ebullated bed reactors. Multi-reactor systems consisting of moving-bed and/or ebullated bed reactors in series with fixed-bed reactors can be used to process difficult feeds. For heavy feeds, the physical properties (e.g. porosity), shape and size of the catalyst particles become crucial parameters. Pretreatment of catalysts by presulfiding improves the performance of the units.

Utilization of the SLOWPOKE-2 research reactor

International Nuclear Information System (INIS)

Lalor, G.C.

2001-01-01

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

Le Projet d’Ole Lando sur les Contrats

DEFF Research Database (Denmark)

Holle, Marie-Louise

2017-01-01

En 2016 le projet de recherche le plus important jusqu’à présent sur le droit nordique des contrats a pris fin et un livre, de presque 400 pages, en a marqué l’aboutissement. Le résultat du projet est un « restatement » tel qu’il existe aux États-Unis. Le Restatement porte sur les règles et les...

Principles of Inherent Self-Protection Realized in the Project of Small Size Modular Reactor SVBR-100

International Nuclear Information System (INIS)

Toshinsky, G.I.; Petrochenko, V.V.; Komlev, O.G.; Tormyshev, I.V.; Dedul, A.V.

2013-01-01

Conclusions: • From the standpoint of population, the opportunity of catastrophic consequences caused by nuclear accident is much more important than very low possibility of its realization. • The level of social acceptability of future large-scale NP must be higher. • Use of the nuclear power technology (NPT) based on RFs, in which the value of stored potential energy of different kinds is minimal, will meet that goal the most efficient. • Those RFs cannot amplify the external impacts, therefore, the scale of damages will be only determined by the external impact energy, the exhaust of radioactivity being localized. • Now there are no developed NPTs with such properties. • The NPT based on modular fast reactors SVBR-100 and verified in conditions of NS operating is to the most extent ready to be demonstrated. • Federal target program “New Generation Nuclear Power Technologies for 2010 – 2015 Years and Future Trends up to 2020” stipulates the construction of experimental-industrial power-unit SVBR-100. • The project is realized within the frameworks of state-private partnership by joint venture JSC “AKME-Engineering” organized on a parity basis by State Atomic Energy Corporation “Rosatom” and Limited Liability Company “Irkutskenergo”. • The first of a kind power unit with RF SVBR-100 will be commissioned in 2017 near the SSC NIIAR site in Dimitrovgrad (Ulyanovsk region). • Widespread common use of this NPT, which potentials are very high, is expected to begin in ∼ 2020 – 2025. In case of earlier starting, the economic risk will be high; in case it is launched later, much profit will be lost

Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

The 20-MW Replacement Research Reactor Project which is currently under construction at ANSTO will have a combination of a state of the art triplicated computer based reactor protection system, and a fully independent, and diverse, triplicated analogue reactor protection system, that has been in use in the nuclear industry, for many decades. The First Reactor Protection System (FRPS) consists of a Triconex triplicated modular redundant system that has recently been approved by the USNRC for use in the USA?s power reactor program. The Second Reactor Protection System is a hardwired analogue system supplied by Foxboro, the Spec 200 system, which is also Class1E qualified. The FRPS is used to drop the control rods when its safety parameter setpoints have been reached. The SRPS is used to drain the reflector tank and since this operation would result in a reactor poison out due to the time it would take to refill the tank the FRPS trip setpoints are more limiting. The FRPS and SRPS have limited hardwired indications on the control panels in the main control room (MCR) and emergency control centre (ECC), however all FRPS and SRPS parameters are capable of being displayed on the reactor control and monitoring system (RCMS) video display units. The RCMS is a Foxboro Series I/A control system which is used for plant control and monitoring and as a protection system for the cold neutron source. This paper will provide technical information on both systems, their trip logics, their interconnections with each other, and their integration into the reactor control and monitoring system and control panels. (author)

Diacronía y sincronía del contacto lingüístico en Patagonia sur

OpenAIRE

Virkel, Ana Ester

2012-01-01

Desde su integración al Estado argentino, la región patagónica fue receptora de inmigrantes de muy diversas procedencias, quienes, integrándose con las etnias aborígenes preexistentes, configuraron un contexto multiétnico y multicultural, con la consiguiente diversidad lingüística. El propósito de este trabajo es trazar un esquemático itinerario a través de la historia lingüística de Chubut, Patagonia sur, focalizando el interés en una serie de procesos migratorios que impactan de modo sus...

Space reactors, a prospective for the future

International Nuclear Information System (INIS)

Wahlquist, E.; Voss, S.S.

1989-01-01

The power requirements for future space missions are increasing and alternate power systems will be required to meet these needs. Therefore, in the early 1980's a tri-agency space reactor program, the SP-100, was initiated that is capable of meeting the higher power requirements. To understand the current space reactor program, it is important to review it in the context of past space nuclear programs - including radioisotopes, nuclear rockets and reactors. Initial effort on these programs began in the mid-1950's. Radioisotope generators have been flown on a variety of missions and are continuing to be used. The space reactor and nuclear rocket programs were technically successful but were both terminated in 1973. The current SP-100 program builds on those earlier programs

Mutations du gène d'apoprotéine B100 au cours des dyslipidémies ...

African Journals Online (AJOL)

2 Centre de Diagnostic et de Recherche sur le SIDA et les autres Maladies Infectieuses, CHU ... d'Apolipoprotéine B100 chez les personnes vivant avec le VIH (PV-VIH) présentant une dyslipidémie. .... d'ions par la technique du kit Qiagen®.

History of fast reactor development in U.S.A.-I

International Nuclear Information System (INIS)

Ninokata, Hisashi; Sasao, Nobuyki

2007-01-01

History and present state of fast reactor was reviewed in series. As a history of fast reactor development in U.S.A. - I, this third lecture presented the dawn of the fast reactor development in the USA. The first fast reactor was the Clementine reactor with plutonium fuels and mercury coolant. The LAMPRE-1 reactor was the first sodium cooled and molten plutonium reactor. Experimental breeder reactor (EBR-1) was the first reactor to produce electricity and four kinds of fuels were loaded. Zero-power reactors were constructed to conduct reactor physics experiments on fast reactors. Today there are renewed interests in fast reactors due to their ability to fission actinides and reduce radioactive wastes. (T. Tanaka)

Droits, justice et réseaux sociaux sur Internet (Amérique latine ...

International Development Research Centre (IDRC) Digital Library (Canada)

... sur la question de la protection des renseignements personnels sur Internet, ... Derechos y justicia y el movimiento social en internet (5 de agosto de 2009 ... Programme de recherche sur le virus Zika Canada-Amérique latine et Caraïbes.

Space nuclear reactor safety

International Nuclear Information System (INIS)

Damon, D.; Temme, M.; Brown, N.

1990-01-01

Definition of safety requirements and design features of the SP-100 space reactor power system has been guided by a mission risk analysis. The analysis quantifies risk from accidental radiological consequences for a reference mission. Results show that the radiological risk from a space reactor can be made very low. The total mission risk from radiological consequences for a shuttle-launched, earth orbit SP-100 mission is estimated to be 0.05 Person-REM (expected values) based on a 1 mREM/yr de Minimus dose. Results are given for each mission phase. The safety benefits of specific design features are evaluated through risk sensitivity analyses

Fast reactor collaboration in Europe

International Nuclear Information System (INIS)

Smith, G.E.I.

1987-01-01

Fast reactors have been developed in several European countries, the United Kingdom, France, Germany and Italy. A suggestion to collaborate on fast reactor research and development resulted in an Intergovernmental Memorandum of Understanding signed in 1984 by the UK, France, Germany, Italy and Belgium. Holland was expected to join later. This provided for co-operation between electric utilities, reactor design, research and development companies and fuel cycle companies. Three steering committees have so far been set up, the European fast reactor utilities Group, the European research and development and the European fuel cycle steering committees. Progress on these is detailed. The main areas of technology exchange are listed in the Appendix. The possibility exists for a series of three large demonstration plants to be built in Europe and a fuel reprocessing plant to confirm the reactor system. (U.K.)

A program for dynamic noise investigations of reactor systems

International Nuclear Information System (INIS)

Antonov, N.A.; Yaneva, N.B.

1980-01-01

A stochastic process analysis in nuclear reactors is used for the state diagnosis and dynamic characteristic investigation of the reactor system. A program DENSITY adapted and tested on an IBM 360 ES type computer is developed. The program is adjusted for fast processing of long series exploiting a relatively small memory. The testing procedure is discussed and the method of the periodic sequences corresponding to characteristic reactivity perturbations of the reactor systems is considered. The program is written for calculating the auto-power spectral density and the cross-power spectral density, as well as the coherence function of stationary statistical time series using the advantages of the fast Fourier transformation. In particular, it is shown that the multi-frequency binary sequences are very useful with respect to the signal-to-noise ratio and the frequency distribution in view of the frequency reactor test

Case series

African Journals Online (AJOL)

abp

27 sept. 2017 ... Résumé. La synovite villonodulaire pigmentée (SVNP) est une prolifération bénigne rare de la synoviale des articulations, des bourses séreuses et des gaines tendineuses, d'étiopathogénie inconnue. Notre travail porte sur 20 cas de SVN du genou colligés à l'hôpital militaire Avicenne de Marrakech sur ...

El espacio cultural transnacional en la post-Transición. El caso de las series televisivas Amores difíciles y La reina del Sur / The Transnational Cultural Space in the Post-Transition. The Case of the TV series Amores difíciles and La Reina del Sur

Directory of Open Access Journals (Sweden)

Manuel Palacio

2017-02-01

Full Text Available Tomando como punto de partida el trabajo de Alberto Elena, Cruces de destinos. Intercambios cinematográficos entre España y América Latina, los autores analizan cómo ha evolucionado el panorama audiovisual latinoamericano al mismo tiempo que cambiaba la política cultural española con respecto a América, ejemplarizado en sendas coproducciones y adaptaciones literarias como son Amores difíciles y La reina del sur.Palabras clave: Post-Transición, televisión transnacional, coproducción, Pérez Reverte, García Márquez, adaptación literaria.Abstract:Taking as a starting point Alberto Elena’s work, Cruces de destinos. Intercambios cinematográficos entre España y América Latina, the authors analyse how the Latin American audio-visual scene has evolved at the same time that the Spanish cultural politics was changing with respect to Latin America. This will be exemplified in both co-productions and literary adaptations: Amores difíciles and La reina del sur.Keywords: Post-Transition, transnational television, co-production, Pérez Reverte, García-Márquez, literary adaptation.

Cleanup Verification Package for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault

International Nuclear Information System (INIS)

Appel, M.J.

2006-01-01

This cleanup verification package documents completion of remedial action for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault. The site consisted of an inactive solid waste storage vault used for temporary storage of slightly contaminated reactor parts that could be recovered and reused for the 100-F Area reactor operations

Cleanup Verification Package for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault

Energy Technology Data Exchange (ETDEWEB)

M. J. Appel

2006-11-02

This cleanup verification package documents completion of remedial action for the 118-F-7, 100-F Miscellaneous Hardware Storage Vault. The site consisted of an inactive solid waste storage vault used for temporary storage of slightly contaminated reactor parts that could be recovered and reused for the 100-F Area reactor operations.

Trade study for kWe class space reactors

Science.gov (United States)

Bost, Donald S.

Recent interest by NASA and other government agencies in space reactor power systems with power levels in the 1 to 100 kWe range has prompted a review of earlier space reactor programs, as well as the ongoing SP-100 program, to identify a system that will best fulfill their needs. The candidate reactor types that were reviewed are listed. They are categorized according to the method of heat removal. The five types are: conduction cooled, heat pipe cooled, liquid metal cooled, in-core thermionic and gas cooled. The UZrH moderated reactor coupled with an organic Rankine cycle power conversion system provides an attractive system for multikilowatt, long lived missions. The reactor requires a minimum development because a similar reactor has already flown and the ORC is being developed for use in the Dynamic Isotope Power System (DIPS) and on the Space Station.

Nuclear reactor power for a space-based radar. SP-100 project

Science.gov (United States)

Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin

1986-01-01

A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

Reactor power control device

International Nuclear Information System (INIS)

Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

1998-01-01

The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

Feasibility study of the underwater neutron radiography facility using the University of Utah 100 kW TRIGA (UUTR) reactor

International Nuclear Information System (INIS)

Choe, D.; Xiao, S.; Jevremovic, T.; Yang, X.

2010-01-01

The University of Utah 100 kW TRIGA (UUTR) reactor provides usable neutron yields for neutron radiography. Currently, UUTR reactor has three irradiators (Central, Pneumatic, and Thermal irradiators) and one Fast neutron Irradiation Facility (FNIF). These irradiators are very small so they are not suitable for neutron radiography. UUTR has three beam ports but they are not available due to the structure of the core. All sides of the core are occupied by FNIF, Thermal Irradiator, and three ion chambers. The only available position for underwater vertical beam port is on the top of the FNIF. There are two factors necessary to fulfill to be able to realize vertical underwater beam port: noninterruption to other facilities and radiation shielding. Designing the vertical beam port as movable ensures good access to the core and pool, while still providing a good neutron radiography environment. Keeping the top of the beam port below the surface of the pool the water represents biological shield. Neutron radiographs, with a simple setup of efficient neutron converters and digital camera systems, can produce acceptable resolution with an exposure time as short as a few minutes. It is important to validate the design with calculations before constructing the beam port. The design of the beam port is modeled using the MCNP5 transport code. A minimum of 10 5 neutrons/cm 2 -sec thermal neutron flux is required for high resolution neutron radiography. Currently, the UUTRIGA is in the process of upgrading its power from 100 kW to 250 kW. Upon the completion of the upgrading, the maximum neutron flux in the core will be ∼7x10 12 neutrons/cm 2 -sec. This paper discusses a modeling and evaluation of capability for a neutron radiography facility. (author)

Recommendations for a restart of molten salt reactor development

International Nuclear Information System (INIS)

Moir, R.W.

2008-01-01

The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often-cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. Strong incentives for the molten salt reactor design are its good fuel utilization, good economics, amazing fuel flexibility and promised large benefits. It can: - use thorium or uranium; - be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have an epithermal neutron spectrum; - fission uranium isotopes and plutonium isotopes; - produces less long-lived wastes than today's reactors by a factor of 10-100; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 deg. C if carbon composites are successfully developed. Enhancing 232 U content in the uranium to over 500 ppm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR are enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high-enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/year base program for 10 years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/year over 20 years). A benefit of liquid fuel is that

SP-100 from ground demonstration to flight validation

International Nuclear Information System (INIS)

Buden, D.

1989-01-01

The SP-100 program is in the midst of developing and demonstrating the technology of a liquid-metal-cooled fast reactor using thermoelectric thermal-to-electric conversion devices for space power applications in the range of tens to hundreds of kilowatts. The current ground engineering system (GES) design and development phase will demonstrate the readiness of the technology building blocks and the system to proceed to flight system validation. This phase includes the demonstration of a 2.4-MW(thermal) reactor in the nuclear assembly test (NAT) and aerospace subsystem in the integrated assembly test (IAT). The next phase in the SP-100 development, now being planned, is to be a flight demonstration of the readiness of the technology to be incorporated into future military and civilian missions. This planning will answer questions concerning the logical progression of the GES to the flight validation experiment. Important issues in planning the orderly transition include answering the need to plan for a second reactor ground test, the method to be used to test the SP-100 for acceptance for flight, the need for the IAT prior to the flight-test configuration design, the efficient use of facilities for GES and the flight experiment, and whether the NAT should be modified based on flight experiment planning

Full-Scale Continuous Mini-Reactor Setup for Heterogeneous Grignard Alkylation of a Pharmaceutical Intermediate

DEFF Research Database (Denmark)

Pedersen, Michael Jønch; Holm, Thomas; Rahbek, Jesper P.

2013-01-01

A reactor setup consisting of two reactors in series has been implemented for a full-scale, heterogeneous Grignard alkylation. Solutions pass from a small filter reactor into a static mixer reactor with multiple side entries, thus combining continuous stirred tank reactor (CSTR) and plug flow...

Mitigation action plan for the 100-KR-4, 100-HR-3 pump and treat

International Nuclear Information System (INIS)

Weiss, S.G.

1996-10-01

This project involves drilling 22 wells, improving access roads to existing new wells, laying connecting pipes, and constructing groundwater treatment facilities in the 100-KR-4 and 100-HR-3 Operable Units. These facilities are located at the Hanford Site in Richland, Washington. The drilling is expected to be completed by October 1996, but the treatment operations will continue for approximately 10 years. Thirteen of the new wells are to be placed in the 100-K Area, five in the 100-H Area, and four in the 100-D Area. A 4 km (2.5 mi) pipeline will run from the wells at the 100-D Area to the treatment facility at the 100-H Area. The 116-K-2 Trench received reactor effluents from 1955 to 1971. It is 1,460 m long by 16.4 m wide by 5.5 m deep with spoil pits at the surface on both sides. Washouts occurred during operation, causing several surface contamination areas between the trench and river that were covered with a few feet of soil

Recherche en partenariat sur la transition nutritionnelle et les ...

International Development Research Centre (IDRC) Digital Library (Canada)

l'évolution du risque a suivi et a débouché sur une stratégie de prévention ... de recherche ont pu avoir un impact sur les politiques et pro- .... Deux études étaient donc entreprises ..... à la même thématique et de recruter par eux-mêmes les.

The direct conversion of heat into electricity in reactors

International Nuclear Information System (INIS)

Devin, B.; Bliaux, J.; Lesueur, R.

1964-01-01

The direct conversion of heat into electricity by thermionic emission in an atomic reactor has been studied with the triple aim of its utilisation: as an energy source for a space device, at the head of a conventional conversion system in power installations, or finally in association with the thermoelectric conversion in very low power installations. The laboratory experiments were mainly orientated towards the electron extraction of metals and compounds and their behaviour at high temperatures. Converters furnishing up to 50 amps at 0. 4 volts with an efficiency close to 10 p. 100 have been constructed in the laboratory; the emitters were heated by electron bombardment and were composed of tungsten covered with an uranium carbide deposit or molybdenum covered with cesium. The main aspects of the coupling between the converter and the reactor have been covered from the point of view of electronics: the influence of the mismatching of the load on the temperature of the emitter and the influence of thermal flux density on the temperature of the emitter and the stability of the converter. Converters using uranium carbide as the electron emitter have been tested in reactors. Tests have been made under dynamic conditions in order to determine the dynamic characteristics. The load matching curves have been constructed and the overall performances of several cells coupled in such a way as to form a reactor rod have been deduced. This information is fundamental to the design of a control system for a thermionic conversion reactor. The problems associated with the reliability of thermionic converters connected in series in the same reactor rod have been examined theoretically. Finally, the absorption isotherms have been drawn at the ambient temperatures for krypton and xenon on activated carbon with the aim of investigating the escape of fission products in a converter. (author) [fr

Renforcement de l'influence de la recherche sur l'élaboration des ...

International Development Research Centre (IDRC) Digital Library (Canada)

Réflexions

influent sur les politiques, mais de façons différentes. Le lien entre recherche et influence sur les politiques n'est ... Renforcement de l'influence de la recherche sur l'élaboration des politiques d'adaptation. PA. NOS. /G .... ceux de la CMAE ont dit avoir besoin de soutien technique. Il était malheureusement difficile pour le ...

Energy from nuclear reactors

International Nuclear Information System (INIS)

Hospe, J.

1977-01-01

This VDI-Nachrichten series has the target to provide a technical-objective basis for the discussion of the pros and cons of nuclear power. The first part deals with LWR-type reactors which so far have prevailed in nuclear power generation. (orig.) [de

Event tree analysis for the system of hybrid reactor

International Nuclear Information System (INIS)

Yang Yongwei; Qiu Lijian

1993-01-01

The application of probabilistic risk assessment for fusion-fission hybrid reactor is introduced. A hybrid reactor system has been analysed using event trees. According to the character of the conceptual design of Hefei Fusion-fission Experimental Hybrid Breeding Reactor, the probabilities of the event tree series induced by 4 typical initiating events were calculated. The results showed that the conceptual design is safe and reasonable. through this paper, the safety character of hybrid reactor system has been understood more deeply. Some suggestions valuable to safety design for hybrid reactor have been proposed

Transmutation of Tc-99 in fission reactors

International Nuclear Information System (INIS)

Kloosterman, J.L.; Li, J.M.

1994-12-01

Transmutation of Tc-99 in three different types of fission reactors is considered: A heavy water reactor, a fast reactor and a light water reactor. For the first type a CANDU reactor was chosen, for the second one the Superphenix reactor, and for the third one a PWR. The three most promising Tc-99 transmuters are the fast reactor with a moderated subassembly in the inner core, a fast reactor with a non-moderated subassembly in the inner core, and a heavy water reactor with Tc-99 target pins in the moderator between the fuel bundles. Transmutation half lives of 15 to 25 years can be achieved, with yearly transmuted Tc-99 masses of about 100 kg at a thermal reactor power of about 3000 MW. (orig.)

Handling and disposal of SP-100 ground test nuclear fuel and equipment

International Nuclear Information System (INIS)

Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

1990-05-01

The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radioactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

Handling and disposal of SP-100 ground test nuclear fuel and equipment

International Nuclear Information System (INIS)

Wilson, C.E.; Potter, J.D.; Hodgson, R.D.

1991-01-01

The post SP-100 reactor testing period will focus on defueling the reactor, packaging the various radiactive waste forms, and shipping this material to the appropriate locations. Remote-handling techniques will be developed to defuel the reactor. Packaging the spent fuel and activated reactor components is a challenge in itself. This paper presents an overview of the strategy, methods, and equipment that will be used during the closeout phase of nuclear testing

Testing and analyses of a high temperature duct for gas-cooled reactors

International Nuclear Information System (INIS)

Black, W.E.; Roberge, A.; Felten, P.; Bastien, D.

1979-01-01

A 0.6 scale model of a steam cycle gas-cooled reactor high temperature duct was tested in a closed loop helium facility. The object of the test series was to determine: 1) the thermal effects of gas permeation within the thermal barrier, 2) the plastic deformation of the metallic components, and 3) the thermal performance of the fibrous insulation. A series of tests was performed with thermal cyclings from 100 0 C to 760 0 C at 50 atmospheres until the system thermal performance had stabilized hence enabling predictions for the reactor life. Additional tests were made to assess permeation by deliberately simulating sealing weld failures thereby allowing gas flow by-pass within the primary thermal barrier. After 100 cycles the entire primary structure was found to have performed without structural failure. Due to high pressures exerted by the insulation on the cover plates and a design oversight, the thin seal sheets were unable to expand in an anticipated manner. Local buckling resulted. The insulation retained an acceptable degree of resiliency. However, some fiber damage was observed within both the high and low temperature insulation blankets. A thermal analysis was conducted to correlate the hot duct heat transfer results with those obtained from the analytical techniques used for the HTGR design using a computer thermal model representative of the duct and test setup. The thermal performance of the insulation, the temperature gradient through the structural components, the heating load to the cooling system and the permeation flow effect on heat transfer were verified. Exellent correlation between the experimental data and the analytical techniques were obtained

Components of SurA required for outer membrane biogenesis in uropathogenic Escherichia coli.

Directory of Open Access Journals (Sweden)

Kristin M Watts

2008-10-01

Full Text Available SurA is a periplasmic peptidyl-prolyl isomerase (PPIase and chaperone of Escherichia coli and other Gram-negative bacteria. In contrast to other PPIases, SurA appears to have a distinct role in chaperoning newly synthesized porins destined for insertion into the outer membrane. Previous studies have indicated that the chaperone activity of SurA rests in its "core module" (the N- plus C-terminal domains, based on in vivo envelope phenotypes and in vitro binding and protection of non-native substrates.In this study, we determined the components of SurA required for chaperone activity using in vivo phenotypes relevant to disease causation by uropathogenic E. coli (UPEC, namely membrane resistance to permeation by antimicrobials and maturation of the type 1 pilus usher FimD. FimD is a SurA-dependent, integral outer membrane protein through which heteropolymeric type 1 pili, which confer bladder epithelial binding and invasion capacity upon uropathogenic E. coli, are assembled and extruded. Consistent with prior results, the in vivo chaperone activity of SurA in UPEC rested primarily in the core module. However, the PPIase domains I and II were not expendable for wild-type resistance to novobiocin in broth culture. Steady-state levels of FimD were substantially restored in the UPEC surA mutant complemented with the SurA N- plus C-terminal domains. The addition of PPIase domain I augmented FimD maturation into the outer membrane, consistent with a model in which domain I enhances stability of and/or substrate binding by the core module.Our results confirm the core module of E. coli SurA as a potential target for novel anti-infective development.

Site specific health and safety plan, 100-HR-3 pump and treat. Revision 2

International Nuclear Information System (INIS)

St John, C.H.

1997-09-01

The 100-HR-3 Pump and Treat system is a groundwater remedial action to remove Hexavalent Chromium (Cr+6) from the groundwater underlying the 100-HR-3 Operable Unit. This plan covers operation, maintenance, repairs, resin exchange and equipment removal/installation. The 100-HR-3 Operable Unit addresses groundwater underlying the 100-D Area. The primary groundwater contaminant is Chromium +6. The chromium contamination resulted from the use of sodium dichromate during past reactor operations. Sodium dichromate was used to treat reactor coolant water during reactor operations. The purpose of this Pump and Treat system is to pump contaminated groundwater through above ground ion exchange resin and then return treated water to aquifer. Chromium levels extracted from the wells are anticipated to range in the low parts per billion (∼50 ppb) which is the drinking water limit for Cr+6

User's guide of DETRAS system-3. Description of the simulated reactor plant

International Nuclear Information System (INIS)

Yamaguchi, Yukichi

2006-12-01

DETRAS system is a PWR reactor simulator system for operation trainings whose distinguished feature is that it can be operated from the remote place of the simulator site. The document which is the third one of a series of three volumes of the user's guide of DETRAS, describes firstly an outline of the simulated reactor system then a user's interface needed for operation of the simulator of interest and finally a series of procedure for startup of the simulated reactor and shutdown of it from its rated operation state. (author)

BWR series pump recirculation system

International Nuclear Information System (INIS)

Dillmann, C.W.

1992-01-01

This patent describes a recirculation system for driving reactor coolant water contained in an annular downcomer defined between a boiling water reactor vessel and a reactor core spaced radially inwardly therefrom. It comprises a plurality of circumferentially spaced second pumps disposed in the downcomer, each including an inlet for receiving from the downcomer a portion of the coolant water as pump inlet flow, and an outlet for discharging the pump inlet flow pressurized in the second pump as pump outlet flow; and means for increasing pressure of the pump inlet flow at the pump inlet including a first pump disposed in series flow with the second pump for first receiving the pump inlet flow from the downcomer and discharging to the second pump inlet flow pressurized in the first pump

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

International Nuclear Information System (INIS)

Benedict, R.W.; Goff, K.M.

1993-01-01

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations

Visualization of Gas Distribution in a Model AP-XPS Reactor by PLIF: CO Oxidation over a Pd(100 Catalyst

Directory of Open Access Journals (Sweden)

Jianfeng Zhou

2017-01-01

Full Text Available In situ knowledge of the gas phase around a catalyst is essential to make an accurate correlation between the catalytic activity and surface structure in operando studies. Although ambient pressure X-ray photoelectron spectroscopy (AP-XPS can provide information on the gas phase as well as the surface structure of a working catalyst, the gas phase detected has not been spatially resolved to date, thus possibly making it ambiguous to interpret the AP-XPS spectra. In this work, planar laser-induced fluorescence (PLIF is used to visualize the CO2 distribution in a model AP-XPS reactor, during CO oxidation over a Pd(100 catalyst. The results show that the gas composition in the vicinity of the sample measured by PLIF is significantly different from that measured by a conventional mass spectrometer connected to a nozzle positioned just above the sample. In addition, the gas distribution above the catalytic sample has a strong dependence on the gas flow and total chamber pressure. The technique presented has the potential to increase our knowledge of the gas phase in AP-XPS, as well as to optimize the design and operating conditions of in situ AP-XPS reactors for catalysis studies.

SP-100 coated-particle fuel development. Phase I. Final report

International Nuclear Information System (INIS)

1983-03-01

This document is the final report of Phase I of the SP-100 Coated-Particle Fuel Development Program conducted by GA Technologies Inc. for the US Department of Energy under contract DE-AT03-82SF11690. The general objective of the study conducted between September and December 1982 was to evaluate coated-particle type fuel as an alternate or backup fuel to the UO 2 tile-and-fin arrangement currently incorporated into the reference design of the SP-100 reactor core. This report presents and discusses the following topics in the order listed: the need for an alternative fuel for the SP-100 nuclear reactor; an abbreviated description of the reference and coated-particle fuel module concepts; the bases and results of the study and analysis leading to the preliminary design of a coated particle suitable for the SP-100 space power reactor; incorporation of the fuel particles into compacts and heat-pipe-cooled modules; initial efforts and plans to fabricate coated-particle fuel and fuel compacts; the design and performance of the proposed alternative core relative that of the reference fuel; and a summary of critical issues and conclusions consistent with the level of effort and duration of the study

Les masques trompeurs de la bipolarité: étude de 100 cas | Nabih ...

African Journals Online (AJOL)

C'est une étude descriptive transversale portant sur 100 patients atteints de TB, inclus selon les critères du DSM V, qui ont été vus en consultation ou bien hospitalisés dans le service de psychiatrie de l'hôpital Militaire Avicenne de Marrakech, durant une période de deux ans. L'âge moyen des patients était de 29,5 ans ...

Overall aspects of control of ISIS-type nuclear reactor

International Nuclear Information System (INIS)

Amato, S.; Santinelli, A.

1996-01-01

The paper describes the main aspects related to the definition of main controls required to operate an ISIS-type nuclear power reactors. ISIS is a PWR-type intrinsically safe nuclear reactor designed by ANSALDO, based on density lock concept; it presents, between the other safety functions, self-depressurization and core cooling capability for unlimited time. Due to its specific characteristics, the ISIS reactor required to development of new control philosophy (if compared with actual nuclear power reactor) with the implementation of new control functions, for instance the density locks hot/cold interface locations control. This paper describes the main control functions implemented, their rationale, as well as the dynamic simulation performed to verify the adequacy of controls definitions. The dynamic simulations here described refers to a step-wise power ramp of 100-90-100 (% of nominal power) and to a power ramp of 100-50-100 with a slope of 5%/min; the results obtained have shown the ISIS capability to perform such operational transients, despite its innovative design was mainly focused on intrinsically safe behaviour. (author)

Nouvelles: KEK: B pour BELLE; NA48 mesure la violation directe de CP; LEL: le laser à électrons libres dépasse le kilowatt; Le LEP à 100 pour 100; Tour d'honneur au RHIC; Mesures de précision; Etude sur les modes de gestion du village planétaire; Conférence PANIC99 à Uppsala; Rencontre électron-noyau à HERA; LPP, Doubna célèbre une décennie; Les physiciens auscultent le photon; Aimants chauds venus du froid

CERN Multimedia

1999-01-01

Nouvelles: KEK: B pour BELLE; NA48 mesure la violation directe de CP; LEL: le laser à électrons libres dépasse le kilowatt; Le LEP à 100 pour 100; Tour d'honneur au RHIC; Mesures de précision; Etude sur les modes de gestion du village planétaire; Conférence PANIC99 à Uppsala; Rencontre électron-noyau à HERA; LPP, Doubna célèbre une décennie; Les physiciens auscultent le photon; Aimants chauds venus du froid

Nuclear research reactors in the world. June 1988 ed.

International Nuclear Information System (INIS)

1988-01-01

This is the third edition of Reference Data Series No. 3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of June 1988, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the IAEA Research Reactor Data Base (RRDB) system. This system contains all the information and data previously published in the Agency's publication Power and Research Reactors in Member States as well as additional information. 12 figs, 19 tabs

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

International Nuclear Information System (INIS)

1967-01-01

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

Definitive design status of the SP-100 Ground Engineering System Test Site

International Nuclear Information System (INIS)

Renkey, E.J. Jr.; Bazinet, G.D.; Bitten, E.J.; Brackenbury, P.J.; Carlson, W.F.; Irwin, J.J.; Edwards, P.A.; Shen, E.J.; Titzler, P.A.

1989-05-01

The SP-100 reactor will be ground tested at the SP-100 Ground Engineering System (GES) Test Site on the US Department of Energy (DOE) Hanford Site near Richland, Washington. Project direction and the flight system design evolution have resulted in a smaller reactor size and the consequential revision to Test Site features to accommodate the design changes and reduce Test Site costs. The significant design events since the completion of the Conceptual Design are discussed in this paper

Definitive design status of the SP-100 Ground Engineering System Test Site

Energy Technology Data Exchange (ETDEWEB)

Renkey, E.J. Jr.; Bazinet, G.D.; Bitten, E.J.; Brackenbury, P.J.; Carlson, W.F.; Irwin, J.J.; Edwards, P.A.; Shen, E.J.; Titzler, P.A.

1989-05-01

The SP-100 reactor will be ground tested at the SP-100 Ground Engineering System (GES) Test Site on the US Department of Energy (DOE) Hanford Site near Richland, Washington. Project direction and the flight system design evolution have resulted in a smaller reactor size and the consequential revision to Test Site features to accommodate the design changes and reduce Test Site costs. The significant design events since the completion of the Conceptual Design are discussed in this paper.

Nuclear research reactors in the world. May 1987 ed.

International Nuclear Information System (INIS)

1987-01-01

This is the second edition of Reference Data Series No.3, Nuclear Research Reactors in the World, which replaces the Agency's publications Power and Research Reactors in Member States and Research Reactors in Member States. This booklet contains general information, as of the end of May 1987, on research reactors in operation, under construction, planned, and shut down. The information is collected by the Agency through questionnaires sent to the Member States through the designated national correspondents. 11 figs, 19 tabs

Espacio herido, tiempo acelerado, imaginario conmovido. Interrogantes sobre la postmodernidad del Sur.

Directory of Open Access Journals (Sweden)

Mabel Franzone

2010-11-01

Full Text Available Estableciendo una relación estrecha entre espacio, tiempo e imaginario, este escrito intenta tratar la gran separación Norte- Sur, acaecida con la hipermodernidad y sus consecuencias sobre la destrucción del espacio del Sur ; la cuestión de fondo encara las huellas que puede dejar en los cuerpos y en los imaginarios. Los límites impuestos se tratan en cuatro niveles : históricos, geográficos, económicos- sociales y de la representación. Para una crítica de la modernidad se han tomado parámetros de pensamiento que operan en una trama mediana, forma de pensar propia al Sur y que muestran la necesidad de una epistemología también propia al Sur, que refleje las cuestiones fundamentales y que tenga en cuenta la emancipación y la solidaridad como categorías para repensar el Sur.

Estilos estructurales del Subandino Sur de Bolivia

OpenAIRE

Rocha, Emilio

2013-01-01

El Subandino Sur de Bolivia es una típica faja plegada y corrida de lámina delgada, con una notable regularidad en la geometría de las estructuras. Esta es una característica típica de las fajas plegadas en las que no se involucra el basamento en la deformación. Sin embargo, cuando se analiza en detalle la geometría y evolución de la deformación del Subandino Sur, se verifica que existen numerosas desviaciones de dicha regularidad. En el presente trabajo se estudiaron los diferentes procesos ...

Safety features and licensing of CNNC-ACP100

Energy Technology Data Exchange (ETDEWEB)

Zhong, F., E-mail: Zhongfj2000@163.com [Nuclear Power Inst. of China, National Key Lab. of Science and Technology on Reactor System Design Technology (China)

2014-07-01

ACP100 is an innovatory modular pressurized water reactor, the engineering safety systems fully adopt passive safety design technology. Its inherent safety and passive features/systems are verified via testing facilities and are highlighted at certain levels of defence in depth. The licensing of ACP 100 is within current LWR framework and meets up-to-date codes and requirements in nuclear safety. (author)

¿Cuáles son las Razones Subyacentes al Éxito Educativo de Corea del Sur?

OpenAIRE

García Ruiz, María José; Arechavaleta Pintó, Carmen

2011-01-01

El éxito educativo de los alumnos de Corea del Sur debe gran parte de su mérito a toda una serie de estructuras, presiones, valores y objetivos que están presentes en la cultura escolar y en la sociedad coreana, más que sólo a virtudes intrínsecas de las escuelas de este país. Las altas puntuaciones de los alumnos coreanos en las diversas evaluaciones internacionales de la educación pueden ser atribuidas, de modo especial, a dos aspectos que constituyen un legado de la ética confuciana en est...

Nuclear power reactors in the world. Apr 1985 ed.

International Nuclear Information System (INIS)

1985-01-01

This is the fifth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which replaces the Agency's publication Power Reactors in Member States. This bulletin contains the following summarized information on nuclear power reactors in the world: General information as of the end of 1984 on reactors operating or under construction and such additional information on planned and shutdown reactors as is available; Performance data on major reactor types operating in the Agency's Member States. The information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of and operating experience with reactors

Le lancement canadien du Rapport sur les politiques alimentaires ...

International Development Research Centre (IDRC) Digital Library (Canada)

1 mai 2018 ... Cet événement sera filmé, puis accessible en ligne sur la chaîne YouTube du CRDI. Le Rapport sur les politiques alimentaires mondiales 2018 passe en revue les principaux faits nouveaux et événements en matière de politiques alimentaires survenus au cours de l'année. D'éminents chercheurs ...

Analyse technico–économique des Aliments densifies sur les ...

African Journals Online (AJOL)

Une étude portant sur l'influence des aliments densifiés sur la performance de croissance des boucs roux de Maradi a été conduite entre juillet 2016 à septembre 2016. Quarante deux (42) boucs ont été répartis au hasard en six lots de sept individus. Chaque lot correspond à un aliment ou traitement préparé à partir des.

Public information circular for shipments of irradiated reactor fuel

International Nuclear Information System (INIS)

1982-06-01

This publication is the third in a proposed series of annual publications issued by the Nuclear Regulatory Commission in response to public information requests regarding the Commission's regulation of shipments of irradiated reactor fuel. Subsequent issues in this series will update the information contained herein. This publication contains basically three kinds of information: (1) routes approved by the Commission for the shipment of irradiated reactor fuel, (2) information regarding any safeguards-significant incidents which have been reported to occur during shipments along such routes, and (3) cumulative amounts of material shipped

Case series

African Journals Online (AJOL)

abp

14 déc. 2015 ... s'agissait d'une étude prospective de base réalisée au Burkina Faso portant sur la chirurgie de la cataracte par la technique de tunnélisation. L'âge moyen de nos ... charge sur place ou d'une référence vers le service d'ophtalmologie du CHR .... trou sténopéique selon les critères de l'OMS. Ces résultats ...

Concilier plaisir et nutrition. Travaux des groupes de travail PNNS sur les lipides et sur le goût

OpenAIRE

Souliac , L.; Bizet, G.; Remy, S.

2010-01-01

Les ministères chargés de l'alimentation et de la santé ont incité les opérateurs à mettre sur le marché des aliments moins riches en lipides, par la signature avec l'État de chartes d'engagement sur la composition nutritionnelle. L'observatoire de la qualité de l'alimentation suit l'évolution de ces efforts. Les lipides ont un rôle technologique et sensoriel important qu'il convient de prendre en considération. Il est aujourd'hui possible d'améliorer le profil nutritionnel des aliments, tout...

Données préliminaires sur le paludisme humain en zones rurale et ...

African Journals Online (AJOL)

SARAH

31 juil. 2017 ... axée sur le diagnostic précoce et la prise en charge rapide des ... recherche investigue sur les connaissances, attitudes et ... paludisme et (vi) sur la source de traitement du paludisme. ... technique ELISA (Burkot et al., 1984).

Development of telerobotic manipulators for reactor dismantling work

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

This paper describes the amphibious electrical manipulators JARM-10, JART-25, JART-100 and JARM-25 which were developed in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. They are multi-functional telerobotic light-duty (10 and 25 daN) and heavy-duty (100 daN) Manipulators which can be used in hostile environments in reactor dismantling work such as high radiation, underwater work and electrical noise. Each manipulator can be operated in either a bilateral master-slave, a teach-and-playback or a programmed control mode. By combining these modes appropriately, it is possible to perform complex tasks of remote handling. The usefulness of the telerobotic systems for dismantling nuclear reactors has been demonstrated by successful application of the JARM-25 for remote underwater dismantlement of highly radioactive reactor internals of complex form of an experimental nuclear power reactor. (author)

Progress in SP-100 tribological coatings

International Nuclear Information System (INIS)

Ring, P.J.; Roy, P.; Schuster, G.B.; Busboom, H.J.

1992-01-01

The SP-100 reactor will operate at temperatures up to 1500K in high vacuum. To address the SP-100 needs, a tribology development program has been established at GE to investigate candidate coating materials. Materials were selected based on their high thermodynamic stability, high melting point, compatibility with the substrate, and coefficients of thermal expansion similar to niobium-1% zirconium-the candidate structural material for SP-100. An additional requirement was that the deposition processes should be commercially available to coat large components. This paper presents the details regarding the SP-100 Tribology Development Program including background information, specific bearing requirements, basis for coating material selection, testing methods and the initial results covering the early years of this program

Enquêtes sur les soucoupes volantes

OpenAIRE

Lagrange, Pierre

2007-01-01

La dispute autour du mot « preuve » se réduit à une question : Qu'est-ce qui constitue une preuve ? Faut-il qu'un ovni atterrisse à l'entrée du Pentagone, près des bureaux des chefs d'État-Major ? Ou bien est-ce une preuve quand une station de radar au sol détecte un ovni, envoie un jet l'intercepter, que le pilote du jet le voit et le suit sur son radar pour finir par le voir disparaître à une vitesse phénoménale ? Ou est-ce une preuve quand un pilote tire sur un ovni et maintient son histoi...

Summary report for 1990 inservice inspection (ISI) of SRS 100-K reactor tank

International Nuclear Information System (INIS)

Morrison, J.M.; Loibl, M.W.

1990-01-01

The integrity of the SRS reactor tanks is a key factor affecting their suitability for continued service since, unlike the external piping system and components, the tanks are virtually irreplaceable. Cracking in various areas of the process water piping systems has occurred beginning in about 1960 as a result of several degradation mechanisms, chiefly intergranular stress corrosion cracking (IGSCC) and chloride-induced transgranular cracking. The purpose of this inspection was to determine if selected welds in the K Reactor tank wall contained any indications of IGSCC. These portions included areas in and beyond the weld HAZ, extending out as far as two to three inches from the centerline of the welds, plus selected areas of base metal at the intersection of the main tank vertical and mid-girth welds. No evidence of such degradation was found in any of the areas examined. This inspection comprised approximately 60% of the accessible weld length in the K Reactor tank. Initial setup of the tank, which prior to inspection contained Mark 60B target assemblies but no Mark 22 fuel assemblies, began on January 14, 1990. The inspection was completed on March 9, 1990

Experimental methods of reactor physics; Methodes experimentales de physique des reacteurs a neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Breton, D; Lafore, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

This paper is a synthesis of various experimental methods in use with the reactors of the Commissariat a l'Energie Atomique. The main techniques used are mentioned and the difficulties encountered and the accuracy obtained are particularly dwelt upon. The application of these various methods to reactors in order to obtain specific results is also indicated. This paper consists of five parts. I - General methods. Macroscopic and microscopic flux distribution (anisotropy effect), power distribution, etc... II - Kinetic measurements a) pulsed neutron technique: apparatus and accuracy; application to {lambda}t and to anti reactivity measurements; application to graphite, light water and beryllium oxide. b) oscillation techniques: equipment and accuracy; application to the measurements of effective cross sections and resonance integrals. c) fluctuations: apparatus and technique of measurement. III - Poison methods. Description of methods for introducing and extracting the poison, difficulties encountered with light and heavy water, measurement of temperature coefficients and anti-reactivity. IV - Spectra measurements. Choice and development of foils, problems of measurement, application to spectral measurements for thermalization studies, application to dosimetry. V - Experimental shielding measurements. The technique and apparatus recently developed in this field are presented. (authors) [French] Cette communication fait une synthese des differentes methodes experimentales mises en oeuvre sur les reacteurs du CEA. Elle presente les principales techniques utilisees et insiste plus particulierement sur les difficultes rencontrees et la precision obtenue; elle indique egalement l'application de ces differentes methodes sur les reacteurs, en vue de l'obtention des resultats determines. Elle comporte cinq parties: I - METHODES GENERALES: Distribution de flux macroscopique et microscopique (effet d'anisotropie), distribution de puissance, etc... II - MESURES CINETIQUES: a

New modelling method for fast reactor neutronic behaviours analysis; Nouvelles methodes de modelisation neutronique des reacteurs rapides de quatrieme Generation

Energy Technology Data Exchange (ETDEWEB)

Jacquet, P.

2011-05-23

Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author) [French] Les criteres de surete qui regissent le developpement de coeurs de reacteurs de quatrieme generation implique l'usage d'outils de calcul neutronique performants. Une premiere partie de la these reprend toutes les etapes de modelisation neutronique des reacteurs rapides actuellement d'usage dans le code de reference ECCO. La capacite des modeles a decrire le phenomene d'autoprotection, a representer les fuites neutroniques au niveau d'un reseau d'assemblages combustibles et a generer des sections macroscopiques representatives est appreciee sur le domaine des reacteurs rapides innovants respectant les criteres de quatrieme generation. La deuxieme partie de ce memoire se consacre a la modelisation des coeurs rapides avec reflecteur acier. Ces derniers necessitent le developpement de methodes avancees de condensation et d'homogenisation. Plusieurs methodes sont proposees et confrontees sur un probleme de modelisation typique: le coeur ZONA2B du reacteur maquette MASURCA

Hausse des taxes sur le tabac dans trois pays d'Amérique centrale ...

International Development Research Centre (IDRC) Digital Library (Canada)

Les chercheurs mèneront des études sur la demande de produits du tabac dans le ... sur les techniques modernes de contrôle de la chaîne d'approvisionnement, ... IWRA/CRDI sur les changements climatiques et la gestion adaptive de l'eau.

Energy memento; Memento sur l'energie

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-07-01

This memento about energy provides a series of tables with numerical data relative to energy resources and uses in France, in the European Union and in the rest of the world: energy consumption (primary energy, forecasting, CO{sub 2} emissions, energy independence, supplies, uses and imports, demand scenarios, energy savings..), power production (production, forecasting, loads, consumption, hydro-power, thermal equipment, exports), nuclear power (production, forecasting, reactors population, characteristics of French PWRs, uranium needs and fuel cycle), energy resources (renewable energies, fossil fuels and uranium reserves and production), economic data (gross national product, economic and energy indicators, prices and cost estimations), energy units and conversion factors (counting, calorific value of coals, production costs, energy units). (J.S.)

The action of the project coordinator with respect to reactor safety

International Nuclear Information System (INIS)

Leclercq, Jacques

1981-01-01

Before describing the various actions of the project coordinator (EDF) entrusted with the building of nuclear power stations, with respect to reactor safety in France, the definition of reactor safety and the various participants are mentioned first. These participants are: the Government Departments and the Experts involved (the Department of Nuclear Safety of the 'Institut de Protection et de Surete Nucleaire' forming the first technical support) and the applicant, namely the EDF. The reactor safety actions of the project coordinator are defined as from the following components: 1 - The targets laid down with respect to safety, the final objective being the protection of workers and the public against the potential dangers of the installations, principally against radiation. 2 - The safety methodology at the design stage of the power station: 'barrier' method, defence method in depth at three levels, lines of assurance method, and probabilistic method. 3 - Safety actions at the construction stage within the context of an assurance of quality programe. 4 - Safety at the trials, commissioning and operating stage, with the backing of the 'Groupe Operationnel de Demarrage (G.O.D.)' and the 'Commission d'Essais sur Site (C.E.S.)'. An initial balance sheet of the reactor safety actions for the PWR units built by the EDF is presented [fr

Power Reactor Design at Zero Power; Etudes de Reacteurs de Puissance, au Moyen de Machines de Puissance Zero; Konstruktsiya ehnergeticheskogo reaktora nulevoj moshchnosti; Diseno de Reactores Generadores con Ayuda de Reactores de Potencia Nula

Energy Technology Data Exchange (ETDEWEB)

Redman, W. C.; Plumlee, K. E.; Baird, Q. L. [Argonne National Laboratory, Argonne, IL (United States)

1964-02-15

reliance placed in the past on exponential and critical systems for fulfilling Argonne's responsibilities in reactor development. An indication of their future role is provided by a brief summary of the current and planned programmes for the existing members of, and anticipated additions to, Argonne's family of operating zero-power reactors. (author) [French] Avec le reacteur de puissance zero du Laboratoire national d'Argonne, on a procede a des etudes de reacteurs tres divers; reacteurs de recherche, generatrices nucleaires, reacteurs pour la propulsion, pour la production de radioisotopes et reacteurs experimentaux; les ensembles associes - exponentiels et critiques non empoisonnes - ont fourni les donnees debase. Afin de rendre compte d'experiences recentes et de montrer quelle masse de renseignements sur la physique des reacteurs on peut obtenir avec des systemes a bas flux, les auteurs exposent les programmes experimentaux ci-apres: 1. Etude des proprietes des elements combustibles en oxydes d'uranium et de thorium, immerges dans l'eau lourde, en s'attachant particulierement aux donnees necessaires pour l'etude d'un deuxieme coeur pour le reacteur experimental a eau bouillante du Laboratoire d'Argonne; 2. Maquette d'un reacteur de recherche a haut flux, qui permettra de verifier les calculs faits au cours de l'etude, de determiner la geometrie optimale et d'estimer l'effet du taux de combustion; 3. Determination des repartitions energetiques et de l'effet de l'immersion des cartouches sur la reactivite pour un reacteur experimental a ebullition et a surchauffe combinees; 4. Etude d'un coeur de reacteur surgenerateur plutonigene a neutrons rapides, alimente en U{sup 235} et refroidi au sodium qui constituerait la charge initiale du Deuxieme reacteur surgenerateur experimental d'Argonne; 5. Etude des caracteristiques d'un reacteur a deux regions, l'une thermique et l'autre rapide, en interaction. Dans l'expose de ces programmes, les auteurs expliquent pourquoi on a

Effect of the plutonium isotopic composition on the performance of fast reactors; Effet de la composition isotopique du plutonium sur le rendement de reacteurs a neutrons rapides; Vliyanie izotopnogo sostava plutoniya na rabotu reaktorov na bystrykh nejtronakh; Efectos de la composicion isotopica del plutonio sobre el funcionamiento de los reactores rapidos

Energy Technology Data Exchange (ETDEWEB)

Yiftah, S [Israel Atomic Energy Commission (Israel)

1962-03-15

The isotopic composition of plutonium to be used as fuel for fast reactors will depend on the source of plutonium. In principle three different sources are possible: (a) production reactors; (6) thermal power reactors (using natural uranium or enriched uranium as fuel); (c) fast reactor blankets. In general, source (a) and to some extent source (c) will provide relatively 'clean' plutonium, that is mostly Pu{sup 239}, while plutonium from source (6) will be 'dirty' plutonium, that is plutonium rich in Pu{sup 240}, Pu{sup 241}, and Pu{sup 242}. The degree of 'dirtiness' will depend on the kind of reactor, amount of burn-up and in general on the irradiation history of the fuel. The question then arises, can one use as fuel for fast reactors any kind of plutonium? To investigate the effect of different isotopic composition of the plutonium fuel, in the metallic, oxide and carbide form, on the performance of fast reactors, a limited series of spherical geometry 16-group diffusion theory calculations were performed, using the 16-group cross-section set developed recently by Yiftah, Okrent and Moldauer and taking three different kinds of plutonium, starting with pure Pu{sup 239} and increasing the amount of higher isotopes. For the systems studied-800, 1500 and 2500-l core-volumes, which are typical for large fast power reactors-the result is, when one takes into account only the thermally fissionable isotopes Pu{sup 239} arid Pu{sup 241}, that the 'dirtier' the plutonium, the smaller the critical mass and the higher the breeding ratio. For the 1500-l reactor, taken as an example, it is further found that in the metallic, oxide and carbide plutonium fuels the reactivity change upon removal of 40% of the sodium initially present in the core is made more negative (or less positive) when the plutonium is richer in higher isotopes. (author) [French] La composition isotopique du plutonium qui doit etre utilise comme combustible dans des reacteurs a neutrons rapides depend de

Small size modular fast reactors in large scale nuclear power

International Nuclear Information System (INIS)

Zrodnikov, A.V.; Toshinsky, G.I.; Komlev, O.G.; Dragunov, U.G.; Stepanov, V.S.; Klimov, N.N.; Kopytov, I.I.; Krushelnitsky, V.N.

2005-01-01

The report presents an innovative nuclear power technology (NPT) based on usage of modular type fast reactors (FR) (SVBR-75/100) with heavy liquid metal coolant (HLMC) i. e. eutectic lead-bismuth alloy mastered for Russian nuclear submarines' (NS) reactors. Use of this NPT makes it possible to eliminate a conflict between safety and economic requirements peculiar to the traditional reactors. Physical features of FRs, an integral design of the reactor and its small power (100 MWe), as well as natural properties of lead-bismuth coolant assured realization of the inherent safety properties. This made it possible to eliminate a lot of safety systems necessary for the reactor installations (RI) of operating NPPs and to design the modular NPP which technical and economical parameters are competitive not only with those of the NPP based on light water reactors (LWR) but with those of the steam-gas electric power plant. Multipurpose usage of transportable reactor modules SVBR-75/100 of entirely factory manufacture assures their production in large quantities that reduces their fabrication costs. The proposed NPT provides economically expedient change over to the closed nuclear fuel cycle (NFC). When the uranium-plutonium fuel is used, the breeding ratio is over one. Use of proposed NPT makes it possible to considerably increase the investment attractiveness of nuclear power (NP) with fast neutron reactors even today at low costs of natural uranium. (authors)

Une nouvelle ouverture sur le monde pour les femmes | CRDI ...

International Development Research Centre (IDRC) Digital Library (Canada)

'appellent ailleurs dans le dédale d'étals et d'allées serrés. En plus de son travail sur le marché, Jertrudes est également secrétaire de l'association des femmes qui vendent sur ce marché. Le groupe est un endroit où elles ...

Thermohydraulic analysis for power increase of IEAR-1 reactor

International Nuclear Information System (INIS)

Umbehaun, Pedro E.; Bastos, Jose L.F.

1996-01-01

In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

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.

Mo-99 production on a LEU solution reactor

International Nuclear Information System (INIS)

Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.

2005-01-01

A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

Influence de l'irrigation sur le rendement et sur la qualité des raisins

Directory of Open Access Journals (Sweden)

J. Cerny

1968-12-01

 Le travail présenté résume les résultats de l'expérience avec l'irrigation supplémentaire de la vigne dans le domaine sec du Sud de la Moravie ainsi que son influence sur le rendement et la qualité des raisins.

Iris reactor conceptual design

International Nuclear Information System (INIS)

Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V.; Galvin, M.; Todreas, N.E.; Lombardi, C.V.; Maldari, F.; Ricotti, M.E.; Cinotti, L.

2001-01-01

IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

Joint Oil Analysis Program Spectrometer Standards SCP Science (Conostan) Qualification Report for D19-0, D3-100, and D12-XXX Series Standards

Science.gov (United States)

2015-05-20

Joint Oil Analysis Program Spectrometer Standards SCP Science (Conostan) Qualification Report For D19-0, D3-100, and D12- XXX Series Standards NF...Candidate Type D19-0 ICP-AES Results ..................................................................... 4 Table V. Candidate Type D12- XXX ...Physical Property Results .................................................. 5 Table VI. Candidate Type D12- XXX Rotrode-AES Results

A systems analysis of the ARIES tokamak reactors

International Nuclear Information System (INIS)

Bathke, C.G.

1992-01-01

The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor

Nuclear energy center site survey reactor plant considerations

International Nuclear Information System (INIS)

1976-05-01

The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers

Transfer function synthesis for reactor spatial dynamics using the modal approach

Energy Technology Data Exchange (ETDEWEB)

Guppy, C B [Control and Instrumentation Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1962-08-15

Techniques are developed below which will enable the construction of transfer functions relating changes in variables such as power or neutron flux with reactivity perturbations when there is a need for taking into account spatial effects within a reactor. Initially each of the transfer functions derived comprises the sum of a series of harmonics each of which has a laplace transform with associated spatial eigenfunction. Series of this kind can then be reduced to pure polynomial form (numerators on denominators) the coefficients of which have implicit allowance for spatial effects. The existence of large reactors having several independent controllers make necessary knowledge of transfer functions of this form. The technique will allow the characteristics of each controlled sector to be obtained as well as the characteristics of the complete control system with its couplings through the reactor core. In addition, the developing use of frequency response testing of reactors makes necessary a knowledge of the spatial behaviour to be expected of a reactor under test. (author)

Analysis of boiling water reactors capacities for the 100% MOX fuel recycling

International Nuclear Information System (INIS)

Knoche, Dietrich

1999-01-01

The electro-nuclear park exploitation leads to plutonium production. The plutonium recycling in boiling water reactors performs a use possibility. The difference between the neutronic characteristics of the uranium and the plutonium need to evaluate the substitution impact of UOX fuel by MOX fuel on the reactor operating and safety. The analysis of the main points reached to the following conclusions: the reactivity coefficients are negative, during a cooling accident the re-divergence depends on the isotopic vector of the used plutonium, the efficiency lost of control cross resulting from the plutonium utilization can be compensate by the increase of the B 4C enrichment by 10 B and the change of the steel structure by an hafnium structure, the reactivity control in evolution can be obtained by the fuel poisoning (gadolinium, erbium) and the power map control by the plutonium content monitoring. (A.L.B.)

Future view of total energy system and reactor engineering and reactor physics

International Nuclear Information System (INIS)

Ozawa, T.

1974-01-01

This paper outlines the present status of fission reactors and fusion reactors. The conversion ratio of light water reactors is 0.5, and the efficiency is 32% because of relatively low temperature. Both pressurized water reactors and boiling water reactors are technically well developed, their performances are well known, and the fuel cycle is well developed, so that both reactors have monopolized power reactor market. But the reprocessing of spent fuel and the treatment of their hazards are inevitable, and the construction and enlargement of reprocessing facilities are indispensable. In LMFBR's tight sealing is easy because they are non-pressurized, and the efficiency is 41%. But liquid sodium is strongly activated and recirculated, so that chemical obstruction due to the breakage of recirculating pumps, pipings, and heat exchangers may occur, and the hazard of plutonium is large. Regarding controlled thermo-nuclear fusion reactors, because Lawson criterion must be satisfied, two methods of plasma confinement are now experimented. One is the plasma confinement by strong magnetic field of 50 KG to 100 KG, and the other is the confinement by the implosion method with high-power laser beam. The latter has much more uncertainties than the former, but recently both methods have made much progress. (Tai, I)

Preliminary Study for Conceptual Design of Advanced Long Life Small Modular Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Kim, T. K. [Argonne National Laboratory, Argonne (United States)

2015-05-15

As one of the non-water coolant Small-Modular Reactor (SMR) core concepts for use in the mid- to long-term, ANL has proposed a 100 MWe Advanced sodium-cooled Fast Reactor core concept (AFR-100) targeting a small grid, transportable from pre-licensed factories to the remote plant site for affordable supply. Various breed-and-burn core concepts have been proposed to extend the reactor cycle length, which includes CANDLE with a cigar-type depletion strategy, TerraPower reactors with fuel shuffling for effective breeding, et al. UNIST has also proposed an ultra-long cycle fast reactor (UCFR) core concept having the power rating of 1000 MWe. By adopting the breed-and-burn strategies, the UCFR core can maintain criticality for a targeting reactor lifetime of 60 years without refueling. The objective of this project is to develop an advanced long-life SMR core concept by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. A conceptual design of long life small modular fast reactor is under development by adopting both the small modular design features of the AFR-100 and the long-life breed-and-burn concept of the UCFR. The feasibility of the long-life fast reactor concepts was reviewed to obtain the core design guidelines and the reactor design requirements of long life small modular fast reactor were proposed in this study.

Effects of hydrogen on the tensile strength characteristics of stainless steels; Effets de l'hydrogene sur les caracteristiques de rupture par traction d'aciers inoxydables

Energy Technology Data Exchange (ETDEWEB)

Blanchard, R; Pelissier, J; Pluchery, M [Commissariat a l' Energie Atomique, Grenoble (France).Centre d' Etudes Nucleaires; Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1961-07-01

This paper deals with the effects of hydrogen on stainless steel, that might possibly be used as a canning material in hydrogen-cooled reactors. Apparent ultimate-tensile strength is only 80 per cent of initial value for hydrogen content about 50 cc NTP/ 100 g, and reduction in area decreases from 80 to 55 per cent. A special two-stage replica technique has been developed which allows fracture surface of small tensile specimens (about 0.1 mm diam.) to be examined in an electron microscope. All the specimens showed evidence of ductile character throughout the range of hydrogen contents investigated, but the aspect of the fracture surfaces gradually changes with increasing amounts. (author) [French] On etudie les effets de l'hydrogene sur des aciers inoxydables, qui sont des materiaux de gainage possibles pour des reacteurs utilisant l'hydrogene comme gaz de refroidissement. On montre que la charge apparente de rupture a la traction n'est plus que 80 pour cent de sa valeur initiale lorsque la teneur en hydrogene atteint 50 cc TPN/ 100 g, et que la striction passe dans ces conditions de 80 a 55 pour cent. L'examen microfractographique qui a ete effectue avec succes par une technique de double replique malgre la petitesse des echantillons (0,3 mm de diametre environ), revele que tout en gardant un caractere ductile, l'aspect des surfaces de rupture evolue notablement avec la teneur en hydrogene. (auteur)

Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

Field-reversed mirror reactor

International Nuclear Information System (INIS)

Carlson, G.A.

1978-01-01

The reactor design is a multicell arrangement wherein a series of field-reversed plasma layers are arranged along the axis of a long superconducting solenoid which provides the background magnetic field. Normal copper mirror coils and Ioffe bars placed at the first wall radius provide shallow axial and radial magnetic wells for each plasma layer. Each of 11 plasma layers requires the injection of 3.6 MW of 200 keV deuterium and tritium and produces 20 MW of fusion power. The reactor has a net electric output of 74 MWe and an estimated direct capital cost of $1200/kWe

First principles study of NH3 molecular adsorption on LiH (100) surfaces

International Nuclear Information System (INIS)

Lu Xiaoxia; Chen Yuhong; Dong Xiao

2012-01-01

The adsorption of NH 3 on LiH (100) crystal surfaces was studied by first principles method. The preferred adsorption sites, adsorption energy, dissociation energy and electronic structure of the LiH (100)/NH 3 systems were calculated separately. It is found that chemical adsorption happened mainly when NH 3 molecules are on the LiH (100) crystal surfaces. When NH 3 is adsorbed on the Li top site, NH 2 is formed on the LiH (100) crystal surfaces after loss of H atom, the calculated adsorption energy, 0.511 eV, belongs to strong chemical adsorption, then the interaction is strongest. The interaction between NH 2 and the neighboring Li, H are ionic. The covalent bonds are formed between N and H atoms in NH 2 . One H 2 molecule is formed by another H atom in NH 3 and H atom from LiH (100) crystal sur- faces. The covalent bonds are formed between H and H atoms in H 2 . (authors)

[Rapid startup and nitrogen removal characteristic of anaerobic ammonium oxidation reactor in packed bed biofilm reactor with suspended carrier].

Science.gov (United States)

Chen, Sheng; Sun, De-zhi; Yu, Guang-lu

2010-03-01

Packed bed biofilm reactor with suspended carrier was used to cultivate ANAMMOX bacteria with sludge inoculums from WWTP secondary settler. The startup of ANAMMOX reactor was comparatively studied using high nitrogen loading method and low nitrogen loading method with aerobically biofilmed on the carrier, and the nitrogen removal characteristic was further investigated. The results showed that the reactor could be started up successfully within 90 days using low nitrogen loading method, the removal efficiencies of ammonium and nitrite were nearly 100% and the TN removal efficiencywas over 75% , however, the high nitrogen loading method was proved unsuccessfully for startup of ANAMMOX reactor probably because of the inhibition effect of high concentration of ammonium and nitrite. The pH value of effluent was slightly higher than the influent and the pH value can be used as an indicator for the process of ANAMMOX reaction. The packed bed ANAMMOX reactor with suspended carrier showed good characteristics of high nitrogen loading and high removal efficiency, 100% of removal efficiency could be achieved when the influent ammonium and nitrite concentration was lower than 800 mg/L.

Nuclear Power Reactors in the World. 2013 Ed

International Nuclear Information System (INIS)

2013-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-third edition of Reference Data Series No. 2 provides a detailed comparison of various statistics through 31 December 2012. The tables and figures contain the following information: - General statistics on nuclear reactors in IAEA Member States; - Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; - Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA's Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects data through designated national correspondents in Member States

Chaire de recherche trilatérale Canada – Afrique du Sud sur les ...

International Development Research Centre (IDRC) Digital Library (Canada)

Le présent projet met l'accent sur les interactions entre l'humain et la faune, notamment sur les pauvres des zones rurales en pays tropical et sur l'effet des changements climatiques que ces derniers subiront. Puisant dans la recherche menée à la station de biologie expérimentale de l'Université Makere, près du parc ...

General Aspects of CAREM-25 Reactor

International Nuclear Information System (INIS)

Delmastro, Dario; Gomez, S.; Mazzi, R.; Gomez de Soler, S.; Santecchia, A.; Ishida, V.

2000-01-01

CAREM project consists on the development and design of an advanced Nuclear Power Plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100MWth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: integrated primary system, primary system cooling by natural convection, selfpressurization, and passive safety systems

Nuclear reactor

International Nuclear Information System (INIS)

Jungmann, A.

1975-01-01

Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

Rapport annuel au Parlement Loi sur la protection des ...

International Development Research Centre (IDRC) Digital Library (Canada)

Larocque, Samia

1 juil. 1983 ... TBS/SCT 350-63 (Rév. 2011/03). 31/03/2014. 01/04/2013. Période visée par le rapport : Reçues pendant la période visée par le rapport. Total. En suspens à la fin de la période de rapport précédente. Rapport statistique sur la Loi sur la protection des renseignements personnels. Nom de l'institution :.

Health physics aspects of advanced reactor licensing reviews

International Nuclear Information System (INIS)

Hinson, C.S.

1995-01-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on open-quotes next-generationclose quotes reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These open-quotes next-generationclose quotes reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four open-quotes next-generationclose quotes reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four open-quotes next-generationclose quotes reactor designs currently being reviewed by the NRC

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1988-05-01

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1988-05-01

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm 3 was by then in routine use, illustrated how far work has progressed

Reactor PIK construction

International Nuclear Information System (INIS)

Konoplev, Kir

2003-01-01

The construction work at the 100 MW researches reactor PIK in year 2002 was in progress. The main activity was concentrated on mechanical, ventilation and electrical equipment. Some systems and subsystems are under adjustment. Hydraulic driving gear for beam shutters are finished in installation, rinsing, and adjusting. Regulating rods test assembling was done. On the critical assembly the first reactor fueling was tested to evaluate the starting neutron source intensity and a sufficiency of existing control and instrument board. Mainline of the PIK facility design and neutron parameters are presented. (author)

Research nuclear reactor RA - Annual report 1992

International Nuclear Information System (INIS)

Sotic, O.

1992-12-01

Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

A stability identification system for boiling water nuclear reactors

International Nuclear Information System (INIS)

Belblidia, L.A.; Chevrier, A.

1994-01-01

Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

Proceedings of the European Research Reactor Conference - RRFM 2013 Transactions

International Nuclear Information System (INIS)

2013-01-01

In 2013 RRFM, the European Research Reactor Conference is jointly organised by ENS and Atomexpo LLC. This time the Research Reactor community meet in St. Petersburg, Russia. The conference programme will revolve around a series of Plenary Sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions will focus on all areas of the Fuel Cycle of Research Reactors, their Utilisation, Operation and Management as well as specific research projects and innovative methods in research reactor analysis and design. In 2013 the European Research Reactor Conference will for the first time give special attention to complementary safety assessments of Research Reactors, following the Fukushima-Dai-Ichi NPP's Accident. (authors)

Experimental determination of the decay constant Alpha in the zero power reactor SUR 100 BE

International Nuclear Information System (INIS)

Lickteig, K.

1975-02-01

The paper discusses experiments with a pulsed source and Rossi-Alpha experiments. In the first case, the effects of higher harmonies and detector position are investigated. In the Rossi-Alpha method, the interest was centered on the correlation between reactivity source strength and method of measurement. (RW/AK) [de

Des pratiques culturales influent sur les attaques de deux ravageurs ...

African Journals Online (AJOL)

La culture de tomate est attaquée par plusieurs ravageurs dont Helicoverpa armigera et Tuta absoluta. Dans le but d'évaluer l'effet des pratiques culturales de la tomate sur ces principaux ravageurs dans les Niayes (Sénégal), un échantillonnage de 98 parcelles est effectué, sur quatre cycles de culture en saison sèche, de ...

Manual for the operation of research reactors

International Nuclear Information System (INIS)

1965-01-01

The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

KAUST Repository

Zhao, Zhenlong

2013-01-17

Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

Environmental assessment of SP-100 ground engineering system test site: Hanford Site, Richland, Washington

Energy Technology Data Exchange (ETDEWEB)

1988-12-01

The US Department of Energy (DOE) proposes to modify an existing reactor containment building (decommissioned Plutonium Recycle Test Reactor (PRTR) 309 Building) to provide ground test capability for the prototype SP-100 reactor. The 309 Building (Figure 1.1) is located in the 300 Area on the Hanford Site in Washington State. The National Environmental Policy Act (NEPA) requires that Federal agencies assess the potential impacts that their actions may have on the environment. This Environmental Assessment describes the consideration given to environmental impacts during reactor concept and test site selection, examines the environmental effects of the DOE proposal to ground test the nuclear subsystem, describes alternatives to the proposed action, and examines radiological risks of potential SP-100 use in space. 73 refs., 19 figs., 7 tabs.

Les mesures de gel et de confiscation en vertu de la Loi sur les embargos

OpenAIRE

Schnyder, Nicolas

2009-01-01

Etude du mécanisme d'adoption des mesures de gel et de confiscation en vertu de la Loi sur les embargos et leurs conséquences tant sur la personne principalement touchée par ces mesures que sur les tiers.

Singular perturbation analysis of relaxation oscillations in reactor systems

International Nuclear Information System (INIS)

Ward, M.E.; Lee, J.C.

1987-01-01

A singular perturbation method for the analysis of large power oscillations in nuclear reactors is applied to obtain phase-plane solutions of the Ergen-Weinberg model. The system equations, recast in an appropriate form, directly give a first approximation to the closed trajectory in which the system behaviour is idealized as relaxation oscillations. Further approximations in the phase plane are determined using separate perturbation series on individual parts of the oscillation, with variations in the assignment of dependent and independent variables to consistently obtain convergent series. The accuracy of each order of the phase-plane solution increases with the magnitude of the power pulse in the actual physical situation. For realistic reactor conditions, both the trajectory and period of oscillation are well predicted using the first two terms of each perturbation series

SXPS study of model GaAs(100)/electrolyte interface

Energy Technology Data Exchange (ETDEWEB)

Lebedev, Mikhail V. [A.F. Ioffe Physico-Technical Institute, Russian Academy of Sciences, St. Petersburg (Russian Federation); Mankel, Eric; Mayer, Thomas; Jaegermann, Wolfram [Institute of Material Sciences, Darmstadt University of Technology, Darmstadt (Germany)

2010-02-15

Model GaAs(100)/electrolyte interfaces are prepared in vacuum by co-adsorption of Cl{sub 2} and 2-propanol molecules at LN{sub 2} temperature. On adsorption of Cl{sub 2} molecules gallium chlorides, elemental arsenic and arsenic chlorides are formed. Co-adsorption of 2-propanol causes formation of additional GaCl{sub 3} and AsCl, as well as soluble/volatile As-based complexes, which are released from the surface depleting the sur- face by arsenic. Comparison of the As 3d and Ga 3d spectra obtained after heating the model interface to room temperature with the corresponding spectra obtained after emersion of the GaAs(100) surface from HCl/2-propanol solution allows to conclude that in HCl solution Cl{sup -} ions attack gallium sites and H{sup +} ions mostly attack arsenic sites. (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Commission des Nations Unies sur les produits indispensables aux ...

International Development Research Centre (IDRC) Digital Library (Canada)

Étude des soins de santé primaires dispensés dans les pays du cône Sud. Les pays membres du Réseau de recherche sur les systèmes et services de santé dans les pays du cône Sud (Red de Investigación en Sistemas y Servicios de Salud en el Cono Sur) ont des systèmes de... Voir davantageÉtude des soins de santé ...

The SNL100-01 blade :

Energy Technology Data Exchange (ETDEWEB)

Griffith, Daniel

2013-02-01

A series of design studies to investigate the effect of carbon on blade weight and performance for large blades was performed using the Sandia 100-meter All-glass Baseline Blade design as a starting point. This document provides a description of the final carbon blade design, which is termed as SNL100-01. This report includes a summary of the design modifications applied to the baseline all-glass 100-meter design and a description of the NuMAD model files that are made publicly available. This document is intended primarily to be a companion document to the distribution of the NuMAD blade model files for SNL100-01.

Space reactor safety, 1985--1995 lessons learned

International Nuclear Information System (INIS)

Marshall, A.C.

1995-01-01

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration

Space reactor safety, 1985--1995 lessons learned

Energy Technology Data Exchange (ETDEWEB)

Marshall, A.C.

1995-12-31

Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

Marginal and joint distributions of S100, HMB-45, and Melan-A across a large series of cutaneous melanomas.

Science.gov (United States)

Viray, Hollis; Bradley, William R; Schalper, Kurt A; Rimm, David L; Gould Rothberg, Bonnie E

2013-08-01

The distribution of the standard melanoma antibodies S100, HMB-45, and Melan-A has been extensively studied. Yet, the overlap in their expression is less well characterized. To determine the joint distributions of the classic melanoma markers and to determine if classification according to joint antigen expression has prognostic relevance. S100, HMB-45, and Melan-A were assayed by immunofluorescence-based immunohistochemistry on a large tissue microarray of 212 cutaneous melanoma primary tumors and 341 metastases. Positive expression for each antigen required display of immunoreactivity for at least 25% of melanoma cells. Marginal and joint distributions were determined across all markers. Bivariate associations with established clinicopathologic covariates and melanoma-specific survival analyses were conducted. Of 322 assayable melanomas, 295 (91.6%), 203 (63.0%), and 236 (73.3%) stained with S100, HMB-45, and Melan-A, respectively. Twenty-seven melanomas, representing a diverse set of histopathologic profiles, were S100 negative. Coexpression of all 3 antibodies was observed in 160 melanomas (49.7%). Intensity of endogenous melanin pigment did not confound immunolabeling. Among primary tumors, associations with clinicopathologic parameters revealed a significant relationship only between HMB-45 and microsatellitosis (P = .02). No significant differences among clinicopathologic criteria were observed across the HMB-45/Melan-A joint distribution categories. Neither marginal HMB-45 (P = .56) nor Melan-A (P = .81), or their joint distributions (P = .88), was associated with melanoma-specific survival. Comprehensive characterization of the marginal and joint distributions for S100, HMB-45, and Melan-A across a large series of cutaneous melanomas revealed diversity of expression across this group of antigens. However, these immunohistochemically defined subclasses of melanomas do not significantly differ according to clinicopathologic correlates or outcome.

A model to describe the performance of the UASB reactor.

Science.gov (United States)

Rodríguez-Gómez, Raúl; Renman, Gunno; Moreno, Luis; Liu, Longcheng

2014-04-01

A dynamic model to describe the performance of the Upflow Anaerobic Sludge Blanket (UASB) reactor was developed. It includes dispersion, advection, and reaction terms, as well as the resistances through which the substrate passes before its biotransformation. The UASB reactor is viewed as several continuous stirred tank reactors connected in series. The good agreement between experimental and simulated results shows that the model is able to predict the performance of the UASB reactor (i.e. substrate concentration, biomass concentration, granule size, and height of the sludge bed).

Estimasi Model Seemingly Unrelated Regression (SUR dengan Metode Generalized Least Square (GLS

Directory of Open Access Journals (Sweden)

Ade Widyaningsih

2015-04-01

Full Text Available Regression analysis is a statistical tool that is used to determine the relationship between two or more quantitative variables so that one variable can be predicted from the other variables. A method that can used to obtain a good estimation in the regression analysis is ordinary least squares method. The least squares method is used to estimate the parameters of one or more regression but relationships among the errors in the response of other estimators are not allowed. One way to overcome this problem is Seemingly Unrelated Regression model (SUR in which parameters are estimated using Generalized Least Square (GLS. In this study, the author applies SUR model using GLS method on world gasoline demand data. The author obtains that SUR using GLS is better than OLS because SUR produce smaller errors than the OLS.

Estimasi Model Seemingly Unrelated Regression (SUR dengan Metode Generalized Least Square (GLS

Directory of Open Access Journals (Sweden)

Ade Widyaningsih

2014-06-01

Full Text Available Regression analysis is a statistical tool that is used to determine the relationship between two or more quantitative variables so that one variable can be predicted from the other variables. A method that can used to obtain a good estimation in the regression analysis is ordinary least squares method. The least squares method is used to estimate the parameters of one or more regression but relationships among the errors in the response of other estimators are not allowed. One way to overcome this problem is Seemingly Unrelated Regression model (SUR in which parameters are estimated using Generalized Least Square (GLS. In this study, the author applies SUR model using GLS method on world gasoline demand data. The author obtains that SUR using GLS is better than OLS because SUR produce smaller errors than the OLS.

Modeling 100,000-year climate fluctuations in pre-Pleistocene time series

Science.gov (United States)

Crowley, Thomas J.; Kim, Kwang-Yul; Mengel, John G.; Short, David A.

1992-01-01

A number of pre-Pleistocene climate records exhibit significant fluctuations at the 100,000-year (100-ky) eccentricity period, before the time of such fluctuations in global ice volume. The origin of these fluctuations has been obscure. Results reported here from a modeling study suggest that such a response can occur over low-altitude land areas involved in monsoon fluctuations. The twice yearly passage of the sun across the equator and the seasonal timing of perihelion interact to increase both 100-ky and 400-ky power in the modeled temperature field. The magnitude of the temperature response is sufficiently large to leave an imprint on the geologic record, and simulated fluctuations resemble those found in records of Triassic lake levels.

Performance testing of refractory alloy-clad fuel elements for space reactors

International Nuclear Information System (INIS)

Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K.

1985-01-01

Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates (Nb-1Zr, PWC-11, and Mo-13Re) and fuel forms (UN and UO 2 ) are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts

Effet du compost à base de Calotropis procera (Aiton) W.T. Aiton sur ...

African Journals Online (AJOL)

Objectifs: Une étude à base de compost de Calotropis procera a été menée afin de mesurer l'effet sur la production de l'arachide sur de sols pauvres. Méthodologie et résultats: L'essai a été conduit sur le site de Doyaba au niveau des zones marginales du Tchad avec la variété fleur 11 d'arachide (Arachis hypogaea L.) ...

Sterilization of swine wastewater treated by anaerobic reactors using UV photo-reactors

Directory of Open Access Journals (Sweden)

Erlon Lopes Pereira

2014-09-01

Full Text Available The use of ultraviolet radiation is an established procedure with growing application forthe disinfection of contaminated wastewater. This study aimed to evaluate the efficiency of artificial UV radiation, as a post treatment of liquid from anaerobic reactors treating swine effluent. The UV reactors were employed to sterilize pathogenic microorganisms. To this end, two photo-reactors were constructed using PVC pipe with100 mm diameter and 1060 mmlength, whose ends were sealed with PVC caps. The photo-reactors were designed to act on the liquid surface, as the lamp does not get into contact with the liquid. To increase the efficiency of UV radiation, photo-reactors were coated with aluminum foil. The lamp used in the reactors was germicidal fluorescent, with band wavelength of 230 nm, power of 30 Watts and manufactured by Techlux. In this research, the HRT with the highest removal efficiency was 0.063 days (90.6 minutes, even treating an effluent with veryhigh turbidity due to dissolved solids. It was concluded that the sterilization method using UV has proved to be an effective and appropriate process, among many other procedures.

Gestion des agroecosystemes sur le mont agou en zone forestiere ...

African Journals Online (AJOL)

Dans le but d'une meilleure mise en valeur de l'espace, plusieurs espèces végétales sont cultivées en association essentiellement avec Persea americana. L'approche méthodologique est basée sur des inventaires floristiques et écologiques sur 45 placettes de 25 m x 25 m dans les agrosystèmes. Une diversité de 85 ...

Taxes sur les cigarettes en Tanzanie | CRDI - Centre de recherches ...

International Development Research Centre (IDRC) Digital Library (Canada)

On s'attend à ce que cette hausse du tabagisme ait d'importantes répercussions sur la santé publique et le développement économique. Il ressort de l'expérience d'autres régions que le moyen le plus efficace de renverser une telle tendance consiste à augmenter les taxes sur les cigarettes. Toutefois, les responsables des ...

General aspects of CAREM-25 reactor

International Nuclear Information System (INIS)

Delmastro, Dario F.; Gomez, Silvia; Ishida, Viviana; Mazzi, Ruben; Santecchia, Alberto; Gomez de Soler, Susana M.

2000-01-01

CAREM project consists on the development and design of an advanced nuclear power plant. In order to verify its innovative features the construction of a prototype is planned. In this paper the main technical characteristics of CAREM-25 prototype reactor are presented. This is a very low power innovative reactor (100 M Wth) conceived with new generation design solutions. Based on an indirect cycle integrated light water reactor using enriched uranium, CAREM has some distinctive features that greatly simplify the reactor and also contribute to a high level of safety: -) Integrated primary system; -) Primary system cooling by natural convection; -) Self pressurization; -) and Passive safety systems. (author)

Problems and prospects of small and medium power reactors

International Nuclear Information System (INIS)

Matin, A.

1977-01-01

Prior to 1973 it was generally believed that small and medium power reactors (SMPRs) had a potentially large market and only their high capital costs prevented their large scale commercial application. In December, 1973, crude oil price rose from US $2.50 per barrel to more than US $11 per barrel. This changed the economic position of SMPRs so much so that even 100-200 MWe nuclear reactors were considered economic compared to oil-fired plants. A Market Survey by the International Atomic Energy Agency in 1974 showed that the potential market for reactors ranging from 150 to 400 MWe during 1980-1990 amounted to 140 units with a total installed capacity of 38,000 MWe. This potential market did not, however, generate the desired interest among the reactor manufacturers. So far only three manufacturers based in Europe have shown interest in SMPRs and at present small reactors are being built commercially only in India. Among developing countries, Bangladesh, Jamaica and Kuwait are seriously looking for reactors in sizes of 100-200 MWe. The paper analyses the historic background of SMPRs and problems related to their commercial application and suggests the following actions: i) The British 100 MWe SGHWR is considered proven and suitable for small grids and hence deserves financial support by British/International Financing Agencies. ii) Any re-engineered or slightly re-designed version of operating small light water reactors will find wider acceptability than available new adaptions of marine reactors. Manufacturers of operating small LWRs may be encouraged through international financial assistance to make such designs commercially available. iii) Small CANDU reactors may be suitable for most developing countries and need technical and economic support from Canada for their export. iv) The Agency must continue their effort more vigorously for making SMPRs commercially available to small developing countries

Unique features of space reactors

International Nuclear Information System (INIS)

Buden, D.

1990-01-01

This paper reports on space reactors that are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K

Health physics aspects of advanced reactor licensing reviews

Energy Technology Data Exchange (ETDEWEB)

Hinson, C.S. [Nuclear Regulatory Commission, Washington, DC (United States)

1995-03-01

The last Construction Permit to be issued by the U.S. Nuclear Regulatory Commission (NRC) for a U.S. light water reactor (LWR) was granted in the late 1970s. In 1989 the NRC issued 10 CFR Part 52 which is intended to serve as a framework for the licensing of future reactor designs. The NRC is currently reviewing four different future on {open_quotes}next-generation{close_quotes} reactor designs. Two of these designs are classified as evolutionary designs (modified versions of current generation LWRs) and two are advanced designs (reactors incorporating simplified designs and passive means for accident mitigation). These {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs incorporate many innovative design features which are intended to maintain personnel doses ALARA and ensure that the annual average collective dose at these reactors does not exceed 100 person-rems (1 person-sievert) per year. This paper discusses some of the ALARA design features which are incorporated in the four {open_quotes}next-generation{close_quotes} reactor designs currently being reviewed by the NRC.

Criticality studies; Etudes de criticite

Energy Technology Data Exchange (ETDEWEB)

Breton, D; Lecorche, P; Clouet d' Orval, Ch [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Criticality studies made at the Commissariat a l'Energie atomique deal on the one hand with experiments on plutonium and uranium solutions, on the other hand with theoretical work on the development and use of computation, methods for the resolution of problems concerning the nuclear safety of chemical and metallurgical plants. I - Since 1958 the experimental studies have dealt with homogeneous media constituted by a fissile salt dissolved in light water. Developed using the reactor Proserpine, the experiments have been carried on at Saclay on the Alecto assemblies where solutions of plutonium or of 90 p.100 - enriched uranium can be made critical. The results already obtained relate to critical masses of cylindrical tanks of diameters from 20 to 50 cm. reflected in several ways (water, concrete, etc. . ) at concentrations up to 100 g/liter. Physical measurements (spectra, reactor noises) and interaction measurements complete the results. Other experiments relating to plutonium solutions were begun in 1963, at the Valduc Center. They deal with the study of critical masses of annular vessels of external diameter 50 cm and internal diameter varying from 10 to 30 cm. These vessels can be water reflected internally, externally, or both. Two of these vessels have been studied in interaction for various geometries. Slabs of various thicknesses were also studied. II - The studies thus undertaken allowed the development of methods of computation which have been tested on several experiments. Particular use has been made of the possibilities of calculations based on transport theory and on Monte Carlo methods. All these theoretical studies are applied to the design and control of industrial plants from the point of view of safety. (authors) [French] Les etudes de criticite effectuees au CEA comportent d'une part des experiences sur des solutions de plutonium et d'uranium enrichi, d'autre part des travaux theoriques portant sur la mise au point et l'exploitation de methodes

Th-100 Nuclear Power Plant

International Nuclear Information System (INIS)

Reitsma, Frederik

2013-01-01

Steenkampskraal Thorium Limited (STL) is a private company which is designing, marketing, licensing and commercializing a 100MWt thorium fueled pebble bed reactor. The concept plant design has been completed and work on the basic design has started. First site to determine the fuel cycle employed. Strong emphasis is placed on modular construction to reduce costs. STL hopes to start the licensing process within the next 6-8 months

Dating of the Francevillian sedimentary series and mineralogic and isotopic (Sm, Nd, Rb, Sr, K, Ar, U, O and C) characterization of the gangue of the reactors 10 and 13. Preliminary report

International Nuclear Information System (INIS)

Gautier-Lafaye, F.; Stille, P.; Bros, R.; Taieb, R.

1993-01-01

This paper summarizes the various ages reported for the diagenetic events in the Francevillian sedimentary series (Precambrian era) and the fission reactors of Oklo. Obviously, differences exist between the ages obtained on the silicate minerals and the ages obtained on the Uranium ores and on the reactors. Clay minerals which crystallized during the fission reactions yield younger ages than the reactors themselves. Similarly, the diagenetic clays (1870 Ma) show younger ages than the Uranium ores (2000 Ma). This is in contrast to mineralogical and field evidence indicating that Uranium mineralization occurred during diagenesis of the Francevillian sediments. These antithetical results give rise to several questions. Does the age obtained on the diagenetic clays date a late thermal event or does the age of the Uranium mineralization reflect a multistage U-Pb history. This work tries to bring answers with the help of new isotopic analysis and studies mineralogy of the gangue of reactors and isotopic compositions in Uranium ores. 8 refs., 4 figs

IAEA high temperature gas-cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2000-01-01

The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

IAEA high temperature gas cooled reactor activities

International Nuclear Information System (INIS)

Kendall, J.M.

2001-01-01

IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

Study relating to the physico-chemical behaviour of heavy water in nuclear reactors; Etudes relatives au comportement physico-chimique de l'eau lourde dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Chenouard, J; Dirian, G; Roth, E; Vignet, P; Platzer, R [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1959-07-01

Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [French] Parmi les degradations subies par l'eau lourde dans les reacteurs nucleaires, les deux plus importantes sont la pollution chimique et isotopique et la decomposition radiolytique. La pollution chimique a conduit a mettre au point pour le cas particulier des reacteurs, des circuits d'epuration par echange d'ions. On decrit ici en detail la mise en oeuvre de cette methode dans les reacteurs de recherche du CEA; les controles qu'elle necessite, les resultats obtenus et leur interpretation. En ce qui concerne la dissociation radiolytique de l'eau, les renseignements obtenus sur ces memes reacteurs sont communiques, ainsi que les details des dispositifs de recombinaison et des moyens de controle. Enfin, on fait le point des etudes poursuivies au CEA sur ces memes problemes de recombinaison dans le cas des reacteurs de puissance. (auteur)

Main results of BN-600 reactor stress-strain state investigations

International Nuclear Information System (INIS)

Panov, V.A.

1983-01-01

The development of BN-600 fast reactor plant needed the solution of a series of complex engineering problems including ones for confirming integrity of the most vital structural components. The particular attention was given to the main vessel since reactor availability end safe operation of the plant as a whole depend on vessel strength end integrity. The present report deals with the main results of theoretical and experimental investigations of the stress-strain state of BN-600 reactor vessel carried out during design, start-up and initial bringing the reactor to power

MOX in reactors: present and future

International Nuclear Information System (INIS)

Arslan, Marc; Gros, Jean Pierre; Niquille, Aurelie; Marincic, Alexis

2010-01-01

In Europe, MOX fuel has been supplied by AREVA for more than 30 years, to 36 reactors: 21 in France, 10 in Germany, 3 in Switzerland, 2 in Belgium. For the present and future, recycling is compulsory in the frame of sustainable development of nuclear energy. By 2030 the overall volume of used fuel will reach about 400 000 t worldwide. Their plutonium and uranium content represents a huge resource of energy to recycle. That is the reason why, the European Utilities issued an EUR (European Utilities Requirement) demanding new builds reactors to be able of using MOX Fuel Assemblies in up to 50 % of the core. AREVA GEN3+ reactors, like EPR TM or ATMEA TM designed with MHI partnership, are designed to answer any utility need of MOX recycling. The example of the EPR TM reactor operated with 100 % MOX core optimized for MOX recycling will be presented. A standard EPR TM can be operated with 100 % MOX core using an advanced homogeneous MOX (single Pu content) with highly improved performances (burn-up and Cycle length). The adaptations needed and the main operating and safety reactor features will be presented. AREVA offers the utilities throughout the world, fuel supply (UO 2 , ERU, MOX), and reactors designed with all the needed capability for recycling. For each country and each utility, an adapted global solution, competitive and non proliferant can be proposed. (authors)

Nuclear Power Reactors in the World. 2014 Ed

International Nuclear Information System (INIS)

2014-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-fourth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2013. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects this data through designated national correspondents in Member States

Nuclear Power Reactors in the World. 2016 Ed

International Nuclear Information System (INIS)

2016-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to reactor units in IAEA Member States. This thirty-sixth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2015. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. Data is collected by the IAEA via designated national correspondents in Member States

Influence du système de logement sur quelques performances ...

African Journals Online (AJOL)

Les effets de deux systèmes de logement (batterie de cages et sol sur litière) sur les performances zootechniques et économiques des poules pondeuses ont été évalués en zone tropicale humide, au sud du Benin. Un total de 180 poules pondeuses (Isa Brown) de 26 semaines d'âge était réparti en deux groupes de 108 et ...

Chernobyl reactor transient simulation study

International Nuclear Information System (INIS)

Gaber, F.A.; El Messiry, A.M.

1988-01-01

This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

Effect of the temperature and of the organic load in two-stage up flow anaerobic sludge blanket reactors treating of swine wastewater

Energy Technology Data Exchange (ETDEWEB)

Bichuette, Alexandre Abud; Duda, Rose Maria; Oliveira, Roberto Alves de [Universidade Estadual Paulista (UNESP), Jaboticabal, SP (Brazil). Dept. de Engenharia Rural], E-mail: oliveira@fcav.unesp.br

2008-07-01

In this work the acting of two-stage up flow anaerobic sludge blanket reactors (UASB) was evaluated, installed in series, in pilot scale (volumes of 908 L and 350 L, respectively) in the treatment swine wastewater, with concentrations of total solids suspended (TSS) around 10000 mg L{sup -1}. The organic loading rates (OLR) applied in first UASB were of 5,2 and of 8,6 g total COD (Ld){sup -1}. The medium efficiencies of removal of the chemical demand of total oxygen (total COD), TSS and TKN were higher than 89; 80 and 55%, respectively, for the system of anaerobic treatment composed by the reactors UASB in two apprenticeships. The rate of volumetric methane production in the system of anaerobic treatment with the reactors UASB were 0,08 and 0,16 m{sup 3}CH{sub 4} (m{sup 3} CH{sub 4} reactor d){sup -1}. The number of total coliforms was reduced to 2,6x10{sup 4} NMP/100 mL. (author)

Development of reactivity feedback effect measurement techniques under sub-critical condition in fast reactors

International Nuclear Information System (INIS)

Kitano, A.; Nishi, H.; Suzuki, T.; Okajima, S.; Kanemoto, S.

2012-01-01

The first-of-a-kind reactor has been licensed by a safety examination of the plant design based on the measured data in precedent mock-up experiments. The validity of the safety design can be confirmed without a mock-up experiment, if the reactor feed-back characteristics can be measured before operation, with the constructed reactor itself. The 'Synthesis Method', a systematic and sophisticated method of sub-criticality measurement, is proposed in this work to ensure the safety margin before operation. The 'Synthesis Method' is based on the modified source multiplication method (MSM) combined with the noise analysis method to measure the reference sub-criticality level for MSM. A numerical simulation for the control-rod reactivity worth and the isothermal feed-back reactivity was conducted for typical fast reactors of 100 MWe-size, 300 MWe-size, 750 MWe-size, and 1500 MWe-size to investigate the applicability of Synthesis Method. The number of neutron detectors and their positions necessary for the measurement were investigated for both methods of MSM and the noise analysis by a series of parametric survey calculations. As a result, it was suggested that a neutron detector located above the core center and three or more neutron detectors located above the radial blanket region enable the measurement of sub-criticality within 10% uncertainty from -$0.5 to -$2 and within 15% uncertainty for the deeper sub-criticality. (authors)

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

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

2014-01-01

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

Energy memento; Memento sur l'energie

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-07-01

This memento about energy provides a series of tables with numerical data relative to energy resources and uses in France, in the European Union and in the rest of the world: energy consumption (primary energy, forecasting, CO{sub 2} emissions, energy independence, supplies, uses and imports, demand scenarios, energy savings..), power production (production, forecasting, loads, consumption, hydro-power, thermal equipment, exports), nuclear power (production, forecasting, reactors population, characteristics of French PWRs, uranium needs and fuel cycle), energy resources (renewable energies, fossil fuels and uranium reserves and production), economic data (gross national product, economic and energy indicators, prices and cost estimations), energy units and conversion factors (counting, calorific value of coals, production costs, energy units). (J.S.)

Simulation of power excursions - Osiris reactor; Simulation des excursions de puissance - pile Osiris

Energy Technology Data Exchange (ETDEWEB)

Pascouet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Following the experimental work accomplished in the U.S.A. on Borax 1 and SPERT 1 and the accident of SL 1, the 'Commissariat a l'Energie Atomique' started a research program about the safety of its own swimming Pool reactors, with regard to power excursions. The first research work led to the design of programmed explosive charges, adapted to the simulation of a power excursion. This report describes the application of these methods to the investigation of Osiris safety. (author) [French] A la suite des essais effectues aux U.S.A. sur BORAX 1 et SPERT 1 et de l'accident survenu a SL 1, le Commissariat a l'Energie Atomique a lance un programme d'etudes sur la surete de ses reacteurs piscines vis-a-vis des excursions de puissance. Les premieres etudes ont abouti A la mise au point de charges programmees capables de simuler une excursion de puissance. On trouvera dans le present rapport l'application de ces methodes a l'etude de la surete d'OSIRIS. (auteur)

Separation of lithium isotopes on ion exchangers; Separation des isotopes du lithium sur echangeurs d'ions

Energy Technology Data Exchange (ETDEWEB)

Menes, F; Saito, E; Roth, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

A survey of the literature shows that little information has been published on the separation of lithium isotopes with ion exchange resins. We have undertaken a series of elutions using the ion-exchange resins 'Dowex 50 x 12' and IRC 50, and various eluting solutions. Formulae derived from the treatment of Mayer and Tompkins permit the calculation of the separation factor per theoretical plate. For the solutions tried out in our experiments the separation factors lie in the interval 1.001 to 1.002. These values are quite low in comparison to the factor 1.022 found by Taylor and Urey for ion exchange with zeolites. (author) [French] Nous avons trouve relativement peu de donnees dans la litterature scientifique sur la separation des isotopes de lithium par les resines echangeuses d'ions. Nous avons effectue un certain nombre d'essais sur Dowex 50 X 12 et IRC 50 utilisant divers eluants. Des formules derivees de celles de Mayer et Tompkins permettent le calcul du coefficient de separation par plateau theorique. Pour les eluants etudies, ces facteurs de separation se trouvent entre 1,001 et 1,002. Ces valeurs sont faibles en comparaison du facteur 1,022 trouve par Taylor et Urey pour les zeolithes. (auteur)

SP-100 nuclear assembly test: Test assembly functional requirements and system arrangement

International Nuclear Information System (INIS)

Fallas, T.T.; Gluck, R.; Motwani, K.; Clay, H.; O'Neill, G.

1991-01-01

This paper describes the functional requirements and the system that will be tested to validate the reactor, flight shield, and flight controller of the SP-100 Generic Flight System (GFS). The Nuclear Assembly Test (NAT) consists of the test article (SP-100 reactor with control devices and the flight shield) and its supporting systems. The NAT test assembly is being designed by GE. Westinghouse Hanford Company (WHC) is designing the test cell and vacuum vessel system that will contain the NAT test assembly (Renkey et al. 1989). Preliminary design reviews have been completed and the final design is under way

Autoridad y memoria entre los killakas. Las estrategias discursivas de don Juan Colque Guarache en el sur andino (Siglo XVI)

OpenAIRE

Graña, Mario Julio

2014-01-01

Entre 1575 y 1577, don Juan Colque Guarache, cacique principal de las etnias killaka, asanaque, sivaroyo,haracapi y uruquilla del sur del lago Poopó, actual Oruro (Bolivia)eleva a consideración de las autoridades españolas una serie de "Probanzasde Méritos y Servicios" de sus antepasados. Esta fuente permite explorar la narración de la historia de un linaje indígena, a la luz de las estrategias discursivas empleadas por las autoridades étnicas dentro del contexto colonial.En este sentido, se ...

Fiscal year 1992 report on archaeological surveys of the 100 Areas, Hanford Site, Washington

Energy Technology Data Exchange (ETDEWEB)

Wright, M.K.

1993-09-01

During FY 1992, the Hanford Cultural Resources Laboratory (HCRL) conducted a field survey of the 100-HR-3 Operable Unit (600 Area) and tested three sites near the 100 Area reactor compounds on the US Department of Energy`s Hanford Site at the request of Westinghouse Hanford Company. These efforts were conducted in compliance with Section 106 of the National Historic Preservation Act (NHPA) and are part of a cultural resources review of 100 Area Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) operable units in support of CERCLA characterization studies.The results of the FY 1992 survey and test excavation efforts are discussed in this report. 518 ha in the 100-HR-3 Operable Unit and conducted test excavations at three prehistoric sites near the 100-F and 100-K reactors to determine their eligibility for listing on the National Register of Historic Places.

Reconfiguración de la Cooperación Sur-Sur en la región latinoamericana y la participación internacional de los actores subnacionales

Directory of Open Access Journals (Sweden)

Mariana Calvento

2015-01-01

Full Text Available El presente trabajo describe y analiza las actua-les características de la Cooperación Sur-Sur en América Latina, especialmente el papel de los actores subnacionales en dicha dinámica. A su vez, analiza las modalidades que adquiere esta cooperación en el contexto político regio-nal de la última década, donde se evidencia un importante avance de la integración de los países latinoamericanos. Finalmente, estudila relación existente entre los principios de la Cooperación Sur-Sur con la participación in-ternacional de los actores subnacionales a partir de la relevancia que adquieren algunas de es-tas formas de cooperación, como es el caso de la Red mercociudades, el Foro Consultivo de Municipios, Estados Federados, Provincias y Departamentos del mercosur (fccr mer-cosur y la Federación Latinoamericana de Ciudades, Municipios y Asociaciones de Go-biernos Locales (flacma, que nuclean a estos actores y promueven mecanismos institucionales que favorecen la participación de las ciudades a través de diferentes instrumentos de intercambio y cooperación.

Study relating to the physico-chemical behaviour of heavy water in nuclear reactors; Etudes relatives au comportement physico-chimique de l'eau lourde dans les reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Chenouard, J.; Dirian, G.; Roth, E.; Vignet, P.; Platzer, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1959-07-01

Chemical and isotope pollution, and radiolytic decomposition are the two most important ways in which heavy water becomes degraded in nuclear reactors. Chemical pollution has led to the creation of ion exchange purification loops specially designed for reactors: the report contains a description in detail of the application of this purification method in CEA research reactors, including the analysis required, results obtained, and their interpretation. The intelligence obtained on radiolytic decomposition with the same facilities is also discussed, as well as the recombination apparatus and control equipment utilized. Finally, investigation to date in the CEA on recombination circuits for power reactors is also discussed. (author) [French] Parmi les degradations subies par l'eau lourde dans les reacteurs nucleaires, les deux plus importantes sont la pollution chimique et isotopique et la decomposition radiolytique. La pollution chimique a conduit a mettre au point pour le cas particulier des reacteurs, des circuits d'epuration par echange d'ions. On decrit ici en detail la mise en oeuvre de cette methode dans les reacteurs de recherche du CEA; les controles qu'elle necessite, les resultats obtenus et leur interpretation. En ce qui concerne la dissociation radiolytique de l'eau, les renseignements obtenus sur ces memes reacteurs sont communiques, ainsi que les details des dispositifs de recombinaison et des moyens de controle. Enfin, on fait le point des etudes poursuivies au CEA sur ces memes problemes de recombinaison dans le cas des reacteurs de puissance. (auteur)

Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

International Nuclear Information System (INIS)

Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

1991-02-01

The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

Pyrometric fuel particle measurements in pressurised reactors

Energy Technology Data Exchange (ETDEWEB)

Hernberg, R.; Joutsenoja, T. [Tampere Univ. of Technology (Finland)

1996-12-01

A fiberoptic two-colour pyrometric technique for fuel particle temperature and size measurement is modified and applied to three pressurized reactors of different type in Finland, Germany and France. A modification of the pyrometric method for simultaneous in situ measurement of the temperature and size of individual pulverized coal particles at the pressurized entrained flow reactor in Jyvaeskylae was developed and several series of measurements were made. In Orleans a fiberoptic pyrometric device was installed to a pressurised thermogravimetric reactor and the two-colour temperatures of fuel samples were measured. Some results of these measurements are presented. The project belongs to EU`s Joule 2 extension research programme. (author)

Shutdown channels and fitted interlocks in atomic reactors; Chaines de securite et verrouillages installes sur les piles atomiques

Energy Technology Data Exchange (ETDEWEB)

Furet, J; Landauer, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

This catalogue consists of tables (one per reactor) giving the following information: number and type of detectors, range of the shutdown channels, nature of the associated electronics, thresholds setting off the alarms, fitted interlocks. These cards have been drawn up with a view to an examination of the reactors safety by the 'Reactor Safety Sub-Commission', they take into account the latest decisions. The reactors involved in this review are: Azur, Cabri, Castor-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, and Ulysse. (authors) [French] Ce catalogue est compose d'un ensemble de tableaux (a raison de un tableau par pile) donnant les renseignements suivants: nombre et nature des detecteurs, dynamique des chaines, nature de l'electronique associee, seuils provoquant des actions de securite, verrouillages installes. Ces fiches ont ete etablies en vue de l'examen de la securite des piles par la 'Sous-Commission de Surete des Piles', et tiennent compte des decisions de celle-ci. Les reacteurs concernes sont: Azur, Cabri, Cator-Pollux, Cesar-Marius-2, Edf-2, EL3, EL4, Eole, G1, G2-G3, Harmonie, Isis, Masurca, Melusine, Minerve, Osiris, Pegase, Peggy, PAT, Rapsodie, SENA, Siloe, Siloette, Triton-Nereide, et Ulysse. (auteurs)

Controverses sur les sucres [Controversies about sugars

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Malika BOUCHENAK

2016-12-01

Full Text Available La surconsommation de sucres ajoutés et leurs effets putatifs sur diverses pathologies cardiométaboliques continuent de susciter des controverses. Dans les années 1950, le taux élevé de maladies cardiovasculaires (MCV, chez les hommes Américains a conduit à des études sur le rôle des facteurs alimentaires, tels que le cholestérol, les phytostérols, l'excès de calories, les acides aminés, les graisses, les glucides, les vitamines et les minéraux, à influencer le risque de MCV. Dans les années 60, deux éminents physiologistes, plaidant pour des hypothèses causales divergentes de MCV, John Yudkin a identifié les sucres ajoutés comme agent primaire, tandis que Ancel Keys identifiait les graisses totales, les graisses saturées et le cholestérol alimentaire. Cependant, dans les années 80, peu de scientifiques croyaient que les sucres ajoutés jouaient un rôle important dans la maladie coronarienne, et les premiers conseils nutritionnels de 1980 (1980 Dietary Guidelines for americans étaient axés sur la réduction des graisses totales, des graisses saturées et du cholestérol alimentaire pour la prévention des MCV.

ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

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Alexander Labutin

2017-02-01

Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

New competition in the world market of nuclear reactors; La nouvelle concurrence sur le marche mondial des reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Finon, D. [Centre National de la Recherche Scientifique (CNRS), CIRED (EHESS et CNRS), 75 - Paris (France)

2005-06-01

As nuclear orders are picking up a little, there are strengths competing against one another in the world industry of reactors, an industry that has been deeply affected for twenty years, by the smallness of the market and the reorganization of the electromechanical industry. Competition remains particularly difficult, even though, in terms of exports, national markets in industrialized countries such as the American market and European market are now open to foreign newcomers. One of the reasons of the difficulty is the increased commercial competition based on advanced reactor techniques untested due to strong faith in technology leading to forget the learning difficulties of older reactor types. On a narrow market, demanding and with very specific political interference, the reasoning is not like on an ordinary capital equipment market. Each builder tries to sell by relying on the assets it has in addition to the offered price and related services: industrial reputation and experience that play confusedly when untested advanced reactors are competing with one another, credit terms offered by the State and the government's influence on the market of emerging economies, the backing o the State's financial insurance in the event of risks taken in the sale of turnkey untested reactors. In the competition of the five manufacturers in the export market, American builders do not seem to have the best place, though even the leading position of Framatome ANP shows some limits. (author)

Measurements with a Pulsed and Modulated Source in a Reactor; Mesures au Moyen d'une Source Pulsee et Modulee dans un Reacteur; Izmereniya v reaktore s pomoshch'yu impul'snogo i moduliruemogo is tochnika; Mediciones Efectuadas en Reactor con una Fuente Pulsada y Modulada

Energy Technology Data Exchange (ETDEWEB)

Rotter, W. [Centre d' Etude de l' Energie Nucleaire, Mol (Belgium)

1965-10-15

generateur: le temps de mesure est donc minimum. Les observations enregistrees sur bande perforee sont depouillees par une calculatrice numerique. (author) [Spanish] Los laboratorios de investigacion Philips han construido un generador neutronico de flujo variable en funcion del tiempo. Con una serie de mediciones efectuadas en el reactor BRO2 en estado subcrftico, se ha demostrado su utililidad practica en la esfera de la fisica de los reactores. El funcionamiento del generador es muy flexible debido a su alta estabilidad, a la posibilidad de variar bruscamente la intensidad neutronica, y de pulsar el flujo o modularlo de manera sinusoidal. El generador permite determinar la reactividad ({rho} = {Delta}k/{beta}) y la vida media de los neutrones ( Script-Small-L /{beta}) segun varios metodos independientes. Es posible proceder a una comparacion exacta de esos metodos, dado que pueden aplicarse sin modificar las condiciones de medicion. El autor ha calculado los siguientes valores: a) p, sobre la base de los neutrones retardados, por reduccion instantanea del flujo neutronico; b) p, sobre la base de los neutrones inmediatos, por impulsos neutronicos; c) Script-Small-L /{beta}, combinando 1) y 2), cuando 0, 5 dolares < {rho} < 2 dolares, y d) Script-Small-L /{beta}, sobre la base de la funcion de transferencia del reactor para una fuente modulada. En la memoria se examinan las funciones de transferencia correspondientes a un oscilador de reactividad y a una fuente de modulacion sinusoidal. Se demuestra que es posible medir Script-Small-L /{beta}, cuando 0,1 dolar < {rho} < 10 dolares utilizando una fuente modulada. Por el mismo metodo se obtiene tambien la reactividad partiendo de la razon neutrones inmediatos/neutrones retardados para una frecuencia optima que es practicamente independiente de los datos relativos a los neutrones retardados y del cociente Script-Small-L /{beta}. La precision estadistica de cada metodo puede aumentarse acumulando un gran numero de ciclos en el

Automated reactor protection testing saves time and avoids errors

International Nuclear Information System (INIS)

Raimondo, E.

1990-01-01

When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

Conceptual design of the advanced marine reactor MRX

International Nuclear Information System (INIS)

1991-02-01

Design studies on the advanced marine reactors have been done continuously since 1983 at JAERI in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type PWR, built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study. (author)

DESIGN SAFETY PROBLEMS OF NUCLEAR REACTORS IN SPACE FOR ELECTRICAL POWER

Energy Technology Data Exchange (ETDEWEB)

Pickler, D A

1963-06-15

A general treatment is presented of some of the problems in the design safety of reactors which are to be operated in space. The basic requirements of these reachigh temperatures. The usual concept of a space reactor is described briefly, and the hazards of an assumed unmanned vehicle with an enriched-U-fueled reactor are examined during its launching, orbit, and reentry. Graphs are given for the dose vs distance downwind for an excursion of 100 Mw-sec, for the activity vs time after shutdown of a reactor which has been operated for 5 yr at 100 kw(t), and for the altitude vs orbital lifetime. Apparent conflicts between the basic requirements are discussed. (D.L.C.)

Hanford B Reactor Building Hazard Assessment Report

International Nuclear Information System (INIS)

Griffin, P. W.

1999-01-01

The 105-B Reactor (hereinafter referred to as B Reactor) is located in the 100 Area of the Hanford Site near Richland, Washington. The B Reactor is one of nine plutonium production reactors that were constructed in the 1940s during the Cold War Era. Construction of the B Reactor began June 7, 1943, and operation began on September 26, 1944. The Environmental Restoration Contractor was requested by RL to provide an assessment/characterization of the B Reactor building to determine and document the hazards that are present and could pose a threat to the environment and/or to individuals touring the building. This report documents the potential hazards, determines the feasibility of mitigating the hazards, and makes recommendations regarding areas where public tour access should not be permitted

The reactor kinetics code tank: a validation against selected SPERT-1b experiments

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

The analysis for inventory of experimental reactor high temperature gas reactor type

International Nuclear Information System (INIS)

Sri Kuntjoro; Pande Made Udiyani

2016-01-01

Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

L’ethnologue et l’assignation, sur le terrain de la parenté

Directory of Open Access Journals (Sweden)

Frédérique Fogel

2009-03-01

Full Text Available L’ethnologue et l’assignation, sur le terrain de la parenté. Sur mon terrain en Nubie égyptienne, j’ai occupé plusieurs positions cohérentes et contradictoires en rapport direct avec les objets principaux de ma recherche, parenté et migration. J’étais la « sœur du couple » qui m’hébergeait, la tante bilatérale des enfants. J’étais aussi cousine, fille, nièce…, en fonction de mes interlocuteurs/trices, de nos âges relatifs, de nos appartenances générationnelles, de notre genre identique ou opposé. J’appartenais à une fratrie, à un lignage, à une tribu, tout en étant étrangère et chercheur. Je reconstitue, dans cet article, mon parcours dans la parenté nubienne pour montrer de quelles manières les rôles successivement ou parallèlement endossés m’ont permis d’aborder les questions qui m’intéressaient. Ce qui revient, en regard, à rendre compte du jeu d’assignations adroitement mis en place par mes hôtes pour me guider sur le terrain de la parenté.The ethnologist and the assignment, in the field of kinship. In my fieldwork in Egyptian Nubia, I have occupied several consistent and contradictory positions in direct relation with the main subjects of my research: kinship and migration. I was the “sister of the couple” who put me up, the children’s bilateral aunt. I was also cousin, daughter, niece, depending on the person with whom I was speaking, our relative ages, our generations, and whether we were of the same or opposite sexes. I belonged to a group of siblings, a lineage and a tribe while at the same time being a foreigner and a researcher. In this article, I trace my journey through Nubian kinship, to show how the roles I assumed, successively or in parallel, enabled me to approach the questions that interested me. From another perspective, this means giving an account of the series of assignments skilfully introduced by my hosts to guide me through the field of kinship.

La estrategia de liderazgo regional de la India a través de la cooperación sur-sur (2003-2012)

OpenAIRE

Jaime Garzón, Carlos Felipe

2014-01-01

La presente investigación tiene como objetivo analizar en qué medida la estrategia de liderazgo regional de la India ha sido impulsada a través de los programas y proyectos de cooperación sur-sur ofrecidos por este país en el periodo de 2003-2012. De igual forma se pretende indagar sobre el papel histórico que ha jugado la India en el establecimiento y posterior evolución de esta nueva forma de cooperación que ha sido vista por la mayoría de los académicos como un complemento de la cooperació...

Enjeux de territoires sur une frontière méconnue

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Françoise Grenand

2012-11-01

Full Text Available « J’ai travaillé dans le temps sur le chantier d’un pont sur le Zambèze. Je me suis toujours demandé pourquoi ils avaient décidé de faire un pont à cet endroit : c’est la même chose sur les deux rives… ». Humphrey Bogart, in African Queen, de John Huston, 1951.Il y a encore quatre décennies, la Guyane ignorait superbement qu’elle eût un voisin nommé Brésil. Une frontière en commun ? Et même la plus longue frontière terrestre de la France ? Les deux seules librairies du département ne proposai...

SurF: an innovative framework in biosecurity and animal health surveillance evaluation.

Science.gov (United States)

Muellner, Petra; Watts, Jonathan; Bingham, Paul; Bullians, Mark; Gould, Brendan; Pande, Anjali; Riding, Tim; Stevens, Paul; Vink, Daan; Stärk, Katharina Dc

2018-05-16

Surveillance for biosecurity hazards is being conducted by the New Zealand Competent Authority, the Ministry for Primary Industries (MPI) to support New Zealand's biosecurity system. Surveillance evaluation should be an integral part of the surveillance life cycle, as it provides a means to identify and correct problems and to sustain and enhance the existing strengths of a surveillance system. The surveillance evaluation Framework (SurF) presented here was developed to provide a generic framework within which the MPI biosecurity surveillance portfolio, and all of its components, can be consistently assessed. SurF is an innovative, cross-sectoral effort that aims to provide a common umbrella for surveillance evaluation in the animal, plant, environment and aquatic sectors. It supports the conduct of the following four distinct components of an evaluation project: (i) motivation for the evaluation, (ii) scope of the evaluation, (iii) evaluation design and implementation and (iv) reporting and communication of evaluation outputs. Case studies, prepared by MPI subject matter experts, are included in the framework to guide users in their assessment. Three case studies were used in the development of SurF in order to assure practical utility and to confirm usability of SurF across all included sectors. It is anticipated that the structured approach and information provided by SurF will not only be of benefit to MPI but also to other New Zealand stakeholders. Although SurF was developed for internal use by MPI, it could be applied to any surveillance system in New Zealand or elsewhere. © 2018 2018 The Authors. Transboundary and Emerging Diseases Published by Blackwell Verlag GmbH.

Los rubios. Recherche sur les différents états de la mémoire et sur la reconstruction d’une identité

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Marta Mariasole Raimondi

2010-11-01

Full Text Available Los rubios, Argentina, 2003. Réalisation : Albertina Carri. Interprètes : Analìa Couceyro, Albertina Carri. Durée : 89 min. Noir & blanc et couleurs. Langue espagnole. Introduction Mélange de souvenirs, d’anecdotes, de fragments et d’imagination. Los rubios [trad. Les Blonds] ne se présente pas comme un film de dénonciation ou comme un documentaire historique sur les detenidos-desparecidos. Dans ce film, Albertina Carri propose plutôt sa vision personnelle sur la disparition de ses parents. L...

Operation and utilization of Indonesia Research Reactors

International Nuclear Information System (INIS)

Kuntoro, Iman; Sujalmo, Saiful; Tarigan, Alim

2004-01-01

For supporting the R and D in nuclear science and technology and its application, BATAN own and operate three research reactors namely, TRIGA-2000, KARTINI and RSG-GAS having thermal power of 2 MW, 100 kW and 30 MW respectively. The main features, operation and utilization progress of the reactors are described in this report. (author)

Sur la plurifonctionnalité du discours direct

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Cigada Sara

2012-07-01

Full Text Available La comparaison entre les résultats de nombreux travaux sur le dialogue dans le texte littéraire, sur l’attestation linguistique de la subjectivité, sur la fonction argumentative des émotions dans le discours et sur les effets de polyphonie, suggère que la structure sémiotique et linguistique du discours direct (DD se trouve au croisement stratégique de plusieurs axes de la construction discursive. Nous étudions donc l’insertion du DD dans le discours (cf. Rosier 2008: Le discours rapporté en français; Kerbrat 2005: Le Discours en interaction et Id. 2008: Le Dialogue comme objet d’analyse linguistique; Maingueneau 2010: Manuel de linguistique pour le texte littéraire en tant que phénomène de rupture sémiotique (Genette 1972: Figures III et Id. 1983: Nouveau discours du récit, dans ses fonctions discursives plurielles, tantôt émotives (Tannen 1989: Talking Voices; Plantin-Traverso-Vosghanian 2008: Parcours des émotions en interaction, tantôt argumentatives (Doury 2001: La Fonction argumentative des échanges rapportés; Stati 1990: Le transphrastique. Du point de vue méthodologique, nous nous proposons de revisiter empiriquement, par l'étude de corpus, les traits linguistiques structuraux qui caractérisent l’insertion du DD dans un récit à l'écrit, en les comparant systématiquement aux traits de l’insertion du DD dans un récit à l'oral. Une analyse parallèle est possible en ce qui concerne les fonctions discursives, que le DD typiquement déroule dans les récits. Les fonctions du DD décrites à partir de l'étude des corpus sont plurielles: on reconnaît des fonctions fortement argumentatives d'autres plus typiquement narratives, tandis que d'autres encore amalgament les deux fonctions. Le « contrat de littéralité », qui selon Genette ne porterait jamais que sur la teneur du discours, doit donc être fortement nuancé selon les contextes, tandis que l'effet de sens le plus directement lié à la

Effets des electrons secondaires sur l'ADN

Science.gov (United States)

Boudaiffa, Badia

Les interactions des electrons de basse energie (EBE) representent un element important en sciences des radiations, particulierement, les sequences se produisant immediatement apres l'interaction de la radiation ionisante avec le milieu biologique. Il est bien connu que lorsque ces radiations deposent leur energie dans la cellule, elles produisent un grand nombre d'electrons secondaires (4 x 104/MeV), qui sont crees le long de la trace avec des energies cinetiques initiales bien inferieures a 20 eV. Cependant, il n'y a jamais eu de mesures directes demontrant l'interaction de ces electrons de tres basse energie avec l'ADN, du principalement aux difficultes experimentales imposees par la complexite du milieu biologique. Dans notre laboratoire, les dernieres annees ont ete consacrees a l'etude des phenomenes fondamentaux induits par impact des EBE sur differentes molecules simples (e.g., N2, CO, O2, H2O, NO, C2H 4, C6H6, C2H12) et quelques molecules complexes dans leur phase solide. D'autres travaux effectues recemment sur des bases de l'ADN et des oligonucleotides ont montre que les EBE produisent des bris moleculaires sur les biomolecules. Ces travaux nous ont permis d'elaborer des techniques pour mettre en evidence et comprendre les interactions fondamentales des EBE avec des molecules d'interet biologique, afin d'atteindre notre objectif majeur d'etudier l'effet direct de ces particules sur la molecule d'ADN. Les techniques de sciences des surfaces developpees et utilisees dans les etudes precitees peuvent etre etendues et combinees avec des methodes classiques de biologie pour etudier les dommages de l'ADN induits par l'impact des EBE. Nos experiences ont montre l'efficacite des electrons de 3--20 eV a induire des coupures simple et double brins dans l'ADN. Pour des energies inferieures a 15 eV, ces coupures sont induites par la localisation temporaire d'un electron sur une unite moleculaire de l'ADN, ce qui engendre la formation d'un ion negatif transitoire

Conceptual design study of small lead-bismuth cooled reactor

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Hori, Toru; Kida, Masanori; Konomura, Mamoru

2004-11-01

In phase 2 of the feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550degC and the reactor electric output increases from 150 MWe to 165 MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal. (author)

Research reactor RB, technical characteristics and experimental possibilities

International Nuclear Information System (INIS)

Sotic, O.; Vranic, S.

1978-01-01

Nuclear research reactor RB tn the Nuclear Engineering Laboratory at the Institute of Nuclear Sciences 'Boris Kidric' in Vinca is the first reactor system built in Yugoslavia in 1958. In this report, the basic technical characteristics of this reactor are described, as well as the experimental possibilities it offers to the users. Its relatively simple construction and flexibility enables direct measurements of a series of physical parameters, and the absence of the biological protection shield makes it very useful for Various biological and other irradiations and dosimetric measurements Where strong neutron source is required. (author) [sr

Reactor D and D at Argonne National Laboratory - lessons learned

International Nuclear Information System (INIS)

Fellhauer, C. R.

1998-01-01

This paper focuses on the lessons learned during the decontamination and decommissioning (D and D) of two reactors at Argonne National Laboratory-East (ANL-E). The Experimental Boiling Water Reactor (EBWR) was a 100 MW(t), 5 MSV(e) proof-of-concept facility. The Janus Reactor was a 200 kW(t) reactor located at the Biological Irradiation Facility and was used to study the effects of neutron radiation on animals

Potential market and characteristics of low-temperature reactors

International Nuclear Information System (INIS)

Lerouge, B.

1975-01-01

The low-temperature (100 to 200 deg C) heat market for industrial applications and district heating is very important. Two main studies have been developed: a swimming pool reactor delivering water at 110 deg C and a prestressed concrete vessel reactor delivering water at 200 deg C [fr

Power reactor noise

International Nuclear Information System (INIS)

Thie, J.A.

1981-01-01

Noise analysis is a growing field that offers advantages such as simplicity, low cost, and natural multivariable interactions. A major advantage, continuous and undisturbed monitoring, supplies a means of obtaining early warnings of possible reactor malfunctions, thus preventing further complications by alerting opeators to a problem - and aiding in the diagnosis of that problem - before it demands major repairs. Dr. Thie hopes to further, through detailed explanations and over 70 illustrations, the acceptance of the use of noise analysis by the nuclear utility industry. Following an introductory chapter, the theoretical basis for the various methods of noise analysis is explained, and full chapters are devoted to the fundamentals of statistics for time-domain analysis and Fourier series and related topics for frequency-domain analysis. General experimental techniques and associated theoretical considerations are reviewed, leading to discussions of practical applications in the latter half of the book. Besides chapters giving examples of neutron noise and acoustical noise, chapters are also devoted to extensive examples from pressurized water reactor and boiling water reactor power plants

100 Areas technical activities report, engineers - March 1951

Energy Technology Data Exchange (ETDEWEB)

1951-04-01

This is the monthly 100 areas technical activities report from the engineering division for the month of March 1951. It reports on engineering activities related directly to the different production reactors, and gives progress reports on various engineering projects which are in development by the engineering group.

The ESKOM pebble bed modular reactor

International Nuclear Information System (INIS)

Gittus, J.H.

1999-01-01

An audit has been made of the design, construction, safety, economics and marketability of the ESKOM pebble bed modular reactor (PBMR). In this paper that audit is briefly summarized. The principal conclusions of the audit are as follows. The design is sound. It is a logical development of the designs proposed for other, modern, high-temperature gas-cooled reactors. More than 80% of the cost of constructing and commissioning a series of PBMRs would be spent in South Africa. The PBMR is much safer than existing nuclear power reactors and for many practical purposes it may be treated as a conventional chemical plant. The PBMR is economically competitive with thermal power stations. There is a substantial global market for the PBMR. (author)

GSTAR-SUR Modeling With Calendar Variations And Intervention To Forecast Outflow Of Currencies In Java Indonesia

Science.gov (United States)

Akbar, M. S.; Setiawan; Suhartono; Ruchjana, B. N.; Riyadi, M. A. A.

2018-03-01

Ordinary Least Squares (OLS) is general method to estimates Generalized Space Time Autoregressive (GSTAR) parameters. But in some cases, the residuals of GSTAR are correlated between location. If OLS is applied to this case, then the estimators are inefficient. Generalized Least Squares (GLS) is a method used in Seemingly Unrelated Regression (SUR) model. This method estimated parameters of some models with residuals between equations are correlated. Simulation study shows that GSTAR with GLS method for estimating parameters (GSTAR-SUR) is more efficient than GSTAR-OLS method. The purpose of this research is to apply GSTAR-SUR with calendar variation and intervention as exogenous variable (GSTARX-SUR) for forecast outflow of currency in Java, Indonesia. As a result, GSTARX-SUR provides better performance than GSTARX-OLS.

Immobilization of Cesium Traps from the BN-350 Fast Reactor (Aktau, Kazakhstan)

Energy Technology Data Exchange (ETDEWEB)

J. A. Michelbacher; C. Knight; O. G. Romanenko; I. L. Tazhibaeva; I. L. Yakovlev; A. V. Rovneyko; V. I. Maev; D. Wells; A. Herrick

2011-03-01

During BN-350 reactor operations and also during the initial stages of decommissioning, cesium traps were used to decontaminate the reactor’s primary sodium coolant. Two different types of carbon-based trap were used – the MAVR series, low ash granulated graphite adsorber (LAG) contained in a carrier designed to be inserted into the reactor core during shutdown; and a series of ex-reactor trap accumulators(TAs) which used reticulated vitreous carbon (RVC) to reduce Cs-137 levels in the sodium after final reactor shutdown. In total four MAVRs and seven TAs were used at BN-350 to remove an estimated cumulative 755 TBq of cesium. The traps, which also contain residual sodium, need to be immobilized in an appropriate way to allow them to be consigned as waste packages for long term storage and, ultimately, disposal. The present paper reports on the current status of the implementation phase, with particular reference to the work done to date on the trap accumulators, which have the most similarity with the cesium traps used at other reactors.

SP-100 Fuel Pin Performance: Results from Irradiation Testing

Science.gov (United States)

Makenas, Bruce J.; Paxton, Dean M.; Vaidyanathan, Swaminathan; Marietta, Martin; Hoth, Carl W.

1994-07-01

A total of 86 experimental fuel pins with various fuel, liner, and cladding candidate materials have been irradiated in the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF) reactor as part of the SP-100 fuel pin irradiation testing program. Postirradiation examination results from these fuel pins are key in establishing performance correlations and demonstrating the lifetime and safety of the reactor fuel system. This paper provides a brief description of the in-reactor fuel pin tests and presents the most recent irradiation data on the performance of wrought rhenium (Re) liner material and high density UN fuel at goal burnup of 6 atom percent (at. %). It also provides an overview of the significant variety of other fuel/liner/cladding combinations which were irradiated as part of this program and which may be of interest to more advanced efforts.

SP-100 technology scales from kilowatts to megawatts

International Nuclear Information System (INIS)

Deane, N.A.; Protsik, R.; Marcille, T.F.; Hoover, D.G.

1992-01-01

System level design studies of space applications ranging in power from 77 kWt to 200 MWt have indicated no practical limit to the thermal power that can be reliably generated by a space reactor system based on the technologies being developed in the SP-100 program. These technologies include uranium nitride fuel, PWC-11/rhenium bonded fuel cladding, PWC-11 structural material for the lithium coolant boundary, electromagnetic coolant pumps, safety and reactivity control drive mechanisms, sensors, shielding materials, etc. at operating temperatures up to 1400K. In this paper the physical arrangements and characteristics of the nuclear reactor materials are described

A comparative study of time series modeling methods for reactor noise analysis

International Nuclear Information System (INIS)

Kitamura, Masaharu; Shigeno, Kei; Sugiyama, Kazusuke

1978-01-01

Two modeling algorithms were developed to study at-power reactor noise as a multi-input, multi-output process. A class of linear, discrete time description named autoregressive-moving average model was used as a compact mathematical expression of the objective process. One of the model estimation (modeling) algorithms is based on the theory of Kalman filtering, and the other on a conjugate gradient method. By introducing some modifications in the formulation of the problem, realization of the practically usable algorithms was made feasible. Through the testing with several simulation models, reliability and effectiveness of these algorithms were confirmed. By applying these algorithms to experimental data obtained from a nuclear power plant, interesting knowledge about the at-power reactor noise was found out. (author)

Current limiting experiment with 600 V/100A rectification type superconducting fault current limiter; 600 V-100A kyu seiryugata chodendo genryuki no genryu shiken

Energy Technology Data Exchange (ETDEWEB)

Matsuzaki, J.; Tsurunaga, K.; Urata, M. [Toshiba Corp., Tokyo (Japan); Okuma, T.; Sato, Y.; Iwata, Y. [Tokyo Electric Power Co., Inc., Tokyo (Japan)

1999-06-07

The rectification type current limiter with the current-limiting system of the new type which combined rectifier circuits with the direct current reactor has been proposed until now, and it has succeeded in the current-limiting test by the normal conduction reactor by the 6.6kV class model vessel. Since the loss of the conductor becomes fundamentally the zero, in the same current limiter, by using superconducting wire rod, because direct current always flows in the reactor, making into low-loss becomes possible. In this report, this paper describes cut-off characteristic of 600V/100A rectification type superconductive current limiter using the metal type superconductive conductor. (NEDO)

EPR (European Pressurized Reactor)

International Nuclear Information System (INIS)

2015-01-01

This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

Dans le labyrinthe de verre. La négociation sur l'effet de serre

OpenAIRE

Hourcade , Jean Charles

2002-01-01

National audience; En parcourant les étapes principales de la « négociation climat », on fait apparaître le jeu de miroirs qui se déroule entre Europe et États-Unis, dans un contexte historique caractérisé par les débats sur la mondialisation, la volonté de leadership de l'Union européenne sur ce thème, la montée aux États-Unis d'inquiétudes sur le maintien de leur suprématie.On montre aussi pourquoi ce jeu, après avoir masqué des possibilités réelles de compromis, a débouché en 2000 sur l'éc...

Funciones agua rendimiento para 14 cultivos agrícolas en condiciones del sur de La Habana

OpenAIRE

González Robaina, Felicita; Herrera Puebla, Julián; López Seijas, Teresa; Cid Lazo, Greco

2013-01-01

El estudio de las funciones agua rendimiento y su uso dentro de la planificación del uso del agua es una vía importante para trazar estrategias de manejo que contribuyan al incremento en la producción agrícola. Utilizando los datos de consumo de agua, agua aplicada por riego, precipitaciones y los rendimientos obtenidos en más de 100 experimentos de campo realizados fundamentalmente en suelo Ferralítico Rojo de la zona sur de La Habana y con ayuda de herramientas de análisis de regresión en e...

Sur les lherzolites et ophites des Pyrénées

NARCIS (Netherlands)

Zwart, H.J.

1953-01-01

Depuis longtemps on trouve dans la littérature sur la géologie des Pyrénées des discussions sur les lherzolites et les ophites, concernant leur origine, leur âge et leur mode de formation. En général ces discussions ont rendu difficile l'éclaircissement de ce problème et un résumé de toute la

L'incidence des pesticides sur les producteurs de pommes de terre ...

International Development Research Centre (IDRC) Digital Library (Canada)

27 janv. 2011 ... De 1990 à 1993, lui et Charles Crissman du Centre international de la pomme de terre (CIP) en Équateur ont participé à une étude appelée Tradeoffs qui portait sur les effets des pesticides sur la santé, l'environnement et la productivité. Tous deux sont chargés de la direction des recherches du projet ...

Etude des répercussions de la pollution industrielle sur la riziculture ...

African Journals Online (AJOL)

Les industries spécialisées dans le textile déversent leurs eaux usées dans le ... sur les sols, les compositions chimiques des plants de riz et sur la production. ... The effluents quality varies on day, the pH from 3.9 to 10.6 and the electrical ...

Safety in decommissioning of research reactors

International Nuclear Information System (INIS)

1986-01-01

This Guide covers the technical and administrative considerations relevant to the nuclear aspects of safety in the decommissioning of reactors, as they apply to the reactor and the reactor site. While the treatment, transport and disposal of radioactive wastes arising from decommissioning are important considerations, these aspects are not specifically covered in this Guide. Likewise, other possible issues in decommissioning (e.g. land use and other environmental issues, industrial safety, financial assurance) which are not directly related to radiological safety are also not considered. Generally, decommissioning will be undertaken after planned final shutdown of the reactor. In some cases a reactor may have to be decommissioned following an unplanned or unexpected event of a series or damaging nature occurring during operation. In these cases special procedures for decommissioning may need to be developed, peculiar to the particular circumstances. This Guide could be used as a basis for the development of these procedures although specific consideration of the circumstances which create the need for them is beyond its scope

Inertia-confining thermonuclear molten salt reactors

International Nuclear Information System (INIS)

Furukawa, Kazuo; Yamanaka, Chiyoe; Nakai, Sadao; Imon, Shunji; Nakajima, Hidenori; Nakamura, Norio; Kato, Yoshio.

1984-01-01

Purpose: To increase the heat generating efficiency while improving the reactor safety and thereby maintaining the energy balance throughout the reactor. Constitution: In an inertia-confining type D-T thermonuclear reactor, the blanket is made of lithium-containing fluoride molten salts (LiF.BeF 2 , LiF.NaF.KF, LiF.KF, etc) which are cascaded downwardly in a large thickness (50 - 100 cm) along the inner wall of the thermonuclear reaction vessel, and neutrons generated by explosive compression are absorbed to lithium in the molten salts to produce tritium, Heat transportation is carried out by the molten salts. (Ikeda, J.)

Limited field investigation report for the 100-DR-1 Operable Unit

International Nuclear Information System (INIS)

1994-06-01

This limited field investigation (LFI) report summarizes the data collection and analysis activities conducted during the 100-DR-1 Source Operable Unite LFI and the associated qualitative risk assessment (QRA), and makes recommendations on the continued candidacy of high-priority sites for interim remedial measures (IRM). The results and recommendations presented in this report are generally independent of future land use scenarios. The 100-DR-1 Operable Unit is one of four operable units associated with the 100 D/DR Area at the Hanford Site. The 100-DR-1 Operable Unit encompasses approximately 1.5 km 2 (0.59 mi 2 ) and is located immediately adjacent to the Columbia River shoreline. In general, it contains waste facilities associated with the original plant facilities constructed to support D Reactor facilities, as well as cooling water retention basin systems for both D and DR Reactors. The 100-DR-1 LFI began the investigative phase of the remedial investigation for a select number of high-priority sites. The LFI was performed to provide additional data needed to support selection, design and implementation of IRM, if needed. The LFI included data compilation, nonintrusive investigations, intrusive investigations, summarization of 100 Area aggregate studies, and data evaluation

Supply of enriched uranium for research reactors

Energy Technology Data Exchange (ETDEWEB)

Mueller, H. [NUKEM GmbH, Alzenau (Germany)

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

Supply of enriched uranium for research reactors

International Nuclear Information System (INIS)

Mueller, H.

1997-01-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA's ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future

SP-100 ground engineering system at Hanford. Revision 1

International Nuclear Information System (INIS)

Ethridge, J.L.

1985-12-01

The SP-100 reactor is intended to provide a reliable power source for space applications. The reactor development program includes a ground test of the reactor systems to demonstrate that reliability and safety issues have been resolved. The use of an existing containment structure provides a unique facility with large safety margins and ample space. Preliminary seismic analysis shows that current site earthquake criteria can be met. The building is currently utilized to house engineering personnel, and the containment area is in use as an assembly facility. Only minimal activity is required to activate major support systems. All of the principal support facilities are in close proximity to the proposed test site. The various systems and facilities and their status are identified

SP-100 Ground Engineering System at Hanford. Volume 1

International Nuclear Information System (INIS)

1985-01-01

The SP-100 reactor is intended to provide a reliable power source for space applications. The reactor development program includes a ground test of the reactor systems to demonstrate that reliability and safety issues have been resolved. The use of an existing containment structure provides a unique facility with large safety margins and ample space. Preliminary seismic analysis shows that current site earthquake criteria can be met. The building is currently utilized to house engineering personnel, and the containment area is in use as an assembly facility. Only minimal activity is required to activate major support systems. All of the principal support facilities are in close proximity to the proposed test site. The various systems and facilities and their status are identified

SP-100 ground engineering system at Hanford. Volume 2

International Nuclear Information System (INIS)

1986-01-01

The SP-100 reactor is intended to provide a reliable power source for space applications. The reactor development program includes a ground test of the reactor systems to demonstrate that reliability and safety issues have been resolved. The use of an existing containment structure provides a unique facility with large safety margins and ample space. Preliminary seismic analysis shows that current site earthquake criteria can be met. The building is currently utilized to house engineering personnel, and the containment area is in use as an assembly facility. Only minimal activity is required to activate major support systems. All of the principal support facilities are in close proximity to the proposed test site. The various systems and facilities and their status are identified

A Modification of Gamma Surveymeter Dosemeter 3007A for Monitoring Use Ethernet by PLC T100MD Series

International Nuclear Information System (INIS)

Ikhsan Shobari; Subchan, M.; Syahrudin Yusuf; Sutomo Budihardjo

2010-01-01

It has been modified a gamma surveymeter Dosemeter 3007A. The Surveymeter represents analogous surveymeter, so that an interface for data acquisition is required. Acquisition system from surveymeter is added to the voltage amplifier module from 0 - 200 mV to 0 - 5 V. This voltage value will represent of doses 0 - 5 mR/hour. Hereinafter the analogous signal 0 - 5 V as signal of input to peripheral of PLC T100MD series. Data in the form of processed analogous signal presented at local display of PLC. For long distance monitoring, data have been sent to a computer from PLC by ethernet. After this modification, the surveymeter can be used to monitor from long distance. By using Internet Service Provider, monitoring can be done at any time and any where as long as network internet is available. (author)

Advanced Safeguards Approaches for New Fast Reactors

International Nuclear Information System (INIS)

Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

2007-01-01

This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to 'breed' nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and 'burn' actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is 'fertile' or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing 'TRU'-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II 'EBR-II' at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line--a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors

Core and shielding analysis of the SCM-100

International Nuclear Information System (INIS)

Olson, A. P.

2002-01-01

It is widely accepted that an intense neutron source can be produced in a suitable target by spallation neutrons generated by a high-current high-energy proton beam. Typical beam energy for such an accelerator is 400 to 2000 MeV. A conventional critical reactor can readily be replaced by a ''sub-critical reactor'' driven by this source. A 5 MW proton beam at 600 MeV can drive a sub-critical reactor to 100 MWt. The accelerator and the associated plant support equipment at these design specifications are complex systems, but they are well within recent technology. The purpose of this study was to examine core design and shielding design issues for a 100 MWt sodium-cooled fast-spectrum Sub-Critical Multiplier (SCM-100) based on LMFBR technology, but driven by an intense neutron source created by spallation reactions. SCM-100 is a component of the Accelerator Driven Test Facility. In this report we provide an overview of the SCM-100 concept. Two designs were investigated: (1) a vertical entry for the beam on the axial centerline; and (2) an inclined entry design where the core is ''C'' shaped and the beam enters the side of the target at an angle of 32 degrees. A brief overview of relevant shielding design data from EBR-II is also provided. The key result of this report is that the inclined entry design cannot achieve design objectives for radial power peaking. Consequently it cannot achieve design objectives for peak neutron flux. Axial power peaking factors are controlled by the axial fuel height and the axial reflector properties. These dimensions and compositions are very similar in SCM-100 to those of EBR-II. EBR-II had an axial power peaking factor of 1.093, and a radial power peaking factor of about 1.46. The radial power peaking of SCM-100 with the inclined entry is too extreme at 2.15, and cannot be made acceptable by modifying the size and detailed shape of the ''C'' shaped core and reflector. The axial power peaking of SCM-100 is very close to that of EBR

The physics design of EBR-II; Physique du reacteur EBR-II; Fizicheskij raschet ehksperimental'nogo reaktora - razmnozhitelya EVR-II; Aspectos fisicos del reactor EBR-II

Energy Technology Data Exchange (ETDEWEB)

Loewenstein, W. B. [Argonne National Laboratory, Argonne, IL (United States)

1962-03-15

notamment l'introduction de taux de reactivite normaux et anormaux, les consequences des effets supposes de reactivite, a partir du comportement physique de l'alliage combustible et de la structure du reacteur, ainsi que par extrapolation des experiences faites sur TREAT au systeme EBR-II. Il examine le probleme de la fusion du coeur de EBR-II. (author) [Spanish] La memoria informa sobre los calculos del comportamiento estatico, dinamico y a largo plazo de la reactividad del reactor reproductor experimental EBR-II, asi como sobre los resultados y analisis de los experimentos criticos en seco del EBR-II y de los experimentos simulados en el reactor de potencia cero ZPR-III. Insiste particularmente en los problemas de fisica del reactor que, en la elaboracion del proyecto, siguen a la eleccion del modelo, pero preceden a la construccion y puesta en marcha del reactor. Presenta diversos analisis del reactor desde el punto de vista de la seguridad y formula consideraciones sobre la evaluacion de los riesgos y su influencia sobre el diseno del reactor. El trabajo explica tambien la manera de emplear los datos obtenidos en los experimentos arriba citados. Estos experimentos, su analisis y sus predicciones teoricas constituyen la base para determinar el comportamiento fisico del reactor. La memoria estudia detalladamente las limitaciones inherentes a la aplicacion de los datos experimentales al funcionamiento del reactor de potencia. Ello incluye datos precisos sobre as dimensiones del cuerpo, el enriquecimiento de la aleacion combustible, o de ambos factores; el establecimiento de una reactividad adecuada para el reactor funcionando o detenido, la determinacion de la variacion de los coeficientes de reactividad en funcion de la temperatura de funcionamiento y de la potencia generadora y detalles de la distribucion de la potencia y del flujo en diversos puntos de la estructura del reactor. La memoria expone tambien el problema general que supone transferir a la verdadera geometria

E-beam heated linear solenoid reactors

International Nuclear Information System (INIS)

Benford, J.; Bailey, V.; Oliver, D.

1976-01-01

A conceptual design and system analysis shows that electron beam heated linear solenoidal reactors are attractive for near term applications which can use low gain fusion sources. Complete plant designs have been generated for fusion based breeders of fissile fuel over a wide range of component parameters (e.g., magnetic fields, reactor lengths, plasma densities) and design options (e.g., various radial and axial loss mechanisms). It appears possible that a reactor of 100 to 300 meters length operating at power levels of 1000 MWt can economically produce 2000 to 8000 kg/yr of 233 U to supply light water reactor fuel needs beyond 2000 A.D. Pure fusion reactors of 300 to 500 meter lengths are possible. Physics and operational features of reactors are described. Beam heating by classical and anomalous energy deposition is reviewed. The technology of the required beams has been developed to MJ/pulse levels, within a factor of 20 of that needed for full scale production reactors. The required repetitive pulsing appears practical

Cooperación Sur-Sur para el fortalecimiento de los laboratorios de control de medicamentos de la Comunidad del Caribe (CARICOM

Directory of Open Access Journals (Sweden)

José María Parisi

Full Text Available RESUMEN Objetivo Describir los beneficios obtenidos a través de la cooperación Sur-Sur y Triangular, como una potencial herramienta para el fortalecimiento en el control de la calidad de los medicamentos en los Laboratorios Oficiales de Control de Medicamentos (LOCM de la Región de las Américas. Métodos Estudio descriptivo del proyecto para el fortalecimiento en el control de la calidad de los medicamentos en los LOCM de la Comunidad del Caribe (CARICOM. Resultados La capacitación fue desarrollada por profesionales de la Administración Nacional de Medicamentos, Alimentos y Tecnología Médica (ANMAT de Argentina, a profesionales de Guyana, Jamaica, Surinam y Trinidad y Tobago. El proyecto contó con financiamiento del Fondo Argentino de Cooperación Sur-Sur y Triangular (FO.AR y coordinación de la Organización Panamericana de la Salud (OPS. Se revisaron los documentos de Buenas Prácticas de Laboratorio (BPL de la Organización Mundial de la Salud (OMS y de la Red Panamericana para la Armonización de la Reglamentación Farmacéutica (Red-PARF y se fortaleció el área de controles físicos químicos, principalmente en relación a medicamentos para el tratamiento de la tuberculosis, la malaria y el VIH/sida, todos de importancia estratégica para esos países. Conclusión Este tipo de colaboraciones permiten transferir experiencia, optimizar los recursos, armonizar procedimientos y regulaciones y reforzar capacidades en término de recursos humanos, y constituyen una herramienta valiosa en la reducción de las asimetrías que pudieron establecerse en diferentes áreas entre diferentes países de nuestra región.

Recommendations for a restart of Molten Salt Reactor development

International Nuclear Information System (INIS)

Moir, R. W.

2007-01-01

The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. A strong incentive for the molten salt reactor design is its good fuel utilization, good economics, amazing flexibility and promised large benefits. It can: - use thorium or uranium; o be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have a fast neutron spectrum reactor; - fission uranium isotopes and plutonium isotopes; - operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon-grade fissile fuel; - be a breeder or near breeder; - operate at temperature >1100 degree C if carbon composites are successfully employed. Enhancing 2 32U content in the uranium to over 500 pm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR is enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10% lower than that from LWRs, and 20% lower for high enriched fuel, with uncertainties of about 10%. The development cost has been estimated at about 1 B$ (e.g., a 100 M$/y base program for ten years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B$ (450 M$/y over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the

L'Avenir de la Vie sur la Terre

CERN Document Server