IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

West, C.D.

1990-05-01

Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

Conceptual design study on inertial confinement reactor ''SENRI-II''

International Nuclear Information System (INIS)

Nakamura, N.; Ouura, H.

1983-01-01

Design features of a laser fusion reactor concept SENRI-II are reviewed and discussed. A conceptual design study of the ICF reactor SENRI-II (an advanced design of SENRI-I) has been carried out over 2 years in the Research Committee of ICF Reactors, Institute of Laser Engineering, Osaka University. While the ICF reactor SENRI-I utilized a magnetic field to guide and control an inner liquid lithium flow, SENRI-II is designed to use porous metal as the liquid lithium flow guide. In the design of SENRI-II, a metal porous lithium blanket serves as the protection of a wall against fusion products and as wall per se. Because of the separation of these two functions, a high power density can be attained

EBR-II Reactor Physics Benchmark Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

Pope, Chad L. [Idaho State Univ., Pocatello, ID (United States); Lum, Edward S [Idaho State Univ., Pocatello, ID (United States); Stewart, Ryan [Idaho State Univ., Pocatello, ID (United States); Byambadorj, Bilguun [Idaho State Univ., Pocatello, ID (United States); Beaulieu, Quinton [Idaho State Univ., Pocatello, ID (United States)

2017-12-28

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

EBR-II [Experimental Breeder Reactor-II] system surveillance using pattern recognition software

International Nuclear Information System (INIS)

Mott, J.E.; Radtke, W.H.; King, R.W.

1986-02-01

The problem of most accurately determining the Experimental Breeder Reactor-II (EBR-II) reactor outlet temperature from currently available plant signals is investigated. Historically, the reactor outlet pipe was originally instrumented with 8 temperature sensors but, during 22 years of operation, all these instruments have failed except for one remaining thermocouple, and its output had recently become suspect. Using pattern recognition methods to compare values of 129 plant signals for similarities over a 7 month period spanning reconfiguration of the core and recalibration of many plant signals, it was determined that the remaining reactor outlet pipe thermocouple is still useful as an indicator of true mixed mean reactor outlet temperature. Application of this methodology to investigate one specific signal has automatically validated the vast majority of the 129 signals used for pattern recognition and also highlighted a few inconsistent signals for further investigation

Liquid metal reactor deactivation as applied to the experimental breeder reactor - II

International Nuclear Information System (INIS)

Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

1999-01-01

The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF

BER of subcarrier MPSK and MDPSK systems in atmospheric turbulence

KAUST Repository

Song, Xuegui

2015-01-01

Bit-error rate (BER) performance of subcarrier $M$-ary phase-shift keying (MPSK) and $M$-ary differential PSK (MDPSK) is analyzed for optical wireless communications over Gamma-Gamma and lognormal turbulence channels. We study the relation between the exact BER and the approximate BER, which is obtained by dividing the symbol-error rate by the number of bits per symbol, for subcarrier MPSK and MDPSK modulations. The asymptotic BER performance gap between the exact and the approximate BERs is quantified analytically through our asymptotic analyses. The accuracy of the approximate BER of both MPSK and MDPSK depends on the channel conditions. Under weak turbulence conditions, the approximate BER expression can be used to predict the system performance with high accuracy, while under strong turbulence conditions the approximate BER becomes inaccurate and can only serve as a loose lower bound of the exact BER. The asymptotic BER performance loss of MDPSK with respect to MPSK is also quantified analytically.

Multiscale Computation. Needs and Opportunities for BER Science

Energy Technology Data Exchange (ETDEWEB)

Scheibe, Timothy D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Jeremy C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-01-01

The Environmental Molecular Sciences Laboratory (EMSL), a scientific user facility managed by Pacific Northwest National Laboratory for the U.S. Department of Energy, Office of Biological and Environmental Research (BER), conducted a one-day workshop on August 26, 2014 on the topic of “Multiscale Computation: Needs and Opportunities for BER Science.” Twenty invited participants, from various computational disciplines within the BER program research areas, were charged with the following objectives; Identify BER-relevant models and their potential cross-scale linkages that could be exploited to better connect molecular-scale research to BER research at larger scales and; Identify critical science directions that will motivate EMSL decisions regarding future computational (hardware and software) architectures.

Data handling at EBR-II [Experimental Breeder Reactor II] for advanced diagnostics and control work

International Nuclear Information System (INIS)

Lindsay, R.W.; Schorzman, L.W.

1988-01-01

Improved control and diagnostics systems are being developed for nuclear and other applications. The Experimental Breeder Reactor II (EBR-II) Division of Argonne National Laboratory has embarked on a project to upgrade the EBR-II control and data handling systems. The nature of the work at EBR-II requires that reactor plant data be readily available for experimenters, and that the plant control systems be flexible to accommodate testing and development needs. In addition, operational concerns require that improved operator interfaces and computerized diagnostics be included in the reactor plant control system. The EBR-II systems have been upgraded to incorporate new data handling computers, new digital plant process controllers, and new displays and diagnostics are being developed and tested for permanent use. In addition, improved engineering surveillance will be possible with the new systems

Kultur als Ordnungshypothese über Raum?

OpenAIRE

Klüter, Helmut

2009-01-01

Inhalt: 1 Einleitung 2 Zur Attraktivität von Kultur 3 Kulturologische Methodik 4 Schnittmengen mit und Bindearme zur Kulturgeographie 5 Kritische Argumente gegen Kultur als Ordnungshypothese über Raum 6 Räumliche Abstraktion als Ordnungsstrategie über „Kultur“? 7 Zusammenfassung

Experimental studies of U-Pu-Zr fast reactor fuel pins in EBR-II [Experimental Breeder Reactor

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1988-01-01

The Integral Fast Reactor (IFR) is a generic reactor concept under development by Argonne National Laboratory. Much of the technology for the IFR is being demonstrated at the Experimental Breeder Reactor II (EBR-II) on the Department of Energy site near Idaho Falls, Idaho. The IFR concept relies on four technical features to achieve breakthroughs in nuclear power economics and safety: (1) a pool-type reactor configuration, (2) liquid sodium cooling, (3) metallic fuel, and (4) an integral fuel cycle with on-site reprocessing. The purpose of this paper will be to summarize our latest results of irradiation testing uranium-plutonium-zirconium (U-Pu-Zr) fuel in the EBR-II. 10 refs., 13 figs., 2 tabs

Technical specifications: Tower Shielding Reactor II

International Nuclear Information System (INIS)

1979-02-01

The technical specifications define the key limitations that must be observed for safe operation of the Tower Shielding Reactor II (TSR-II) and an envelope of operation within which there is reasonable assurance that these limits cannot be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

The influence of a spatial displacement of hydrogen on the reactivity and neutron flux density distribution in a ZrH-moderated reactor

International Nuclear Information System (INIS)

Doehler, J.; Bartsch, G.

1975-08-01

The effect of changes of the hydrogen concentration in uranium zirconium hydride resulting from spatially varying temperatures on the reactivity and neutron flux distribution of the BER-II core (power 2.2 MW) are shown. Furthermore, in general, the influence of the hydrogen concentration on important reactor parameters of a fuel cell of BER-II is calculated and presented. A comparison of the diffusion calculation with spatially constant hydrogen concentrations shows a decrease of the thermal neutron flux density in regions with a low hydrogen content (high temperature) and inversely an increase for high hydrogen concentrations. Furthermore, a change of the effective multiplication factor by 0.6% was found in the case of a spatially varying hydrogen concentration as compared with the calculation for a constant concentration. (orig.) [de

The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

1990-01-01

This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

Characterization of the TRIGA Mark II reactor full-power steady state

Energy Technology Data Exchange (ETDEWEB)

Cammi, Antonio, E-mail: antonio.cammi@polimi.it [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Zanetti, Matteo [Politecnico di Milano – Department of Energy, CeSNEF (Enrico Fermi Center for Nuclear Studies), via La Masa 34, 20156 Milano (Italy); Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica [University of Milano-Bicocca, Physics Department “G. Occhialini” and INFN Section, Piazza dell’Ateneo Nuovo, 20126 Milan (Italy); Magrotti, Giovanni; Prata, Michele; Salvini, Andrea [University of Pavia, Applied Nuclear Energy Laboratory (L.E.N.A.), Via Gaspare Aselli 41, 27100 Pavia (Italy)

2016-04-15

Highlights: • Full-power steady state characterization of the TRIGA Mark II reactor. • Monte Carlo and Multiphysics simulation of the TRIGA Mark II reactor. • Sub-cooled boiling effects in the TRIGA Mark II reactor. • Thermal feedback effects in the TRIGA Mark II reactor. • Experimental data based validation. - Abstract: In this paper, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor at the University of Pavia is achieved by coupling the Monte Carlo (MC) simulation for neutronics with the “Multiphysics” model for thermal-hydraulics. Neutronic analyses have been carried out with a MCNP5 based MC model of the entire reactor system, already validated in fresh fuel and zero-power configurations (in which thermal effects are negligible) and using all available experimental data as a benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core must be established. To evaluate this, a thermal-hydraulic model has been developed, using the power distribution results from the MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then entered into the MC model and a benchmark analysis is carried out to validate the model in fresh fuel and full-power configurations. An acceptable correspondence between experimental data and simulation results concerning full-power reactor criticality proves the reliability of the adopted methodology of analysis, both from the perspective of neutronics and thermal-hydraulics.

Over Twenty Years Of Experience In ITU TRIGA MARK-II Reactor

International Nuclear Information System (INIS)

Yavuz, Hasbi

2008-01-01

I.T.U. TRIGA MARK-II Training and Research Reactor, rated at 250 kW steady-state and 1200 MW pulsing power is the only research and training reactor owned and operated by a university in Turkey. Reactor has been operating since March 11, 1979; therefore the reactor has been operating successfully for more than twenty years. Over the twenty years of operation: - The tangential beam tube was equipped with a neutron radiography facility, which consists of a divergent collimator and exposure room; - A computerized data acquisition system was designed and installed such that all parameters of the reactor, which are observed from the console, could be monitored both in normal and pulse operations; - An electrical power calibration system was built for the thermal power calibration of the reactor; - Publications related with I.T.U. TRIGA MARK-II Training and Research Reactor are listed in Appendix; - Two majors undesired shutdown occurred; - The I.T.U. TRIGA MARK-II Training and Research Reactor is still in operation at the moment. (authors)

Decommissioning of TRIGA Mark II type reactor

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

IAEA fast reactor knowledge preservation initiative. Project focus: KNK-II reactor, Karlsruhe, Germany

International Nuclear Information System (INIS)

2004-08-01

This Working Material (including the attached CD-ROM) documents progress made in the IAEA's initiative to preserve knowledge in the fast reactor domain. The brochure describes briefly the context of the initiative and gives an introduction to the contents of the CD-ROM. In 2003/2004 a first focus of activity was concentrated on the preservation of knowledge related to the KNK-II experimental fast reactor in Karlsruhe, Germany. The urgency of this project was given by the impending physical destruction of the installation, including the office buildings. Important KNK-II documentation was brought to safety and preserved just in time. The CD-ROM contains the full texts of 264 technical and scientific documents describing research, development and operating experience gained with the KNK-II installation over a period of time from 1965 to 2002, extending through initial investigations, 17 years of rich operating experience, and final shutdown and decommissioning. The index to the documents on the CD-ROM is printed at the end of this booklet in chronological order and is accessible on the CD by subject index and chronological index. The CD-ROM contains in its root directory also the document 'fr c lassification.pdf' which describes the classification system used for the present collection of documents on the fast reactor KNK-II

Analysis of Topaz-II reactor performance using MCNP and TFEHX

International Nuclear Information System (INIS)

Lee, H.H.; Klein, A.C.

1993-01-01

Data reported by Russian scientist and engineers for the TOPAZ-II Space Nuclear Power is compared with analytical results calculated using the Monte Carlo Neutron and Photon (MCNP) and TFEHX computer codes. The results of these comparisons show good agreement with the TOPAZ-II neutronics, thermionic and thermal hydraulics performance. A detailed description of the TOPAZ-II reactor and of the TFE should enhance the performance of the both codes in modeling the reactor and TFE performances

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

The physics design oi EBR-II. Calculations of the static, dynamic and long-term reactivity behaviour of EBR-II are reported together with results and analysis of EBR-II dry critical and ZPR-III mock-up experiments. Particular emphasis is given to reactor-physics design problems which arise after the conceptual design is established and before the reactor is built or placed into operation. Reactor-safety analyses and hazards-evaluation considerations are described with their influence on the reactor design. The manner of utilizing the EBR-II mock-up on ZPR-III data and the EBR-II dry critical data is described. These experiments, their analysis and theoretical predictions are the basis for predetermining the physics behaviour of the reactor system. The limitations inherent in applying the experimental data to the performance of the power-reactor system are explored in some detail. This includes the specification of reactor core size and/or fuel-alloy enrichment, provisions for adequate operating and shut-down reactivity, determination of operative temperature and power coefficients of reactivity, and details of power- and flux-distribution as a function of position within the reactor structure. The overall problem of transferring information from simple idealized analytical or experimental geometry to actual hexagonal reactor geometry is described. Nuclear performance, including breeding, of the actual reactor system is compared with that of the idealized conceptual system. The long-term reactivity and power behaviour of the reactor blanket is described within the framework of the proposed cycling of the fuel and blanket alloy. Safety considerations, including normal and abnormal rates of reactivity-insertion, the implication of postulated reactivity effects based on the physical behaviour of the fuel alloy and reactor structure as well as extrapolation of TREAT experiments to the EBR-II system are analysed. The EBR-II core melt-down problem is reviewed. (author

Optical fiber communication — An overview

Indian Academy of Sciences (India)

This paper deals with the historical development of optical communication systems and their failures initially. Then the different generations in optical fiber communication along with their features are discussed. Some aspects of total internal reflection, different types of fibers along with their size and refractive index profile, ...

Design guide for category II reactors light and heavy water cooled reactors

International Nuclear Information System (INIS)

Brynda, W.J.; Lobner, P.R.; Powell, R.W.; Straker, E.A.

1978-05-01

The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems

BER of subcarrier MPSK/MDPSK modulated OWC systems in Gamma-Gamma turbulence

KAUST Repository

Song, Xuegui

2013-12-01

Bit-error rate (BER) performance of subcarrier Mary phase-shift keying (MPSK) and M-ary differential PSK (MDPSK) is analyzed for optical wireless communications in Gamma-Gamma turbulence. We study the relation between the exact BER and the approximate BER, which is obtained by dividing the symbol error rate by the number of bits per symbol, for subcarrier MPSK and MDPSK modulations. The asymptotic performance gap between the exact and the approximate BERs is quantified analytically through our asymptotic analyses. The accuracy of the approximate BER of both MPSK and MDPSK depends on the channel conditions. Under weak turbulence conditions, the approximate BER expression can be used to predict the system performance with high accuracy, while under strong turbulence conditions the approximate BER becomes inaccurate and can only serve as a loose lower bound of the exact BER. The asymptotic BER performance loss of MDPSK with respect to MPSK is also quantified analytically. © 2013 IEEE.

Utilization of Slovenian TRIGA Mark II reactor

International Nuclear Information System (INIS)

Snoj, L.; Smodis, B.

2010-01-01

TRIGA Mark II research reactor at the Jozef Stefan Institute [JSI] is extensively used for various applications, such as: irradiation of various samples, training and education, verification and validation of nuclear data and computer codes, testing and development of experimental equipment used for core physics tests at a nuclear power plant. The paper briefly describes the aforementioned activities and shows that even such small reactors are still indispensable in nuclear science and technology. (author)

The ARIES-II and ARIES-IV second-stability tokamak reactors

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.; Hasan, M.Z.; Mau, T.-K.; Sharafat, S.; Baxi, C.B.; Leuer, J.A.; McQuillan, B.W.; Puhn, F.A.; Schultz, K.R.; Wong, C.P.C.; Brooks, J.; Ehst, D.A.; Hassanein, A.; Hua, T.; Hull, A.; Mattis, R.; Picologlou, B.; Sze, D.-K.; Dolan, T.J.; Herring, J.S.; Bathke, C.G.; Krakowski, R.A.; Werley, K.A.; Bromberg, L.; Schultz, J.; Davis, F.; Holmes, J.A.; Lousteau, D.C.; Strickler, D.J.; Jardin, S.C.; Kessel, C.; Snead, L.; Steiner, D.; Valenti, M.; El-Guebaly, L.A.; Emmert, G.A.; Khater, H.Y.; Santarius, J.F.; Sawan, M.; Sviatoslavsky, I.N.; Cheng, E.T.

1992-01-01

The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. Four ARIES visions are currently planned for the ARIES program. The ARIES-I design is a DT-burning reactor based on modest extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. The ARIES-III study focuses on the potential of tokamaks to operate with D- 3 He fuel system as an alternative to deuterium and tritium. The ARIES-II and ARIES-IV designs have the same fusion plasma but different fusion-power-core designs. The ARIES-II reactor uses liquid lithium as the coolant and tritium breeder and vanadium alloy as the structural material in order to study the potential of low-activation metallic blankets. The ARIES-IV reactor uses helium as the coolant, a solid tritium-breeding material, and silicon carbide composite as the structural material in order to achieve the safety and environmental characteristic of fusion. In this paper the authors describe the trade-off leading to the optimum regime of operation for the ARIES-II and ARIES-IV second-stability reactors and review the engineering design of the fusion power cores

HYLIFE-II reactor chamber design refinements

International Nuclear Information System (INIS)

House, P.A.

1994-06-01

Mechanical design features of the reactor chamber for the HYLIFE-II inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams (Li 2 BeF 4 ) are used for shielding and blast protection of the chamber walls. The system is designed for a 6 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (>12 m/s) salt streams and also recover up to half of the dynamic head. Cost estimates for a 1 GWe and 2 GWe reactor chamber are presented

Transforming criticality control methods for EBR-II fuel handling during reactor decommissioning

International Nuclear Information System (INIS)

Eberle, C.S.; Dean, E.M.; Angelo, P.L.

1995-01-01

A review of the Department of Energy (DOE) request to decommission the Experimental Breeder Reactor-II (EBR-II) was conducted in order to develop a scope of work and analysis method for performing the safety review of the facility. Evaluation of the current national standards, DOE orders, EBR-II nuclear safeguards and criticality control practices showed that a decommissioning policy for maintaining criticality safety during a long term fuel transfer process did not exist. The purpose of this research was to provide a technical basis for transforming the reactor from an instrumentation and measurement controlled system to a system that provides both physical constraint and administrative controls to prevent criticality accidents. Essentially, this was done by modifying the reactor core configuration, reactor operations procedures and system instrumentation to meet the safety practices of ANS-8.1-1983. Subcritical limits were determined by applying established liquid metal reactor methods for both the experimental and computational validations

Corrosion problem in the CRENK Triga Mark II research reactor

International Nuclear Information System (INIS)

Kalenga, M.

1990-01-01

In August 1987, a routine underwater optical inspection of the aluminum tank housing the core of the CRENK Triga Mark II reactor, carried out to update safety condition of the reactor, revealed pitting corrosion attacks on the 8 mm thick aluminum tank bottom. The paper discuss the work carried out by the reactor staff to dismantle the reactor in order to allow a more precise investigation of the corrosion problem, to repair the aluminum tank bottom, and to enhance the reactor overall safety condition

UWMAK-II: a conceptual tokamak reactor design

International Nuclear Information System (INIS)

1975-10-01

This report describes the conceptual design of a Tokamak fusion power reactor, UWMAK-II. The aim of this study is to perform a self consistent and thorough analysis of a probable future fusion power reactor in order to assess the technological problems posed by such a system and to examine feasible solutions. UWMAK-II is a conceptual Tokamak fusion reactor designed to deliver 1716 MWe continuously and to generate 5000 MW(th) during the plasma burn. The structural material is 316 stainless steel and the primary coolant is helium. UWMAK-II is a low aspect ratio, low field design and includes a double null, axisymmetric poloidal field divertor for impurity control. In addition, a carbon curtain, made of two dimensional woven carbon fiber, is mounted on the first vacuum chamber wall to protect the plasma from high Z impurities and to protect the first wall from erosion by charged particle bombardment. The blanket is designed to minimize the inventory of both tritium and lithium while achieving a breeding ratio greater than one. This has led to a blanket design based on the use of a solid breeding material (LiAlO 2 ) with beryllium as a neutron multiplier. The lithium is enriched to 90 percent 6 Li and the blanket coolant is helium at a maximum pressure of 750 psia (5.2 x 10 6 N/m 2 ). A cell of the UWMAK-II blanket design is shown. The breeding ratio is between 1.11 and 1.19 based on one-dimensional discrete ordinates transport calculations, depending on the method of homogenization. Detailed Monte Carlo calculations, which take into account the more complicated geometry, give a breeding ratio of 1.06. The total energy per fusion is 21.56 MeV, which is fairly high

Decontamination of TRIGA Mark II reactor, Indonesia

International Nuclear Information System (INIS)

Suseno, H.; Daryoko, M.; Goeritno, A.

2002-01-01

The TRIGA Mark II Reactor in the Centre for Research and Development Nuclear Technique Bandung has been partially decommissioned as part of an upgrading project. The upgrading project was carried out from 1995 to 2000 and is being commissioned in 2001. The decommissioning portion of the project included disassembly of some components of the reactor core, producing contaminated material. This contaminated material (grid plate, reflector, thermal column, heat exchanger and pipe) will be sent to the Decontamination Facility at the Radioactive Waste Management Development Centre. (author)

The optimal control of ITU TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Can, Burhanettin

2008-01-01

In this study, optimal control of ITU TRIGA Mark-II Reactor is discussed. A new controller has been designed for ITU TRIGA Mark-II Reactor. The controller consists of main and auxiliary controllers. The form is based on Pontragyn's Maximum Principle and the latter is based on PID approach. For the desired power program, a cubic function is chosen. Integral Performance Index includes the mean square of error function and the effect of selected period on the power variation. YAVCAN2 Neutronic - Thermal -Hydraulic code is used to solve the equations, namely 11 equations, dealing with neutronic - thermal - hydraulic behavior of the reactor. For the controller design, a new code, KONTCAN, is written. In the application of the code, it is seen that the controller controls the reactor power to follow the desired power program. The overshoot value alters between 100 W and 500 W depending on the selected period. There is no undershoot. The controller rapidly increases reactivity, then decreases, after that increases it until the effect of temperature feedback is compensated. Error function varies between 0-1 kW. (author)

Adaptive robust control of the EBR-II reactor

International Nuclear Information System (INIS)

Power, M.A.; Edwards, R.M.

1996-01-01

Simulation results are presented for an adaptive H ∞ controller, a fixed H ∞ controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H ∞ controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H ∞ and classical controllers. This makes for a superior and more robust controller

Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

International Nuclear Information System (INIS)

Smith, M.S.; Ragland, W.A.

1989-01-01

The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

U-233 fuelled low critical mass solution reactor experiment PURNIMA II

International Nuclear Information System (INIS)

Srinivasan, M.; Chandramoleshwar, K.; Pasupathy, C.S.; Rasheed, K.K.; Subba Rao, K.

1987-01-01

A homogeneous U-233 uranyl nitrate solution fuelled BeO reflected, low critical mass reactor has been built at the Bhabha Atomic Research Centre, India. Christened PURNIMA II, the reactor was used for the study of the variation of critical mass as a function of fuel solution concentration to determine the minimum critical mass achievable for this geometry. Other experiments performed include the determination of temperature coefficient of reactivity, study of time behaviour of photoneutrons produced due to interaction between decaying U-233 fission product gammas and the beryllium reflector and reactor noise measurements. Besides being the only operational U-233 fuelled reactor at present, PURNIMA II also has the distinction of having attained the lowest critical mass of 397 g of fissile fuel for any operating reactor at the current time. The paper briefly describes the facility and gives an account of the experiments performed and results achieved. (author)

Scram reliability under seismic conditions at the Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Roglans, J.; Wang, C.Y.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

Mittheilungen über Scyphomedusen I

NARCIS (Netherlands)

Stiasny, G.

1921-01-01

Hiemit beabsichtige ich eine Reihe kleinerer Mittheilungen über Semaeostomeen und Rhizostomeen des Rijksmuseums van Natuurlijke Historie in Leiden zu veröffentlichen, welche die Systematik, Biologie, Entwicklungsgeschichte und Anatomie verschiedener Scyphomedusen behandeln. Die vorliegenden beiden

SNR and BER Models and the Simulation for BER Performance of Selected Spectral Amplitude Codes for OCDMA

Directory of Open Access Journals (Sweden)

Abdul Latif Memon

2014-01-01

Full Text Available Many encoding schemes are used in OCDMA (Optical Code Division Multiple Access Network but SAC (Spectral Amplitude Codes is widely used. It is considered an effective arrangement to eliminate dominant noise called MAI (Multi Access Interference. Various codes are studied for evaluation with respect to their performance against three noises namely shot noise, thermal noise and PIIN (Phase Induced Intensity Noise. Various Mathematical models for SNR (Signal to Noise Ratios and BER (Bit Error Rates are discussed where the SNRs are calculated and BERs are computed using Gaussian distribution assumption. After analyzing the results mathematically, it is concluded that ZCC (Zero Cross Correlation Code performs better than the other selected SAC codes and can serve larger number of active users than the other codes do. At various receiver power levels, analysis points out that RDC (Random Diagonal Code also performs better than the other codes. For the power interval between -10 and -20 dBm performance of RDC is better ZCC. Their lowest BER values suggest that these codes should be part of an efficient and cost effective OCDM access network in the future.

Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

International Nuclear Information System (INIS)

Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

1982-01-01

Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components

Method of power self-regulation of CFBR-II reactor based on DSP

International Nuclear Information System (INIS)

Bai Zhongxiong; Zhou Wenxiang

2007-01-01

To the control system of Power Self-regulation of CFBR-II Reactor, a new digital control scheme based on DSP has been brought forward. The TMS320F2812 DSP chip is adopted as the core controller to realize Power self-regulation of CFBR-II Reactor. In this paper, the successful program of DSP control system is introduced in both hardware and software technology in detail. (authors)

Reports from the working group on neutron scattering

International Nuclear Information System (INIS)

1979-06-01

The present report contains papers dating from July 1978 until May 1979. During this period the experimental facilities have been expanded; a new four-circuit neutron spectrometer was installed and, together with the Fritz Hafer Institute, a measuring point was set up for investigations of ideal crystals, the Compton scattering equipment has been essentially improved. The report contains a contribution on the mechanics and the control of the neutron diffractometers existing at BER II. The main subjects of the scientific research work were magnetic structures and phase transitions, electron densities and chemical bonds, structure and dynamics of molecular crystals. At the BER II reactor measuring opportunities could be offered to a number of guest groups. Their research activities are reported, too. In addition to those made at the Berlin reactor BER II measurements could be made at the accelerator VICKSI of the Hahn-Meitner Institute and at the reactors of the Institute Laue-Langevin at Grenoble and of the Research Establishment at Riso by the working groups. (orig.) [de

Experience with automatic reactor control at EBR-II

International Nuclear Information System (INIS)

Lehto, W.K.; Larson, H.A.; Christensen, L.J.

1985-01-01

Satisfactory operation of the ACRDS has extended the capabilities of EBR-II to a transient test facility, achieving automatic transient control. Test assemblies can now be irradiated in transient conditions overlapping the slower transient capability of the TREAT reactor

BER estimation for all-optical regenerators influenced by pattern effects

DEFF Research Database (Denmark)

Bischoff, Svend; Lading, B.; Mørk, Jesper

2002-01-01

An efficient method is presented for the estimation of the bit-error rate (BER) of a system employing all-optical regenerators influenced by pattern effects. We theoretically study noise accumulation and noise redistribution in long distance transmission systems employing a delayed interference...... signal wavelength converter for all-optical regeneration. The BER is studied for return-to-zero signals at bit rates of 2.5 Gb/s (no patterning) up to 40 Gb/s (strong patterning). The calculation of the BER is based on pattern dependent transfer functions, which may be obtained numerically or measured....

HYLIFE-II reactor chamber mechanical design: Update

International Nuclear Information System (INIS)

House, P.A.

1992-01-01

Mechanical design features of the reactor chamber for the HYLIFE-II inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams (Li 2 BeF 4 ) are used for shielding and blast protection of the chamber walls. The system is designed for a 6 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (17 m/s) salt streams and also recover up to half of the dynamic head. Cost estimates for a 1 GW e and 2 GW e reactor chamber are presented

Reactor costs and maintenance, with reference to the Culham Mark II conceptual tokamak reactor design

International Nuclear Information System (INIS)

Hancox, R.; Mitchell, J.T.D.

1977-01-01

Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are the capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, mainly because of the low power density of the fusion reactor which affects both the reactor and building costs. To reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (author)

Reactor costs and maintenance, with reference to the Culham Mark II conceptual Tokamak reactor design

International Nuclear Information System (INIS)

Hancox, R.; Mitchell, J.T.D.

1976-01-01

Published designs of tokamak reactors have proposed conceptual solutions for most of the technological problems encountered. Two areas which remain uncertain, however, are capital cost of the reactor and the practicability of reactor maintenance. A cost estimate for the Culham Conceptual Tokamak Reactor (Mk I) is presented. The capital cost of a power station incorporating this reactor would be significantly higher than that of an equivalent fast breeder fission power station, due mainly to the low power density of the fusion reactor which affects both the reactor and building costs. In order to reduce the fusion station capital costs a new conceptual design is proposed (Mk II) which incorporates a shaped plasma cross-section to give a higher plasma pressure ratio, βsub(t) approximately 0.1. Since the higher power density implies more severe radiation damage of the blanket structure, the question of reactor maintenance assumes greater importance. With the proposed scheme for regular replacement of the blanket, a fusion power station availability around 0.9 should be achievable. (orig.) [de

Thermal-structural response of EBR-II major components under reactor operational transients

International Nuclear Information System (INIS)

Chang, L.K.; Lee, M.J.

1983-01-01

Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

Preliminary neutronic design of TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Sarikaya, B.; Tombakoglu, M.; Cecen, Y.; Kadiroglu, O. K.

2001-01-01

It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

Neutronics and thermohydraulics of the reactor C.E.N.E. Part II; Analisis neutronico y termohidraulico del reactor C.E.N.E. Parte II

Energy Technology Data Exchange (ETDEWEB)

Caro, R

1976-07-01

In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs.

Operating experience of TRIGA MK-II Research Reactor in Bangladesh

International Nuclear Information System (INIS)

Mannan, M.A.; Ahmed, K.

1992-01-01

A 3 MW TRIGA MK II Research Reactor was installed in Bangladesh in 1986. The reactor is being utilized for research, training and for production of radioisotopes. Recently two faults were detected, one in the Emergency Core Cooling System and the other in the Primary Coolant Loop, which hindered the operation of the reactor partially. The faults were investigated by a team of local experts. Results of analyses of possible initiating events of the faults and the remedial steps are briefly discussed in the paper. (author)

Eight joint BER II and BESSY II users meeting. Abstracts

International Nuclear Information System (INIS)

2016-01-01

The following topics were dealt with: Accelerator operation and projecs, photon science and instrumentation at BESSY II, status of energy materials in-situ Lab at BESSY II, high resolution spectrometer PEAXIS at BESSY II, sample environment at BESSY II, molecular control mechanisms in the Brr2 RNA helicase for efficient and regulated splicing, the Li conversion reaction of 4CoFe_2O_4 nanoparticles, buried interfaces in lithium ion batteries probed with HAXPES, ARPES studies of the STO(001) 2DEG, all-in/all-out magnetic order in rare earth iridates, oxygen reduction reaction on graphene in Li-air batteries, electronic order in high-T_c superconductors, in-siu observation of novel switching phenomena in highly porous metal-organic frameworks, photoinduced demagnetization and insulator-to-metal transition in ferromagnetic insulating BaFeO_3 thin films, ARPES measurement of the ferroelectric bulk Rashba system GeTe, bisphenol A on Cu(111) and Ag(111), reverse water-gas shift or Sabathier methanation on N(110), structural studies of molecular machines, multi-MHz time-of-flight electronic band-structure imaging of graphene on Ir(111), diffusion pathways in ion conductors, ground-state potential energy surfaces around selected atoms from resonant inelastic X-ray scattering, solar energy in an emerging country, in-situ neutron analysis of electrode materials for electrochemical energy storage, structure and transport properties in thermoelectric skutterudites, investigation of the interphase formation on solid lithium-ion conductors by neutron reflectometry, load partitin and damage characterization of cast AlSi_1_2CuMgNi alloy with ceramic reinforcement, methane adsorption in highly porous metal-organics, structure and magnetic interactions in dimer system Ba_(_3_-_x_)Sr_xCr_2O_8, distribution of S in C-S nanocomposites, current status of HFM-EXED FACITIY; SPIN NEAMTICITY IN s=1/2 frustrated zigzag chaIN β-TeVO_4, electronic properties of U(Ru_0_._9_2Rh_0_._0_8)_2Si_2 in

The neutron utilization and promotion program of TRR-II research reactor project in Taiwan

International Nuclear Information System (INIS)

Gone, J.K.; Huang, Y.H.

2001-01-01

The objective of the Taiwan research reactor system improvement and utilization promotion project is to reconstruct the old Taiwan research reactor (TRR), which was operated by the Institute of Nuclear Energy Research (INER) between 1973 and 1988, into a multi-purpose medium flux research reactor (TRR-II). The project started in 1998, and the new reactor is scheduled to have its first critical in June of 2006. The estimated maximum unperturbed thermal neutron flux (E 14 n/cm 2 sec, and it is about one order of magnitude higher than other operating research reactors in Taiwan. The new reactor will equip with secondary neutron sources to provide neutrons with different energies, which will be an essential tool for advanced material researches in Taiwan. One of the major tasks of TRR-II project is to promote domestic utilization of neutrons generated at TRR-II. The traditional uses of neutrons in fuel/material research, trace element analysis, and isotope production has been carried out at INER for many years. On the other hand, it is obvious that promotions of neutron spectrometric technique will be a major challenge for the project team. The limited neutron flux from operating research reactors had discouraged domestic users in developing neutron spectrometric technique for many years, and only few researchers in Taiwan are experienced in using spectrometers. It is important for the project team to encourage domestic researchers to use neutron spectrometers provided by TRR-II as a tool for their future researches in various fields. This paper describes the current status of TRR-II neutron utilization and promotion program. The current status and future plans for important issues such as staff recruiting, personnel training, international collaboration, and promotion strategy will be described. (orig.)

Lethal and mutagenic properties of MMS-generated DNA lesions in Escherichia coli cells deficient in BER and AlkB-directed DNA repair.

Science.gov (United States)

Sikora, Anna; Mielecki, Damian; Chojnacka, Aleksandra; Nieminuszczy, Jadwiga; Wrzesinski, Michal; Grzesiuk, Elzbieta

2010-03-01

Methylmethane sulphonate (MMS), an S(N)2-type alkylating agent, generates DNA methylated bases exhibiting cytotoxic and mutagenic properties. Such damaged bases can be removed by a system of base excision repair (BER) and by oxidative DNA demethylation catalysed by AlkB protein. Here, we have shown that the lack of the BER system and functional AlkB dioxygenase results in (i) increased sensitivity to MMS, (ii) elevated level of spontaneous and MMS-induced mutations (measured by argE3 --> Arg(+) reversion) and (iii) induction of the SOS response shown by visualization of filamentous growth of bacteria. In the xth nth nfo strain additionally mutated in alkB gene, all these effects were extreme and led to 'error catastrophe', resulting from the presence of unrepaired apurinic/apyrimidinic (AP) sites and 1-methyladenine (1meA)/3-methylcytosine (3meC) lesions caused by deficiency in, respectively, BER and AlkB dioxygenase. The decreased level of MMS-induced Arg(+) revertants in the strains deficient in polymerase V (PolV) (bearing the deletion of the umuDC operon), and the increased frequency of these revertants in bacteria overproducing PolV (harbouring the pRW134 plasmid) indicate the involvement of PolV in the error-prone repair of 1meA/3meC and AP sites. Comparison of the sensitivity to MMS and the induction of Arg(+) revertants in the double nfo alkB and xth alkB, and the quadruple xth nth nfo alkB mutants showed that the more AP sites there are in DNA, the stronger the effect of the lack of AlkB protein. Since the sum of MMS-induced Arg(+) revertants in xth, nfo and nth xth nfo and alkB mutants is smaller than the frequency of these revertants in the BER(-) alkB(-) strain, we consider two possibilities: (i) the presence of AP sites in DNA results in relaxation of its structure that facilitates methylation and (ii) additional AP sites are formed in the BER(-) alkB(-) mutants.

Multifrequency tests in the EBR-II reactor plant

International Nuclear Information System (INIS)

Feldman, E.E.; Mohr, D.; Gross, K.C.

1989-01-01

A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

Eight joint BER II and BESSY II users meeting. Abstracts

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-07-01

The following topics were dealt with: Accelerator operation and projecs, photon science and instrumentation at BESSY II, status of energy materials in-situ Lab at BESSY II, high resolution spectrometer PEAXIS at BESSY II, sample environment at BESSY II, molecular control mechanisms in the Brr2 RNA helicase for efficient and regulated splicing, the Li conversion reaction of 4CoFe{sub 2}O{sub 4} nanoparticles, buried interfaces in lithium ion batteries probed with HAXPES, ARPES studies of the STO(001) 2DEG, all-in/all-out magnetic order in rare earth iridates, oxygen reduction reaction on graphene in Li-air batteries, electronic order in high-T{sub c} superconductors, in-siu observation of novel switching phenomena in highly porous metal-organic frameworks, photoinduced demagnetization and insulator-to-metal transition in ferromagnetic insulating BaFeO{sub 3} thin films, ARPES measurement of the ferroelectric bulk Rashba system GeTe, bisphenol A on Cu(111) and Ag(111), reverse water-gas shift or Sabathier methanation on N(110), structural studies of molecular machines, multi-MHz time-of-flight electronic band-structure imaging of graphene on Ir(111), diffusion pathways in ion conductors, ground-state potential energy surfaces around selected atoms from resonant inelastic X-ray scattering, solar energy in an emerging country, in-situ neutron analysis of electrode materials for electrochemical energy storage, structure and transport properties in thermoelectric skutterudites, investigation of the interphase formation on solid lithium-ion conductors by neutron reflectometry, load partitin and damage characterization of cast AlSi{sub 12}CuMgNi alloy with ceramic reinforcement, methane adsorption in highly porous metal-organics, structure and magnetic interactions in dimer system Ba{sub (3-x)}Sr{sub x}Cr{sub 2}O{sub 8}, distribution of S in C-S nanocomposites, current status of HFM-EXED FACITIY; SPIN NEAMTICITY IN s=1/2 frustrated zigzag chaIN β-TeVO{sub 4}, electronic

Bemerkungen über Pedipalpi (Fortsetzung)

NARCIS (Netherlands)

Speijer, E.A.M.

1931-01-01

In meiner vorigen Mitteilung (Bemerkungen über Pedipalpi, Zoologische Mededeelingen 's Rijks Museum van Natuurlijke Historie te Leiden Deel XIV, Afl. 1—2, 1931, p. 79—88) habe ich schon darauf hingewiesen, dass man bei der Bearbeitung der Thelyphonidae auf eigentümliche Schwierigkeiten stösst. Diese

Data base formation for important components of reactor TRIGA MARK II

International Nuclear Information System (INIS)

Jordan, R.; Mavko, B.; Kozuh, M.

1992-01-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [sl

Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

International Nuclear Information System (INIS)

Rafhayudi Jamro; Redzuwan Yahaya; Abdul Aziz Mohamed; Eid Abdel-Munem; Megat Harun Al-Rashid; Julia Abdul Karim; Ikki Kurniawan; Hafizal Yazid; Azraf Azman; Shukri Mohd

2008-01-01

A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (k e ff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of k e ff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of k e ff from MCNP5 were obtained when the control fuel rods were fully inserted, the k e ff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of k e ff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

Comparative analyses of ber (zizyphus) honey with other types of honey

International Nuclear Information System (INIS)

Iftikhar, F.; Mehmood, R.

2007-01-01

Eight different honey samples were collected from different ecological areas of Pakistan and analyzed for various physico-chemical parameters. The aim was to evaluate the quality, difference in the composition and comparison of Ber/Sidder (Zizyphus spp.) honey with other types of honey. The two samples of Ber honey and six other floral honey i.e., Citrus, mixed (citrus + Justicia), Acacia modesta, multi-flora and mixed (zizyphus + sesamum) were collected from Apis mellifera colonies. The colonies of Apis mellifera foraged on the above-mentioned flora in different ecological areas i.e. ber honey from Bannu and Karak, Citrus and mixed (citrus + Justicia) honey from Mandi Bahauddin, Acacia honey from Peshawar + Islamabad and multi-flora from Lahore. Nine legal parameters of honey quality control were followed. All honey samples were well within the limits of Codex Alimentarius Commission except ber and citrus. The ber honey showed high pH (6.85) and low reducing sugars (67.66%), sucrose (1.0%), free acidity (6.09 meq/kg) and total acidity (7.68 meq/kg). Mixed honey (zizyphus + sesamum) exhibited highest pH (6.88), lowest free acidity (2.80 meq/kg) and total acidity (5.50 meq/kg). Citrus honey was also found high in Hydroxy-methyl-furfural (HMF, 22.75 meql kg). (author)

Reactor physics studies in the GCFR phase-II critical assembly

International Nuclear Information System (INIS)

Pond, R.B.

1976-09-01

The reactor physics studies performed in the gas cooled fast reactor (GCFR) mockup on ZPR-9 are covered. This critical assembly, designated Phase II in the GCFR program, had a single zone PuO 2 -UO 2 core composition and UO 2 radial and axial blankets. The assembly was built both with and without radial and axial stainless steel reflectors. The program included the following measurements: small-sample reactivity worths of reactor constituent materials (including helium); 238 U Doppler effect; uranium and plutonium reaction rate distributions; thorium, uranium, and plutonium α and reactor kinetics. Analysis of the measurements used ENDF/B-IV nuclear data; anisotropic diffusion coefficients were used to account for neutron streaming effects. Comparison of measurements and calculations to GCFR Phase I are also made

Ageing Management in the CENM Triga Mark II Research Reactor

Energy Technology Data Exchange (ETDEWEB)

El Younoussi, C.; Nacir, B.; El Bakkari, B.; Boulaich, Y. [Centre for Nuclear Studies of Maâmora (CENM), National Centre of Energy Sciences and Nuclear Techniques (CNESTEN), Rabat (Morocco)

2014-08-15

Physical ageing is one of the most important factors that may reduce the safety margins calculated in the design of safety system components of a research reactor. In this context, special efforts are necessary for ensuring the safety of research reactors through appropriate ageing management actions. The paper deals with the overall aspects of the ageing management system of the Moroccan TRIGA Mark II research reactor. The management system covers among others, management of structures, critical components inspections, the control command system and nuclear instrumentation verification. The paper presents also how maintenance and periodic testing are organized and managed in the reactor module. Practical examples of ageing management actions of some systems and components during recent years are presented. (author)

TIBER II: an upgraded tokamak igntion/burn experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Perkins, L.J.

1986-01-01

We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

Final project report: TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

International Nuclear Information System (INIS)

Montoya, G.M.

1993-02-01

This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project including, Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 M 3 of mixed waste

Experience with EBR-II [Experimental Breeder Reactor] driver fuel

International Nuclear Information System (INIS)

Seidel, B.R.; Porter, D.L.; Walters, L.C.; Hofman, G.L.

1986-01-01

The exceptional performance of Experimental Breeder Reactor-II (EBR-II) metallic driver fuel has been demonstrated by the irradiation of a large number of elements under steady-state, transient overpower, and loss-of-flow conditions. High burnup with high reliability has been achieved by a close coupling of element design and materials selection. Quantification of reliability has allowed full utilization of element lifetime. Improved design and duct materials currently under test are expected to increase the burnup from 8 to 14 at.%

Evaluation of Fe(II) oxidation at an acid mine drainage site using laboratory-scale reactors

Science.gov (United States)

Brown, Juliana; Burgos, William

2010-05-01

Acid mine drainage (AMD) is a severe environmental threat to the Appalachian region of the Eastern United States. The Susquehanna and Potomac River basins of Pennsylvania drain to the Chesapeake Bay, which is heavily polluted by acidity and metals from AMD. This study attempted to unravel the complex relationships between AMD geochemistry, microbial communities, hydrodynamic conditions, and the mineral precipitates for low-pH Fe mounds formed downstream of deep mine discharges, such as Lower Red Eyes in Somerset County, PA, USA. This site is contaminated with high concentrations of Fe (550 mg/L), Mn (115 mg/L), and other trace metals. At the site 95% of dissolved Fe(II) and 56% of total dissolved Fe is removed without treatment, across the mound, but there is no change in the concentration of trace metals. Fe(III) oxides were collected across the Red Eyes Fe mound and precipitates were analyzed by X-ray diffraction, electron microscopy and elemental analysis. Schwertmannite was the dominant mineral phase with traces of goethite. The precipitates also contained minor amounts of Al2O3, MgO,and P2O5. Laboratory flow-through reactors were constructed to quantify Fe(II) oxidation and Fe removal over time at terrace and pool depositional facies. Conditions such as residence time, number of reactors in sequence and water column height were varied to determine optimal conditions for Fe removal. Reactors with sediments collected from an upstream terrace oxidized more than 50% of dissolved Fe(II) at a ten hour residence time, while upstream pool sediments only oxidized 40% of dissolved Fe(II). Downstream terrace and pool sediments were only capable of oxidizing 25% and 20% of Fe(II), respectively. Fe(II) oxidation rates measured in the reactors were determined to be between 3.99 x 10-8and 1.94 x 10-7mol L-1s-1. The sediments were not as efficient for total dissolved Fe removal and only 25% was removed under optimal conditions. The removal efficiency for all sediments

Effect of gamma irradiation on antioxidant properties of ber (Zizyphus mauritiana) fruit

International Nuclear Information System (INIS)

Kavitha, C.; Kuna, Aparna; Sagar, S.B.; Padmavati, T.V.N.; Supraja, T.; Prabhakar, N.

2015-01-01

Effect of gamma irradiation (0.25 to 1.0kGy) on antioxidant properties of ber fruit was studied. Antioxidant properties of ber fruits were determined by scavenging DPPH radical activity, reducing power assay, super oxide anion radical activity, TBARS, total phenolic content and total flavonoid content. Gamma irradiation treatment up to 1.0 kGy elevated the scavenging DPPH radical activity (9 %), super oxide anion radical activity (26 %) and total flavonoid content (208 %) compared to fresh ber fruit. On the other hand it brought down the reducing power activity (65 %) and total phenolic content (18 %) as compared to raw fruit. The TBARS activity statistically increased upon irradiation of ber fruit. It indicated that total antioxidant activity decreased as TBARS value increased. Therefore 0.25 to 0.5kGy is better dose to retain the natural antioxidant in fruit. (author)

Safety and environmental aspects of the HYLIFE-II and ARIES fusion reactor designs

International Nuclear Information System (INIS)

Dolan, T.J.; Longhurst, G.R.; Herring, J.S.

1993-01-01

The HYLIFE-II inertial confinement fusion reactor design uses jets of Flibe molten salt to protect the blast chamber walls and to breed tritium. It has a low tritium inventory and effective tritium removal. The issue with this design is not one of safety but of economics. The ARIES reactor designs have safety concerns associated with fires. These reactors designs are described

Scientific report 3 - 1975

International Nuclear Information System (INIS)

1976-01-01

The reactor BER II went into routine operation in 1975. Despite some difficulties and failures it was possible to carry out the research programme in this field mainly with this reactor. The research projects were: Neutron scattering, radiation damage in solids, reactor chemistry, trace elements, preparative nuclear chemistry. (orig./HK) [de

BER Science Network Requirements

Energy Technology Data Exchange (ETDEWEB)

Alapaty, Kiran; Allen, Ben; Bell, Greg; Benton, David; Brettin, Tom; Canon, Shane; Dart, Eli; Cotter, Steve; Crivelli, Silvia; Carlson, Rich; Dattoria, Vince; Desai, Narayan; Egan, Richard; Tierney, Brian; Goodwin, Ken; Gregurick, Susan; Hicks, Susan; Johnston, Bill; de Jong, Bert; Kleese van Dam, Kerstin; Livny, Miron; Markowitz, Victor; McGraw, Jim; McCord, Raymond; Oehmen, Chris; Regimbal, Kevin; Shipman, Galen; Strand, Gary; Flick, Jeff; Turnbull, Susan; Williams, Dean; Zurawski, Jason

2010-11-01

The Energy Sciences Network (ESnet) is the primary provider of network connectivity for the US Department of Energy Office of Science, the single largest supporter of basic research in the physical sciences in the United States. In support of the Office of Science programs, ESnet regularly updates and refreshes its understanding of the networking requirements of the instruments, facilities, scientists, and science programs that it serves. This focus has helped ESnet to be a highly successful enabler of scientific discovery for over 20 years. In April 2010 ESnet and the Office of Biological and Environmental Research, of the DOE Office of Science, organized a workshop to characterize the networking requirements of the science programs funded by BER. The requirements identified at the workshop are summarized and described in more detail in the case studies and the Findings section. A number of common themes emerged from the case studies and workshop discussions. One is that BER science, like many other disciplines, is becoming more and more distributed and collaborative in nature. Another common theme is that data set sizes are exploding. Climate Science in particular is on the verge of needing to manage exabytes of data, and Genomics is on the verge of a huge paradigm shift in the number of sites with sequencers and the amount of sequencer data being generated.

Do berço ao berçário: a instituição como morada e lugar de contato

Directory of Open Access Journals (Sweden)

Marta Nörnberg

2013-12-01

Full Text Available Este ensaio aborda o sentido da instituição infantil a partir de uma perspectiva ético-estético-afetiva do viver humano. A abordagem ampara-se em perspectivas filosóficas sobre a existência humana. Os argumentos apresentam o berçário como morada, lugar do viver juntos, instituição que se faz em coletividade, nela e com ela, decorrente do encontro de bebês e adultos. Ao constituir-se como morada dos bebês, o berçário ritualiza e reatualiza a prática pedagógica como lugar de relação entre bebês e adultos, requerendo uma pedagogia do contato.

Minimum BER Receiver Filters with Block Memory for Uplink DS-CDMA Systems

Directory of Open Access Journals (Sweden)

Debbah Mérouane

2008-01-01

Full Text Available Abstract The problem of synchronous multiuser receiver design in the case of direct-sequence single-antenna code division multiple access (DS-CDMA uplink networks is studied over frequency selective fading channels. An exact expression for the bit error rate (BER is derived in the case of BPSK signaling. Moreover, an algorithm is proposed for finding the finite impulse response (FIR receiver filters with block memory such that the exact BER of the active users is minimized. Several properties of the minimum BER FIR filters with block memory are identified. The algorithm performance is found for scenarios with different channel qualities, spreading code lengths, receiver block memory size, near-far effects, and channel mismatch. For the BPSK constellation, the proposed FIR receiver structure with block memory has significant better BER with respect to and near-far resistance than the corresponding minimum mean square error (MMSE filters with block memory.

Performances on nuclear activation analysis by TRIGA Mark II reactor

International Nuclear Information System (INIS)

Capannesi, G.; Rosada, A.

1986-01-01

Progresses in methodological research and connected applications in the field of activation analysis are introduced. Some peculiar characteristics on the TRIGA MARK II reactor have enabled the possibility of obtaining interesting results. The particular, the rotating radiation device Lazy Susan, with a capability of 40 positionings, permits homogeneity in neutron flux and energy spectrum stability within 15%. High level of precision and accuracy are obtained in analytic. Applications of major interest have been: - reference material certification; - forensic applications; - electrolytic cell productivity evaluation. The TRIGA MARK II reactor is equipped with a thermal column throughout a D 2 O diaphragm with a thickness of 70 cm. The available neutron flux has no fast and epithermal components. Via this facility a method has been tested for the instrumental determination of Al in Si metal of solar and electronic degree. (author)

Monte Carlo analysis of Musashi TRIGA mark II reactor core

International Nuclear Information System (INIS)

Matsumoto, Tetsuo

1999-01-01

The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k eff ) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k eff overestimated the experimental data by about 1.0%Δk/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

Neue Werkstoffe über additive Fertigung

OpenAIRE

Günther, Johannes; Niendorf, Thomas

2016-01-01

Über die additive Fertigung, oftmals bezeichnet als 3D-Druck, lassen sich Bauteile nahezu beliebiger geometrischer Komplexität herstellen. Gleichzeitig lassen die Prozessrandbedingungen die direkte Einstellung der Mikrostruktur in den verwendeten metallischen Werkstoffen zu. Hieraus ergeben sich weitreichende Möglichkeiten bezüglich der Eigenschaftsoptimierung aktueller Hochleistungswerkstoffe.

A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

International Nuclear Information System (INIS)

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

1987-01-01

Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

Minimum BER Receiver Filters with Block Memory for Uplink DS-CDMA Systems

Directory of Open Access Journals (Sweden)

Mérouane Debbah

2008-05-01

Full Text Available The problem of synchronous multiuser receiver design in the case of direct-sequence single-antenna code division multiple access (DS-CDMA uplink networks is studied over frequency selective fading channels. An exact expression for the bit error rate (BER is derived in the case of BPSK signaling. Moreover, an algorithm is proposed for finding the finite impulse response (FIR receiver filters with block memory such that the exact BER of the active users is minimized. Several properties of the minimum BER FIR filters with block memory are identified. The algorithm performance is found for scenarios with different channel qualities, spreading code lengths, receiver block memory size, near-far effects, and channel mismatch. For the BPSK constellation, the proposed FIR receiver structure with block memory has significant better BER with respect to Eb/N0 and near-far resistance than the corresponding minimum mean square error (MMSE filters with block memory.

Ein Lehrbuch über Psychiatrie und Psychotherapie – Geschlechtsunterschiede unter der Lupe?

Directory of Open Access Journals (Sweden)

Isabella Heuser

2007-07-01

Full Text Available Anke Rohde und Andreas Marneros haben sich ein hohes Ziel gesetzt: ein übersichtliches Lehrbuch für Psychiatrie und Psychotherapie, das erstmals eine systematische Zusammenstellung des Wissens über geschlechtsspezifische Aspekte von Symptomatologie, Epidemiologie, Diagnostik, Pharmakotherapie und Psychotherapie psychischer Störungen enthält. Bemerkenswert ist dabei zweierlei: Erstens werden bei der Betrachtung geschlechtsspezifischer Unterschiede im Sinne der Genderforschung Frauen und Männer berücksichtigt, zweitens wird ein vollständiger Überblick sowohl über die häufigen und bekannten psychischen Störungen als auch die äußerst seltenen psychiatrischen Erkrankungen der gesamten Altersspanne gegeben. Wenngleich die Qualität der einzelnen Beiträge deutlich zwischen sehr differenzierter und leider auch für ein Handbuch eindeutig zu oberflächlicher Betrachtung variiert, liegt in dem Band ein weitgehend informatives Nachschlagewerk vor, das einen ersten Überblick über die geschlechtsspezifische Psychiatrie gibt.

BER performance comparison of optical CDMA systems with/without turbo codes

Science.gov (United States)

Kulkarni, Muralidhar; Chauhan, Vijender S.; Dutta, Yashpal; Sinha, Ravindra K.

2002-08-01

In this paper, we have analyzed and simulated the BER performance of a turbo coded optical code-division multiple-access (TC-OCDMA) system. A performance comparison has been made between uncoded OCDMA and TC-OCDMA systems employing various OCDMA address codes (optical orthogonal codes (OOCs), Generalized Multiwavelength Prime codes (GMWPC's), and Generalized Multiwavelength Reed Solomon code (GMWRSC's)). The BER performance of TC-OCDMA systems has been analyzed and simulated by varying the code weight of address code employed by the system. From the simulation results, it is observed that lower weight address codes can be employed for TC-OCDMA systems that can have the equivalent BER performance of uncoded systems employing higher weight address codes for a fixed number of active users.

TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

International Nuclear Information System (INIS)

Lee, J.D.; Sawan, M.

1987-01-01

TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1989-01-01

Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

Energy Technology Data Exchange (ETDEWEB)

1984-01-01

This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

International Nuclear Information System (INIS)

1984-01-01

This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

Data base formation for important components of reactor TRIGA MARK II

Energy Technology Data Exchange (ETDEWEB)

Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

1992-07-01

The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

Final project report, TA-35 Los Alamos Power Reactor Experiment No. II (LAPRE II) decommissioning project

International Nuclear Information System (INIS)

Montoya, G.M.

1992-01-01

This final report addresses the decommissioning of the LAPRE II Reactor, safety enclosure, fuel reservoir tanks, emergency fuel recovery system, primary pump pit, secondary loop, associated piping, and the post-remediation activities. Post-remedial action measurements are also included. The cost of the project, including Phase I assessment and Phase II remediation was approximately $496K. The decommissioning operation produced 533 m 3 of low-level solid radioactive waste and 5 m 3 of mixed waste

Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Ustun, G.; Durmayaz, A.

2002-01-01

Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

Control console conceptual design for sheet type fuels of Triga Mark-II reactor

International Nuclear Information System (INIS)

Eko Priyono; Kurnia Wibowo; Anang Susanto

2016-01-01

The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

Alfredo Häberli ja Iittala : Senta / Piret Veigel

Index Scriptorium Estoniae

Veigel, Piret, 1961-

2004-01-01

Jaanuaris Pariisi messil Maison & Objet aasta disaineriks kuulutatud Alfredo Häberli esitles veebruaris Frankfurdi messil Ambiente oma neljandat ühisprojekti Soome Iittalaga: Senta-nimelisi valge, punase ja vahuveini klaase

Sodium fast reactor safety and licensing research plan - Volume II

International Nuclear Information System (INIS)

Ludewig, H.; Powers, D.A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.

2012-01-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

Sodium fast reactor safety and licensing research plan. Volume II.

Energy Technology Data Exchange (ETDEWEB)

Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

2012-05-01

Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

In-reactor cladding breach of EBR-II driver-fuel elements

International Nuclear Information System (INIS)

Seidel, B.R.; Einziger, R.E.

1977-01-01

Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter

Tritium handling, breeding, and containment in two conceptual fusion reactor designs: UWMAK-II and UWMAK-III

International Nuclear Information System (INIS)

Clemmer, R.G.; Larsen, E.M.

1976-01-01

Tritium is an essential component of near-term controlled thermonuclear reactor systems. Since tritium is not likely to be available on a large scale at a modest cost, fusion reactor designs must incorporate blanket systems which will be capable of breeding tritium. Because of the radiological activity and capability of assimilation into living tissues, tritium release to the environment must be strictly controlled. The University of Wisconsin has completed three conceptual designs of fusion reactors, UWMAK-I, UWMAK-II, and UWMAK-III. This report discusses the tritium systems for UWMAK-II, a reactor design with a helium cooled solid breeder blanket, and UWMAK-III, a reactor design with a high-temperature liquid breeder blanket. Tritium systems for fueling and recycling, breeding and recovery, and plant containment and control are discussed. (Auth.)

Neutronics and thermohydraulics of the reactor C.E.N.E. Part II

International Nuclear Information System (INIS)

Caro, R.

1976-01-01

In this report the analysis of neutronics thermohydraulics and shielding of the 10 HWt swimming pool reactor C.E.N.E is included. In each of these chapters is given a short description of the theoretical model used, along with the theoretical versus experimental checking carried out, whenever possible, with the reactors JEN-I and JEN-II of Junta de Energia Nuclear. (Author) 11 refs

Fuel burnup analysis of the TRIGA Mark II reactor at the University of Pavia

International Nuclear Information System (INIS)

Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

2016-01-01

Highlights: • A fuel evolution model for a TRIGA Mark II reactor has been developed. • Reproduction of nearly 50 years of reactor operation. • The model was used to predict the best reactor reconfiguration. • Reactor life was extended without adding fresh fuel elements. - Abstract: A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyze neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate low power experimental reactors from those used for power production, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the core and to obtain a substantial increase in the Core Excess reactivity value. The evaluation of fuel burnup and the reconfiguration results are presented in this paper.

Tritium system design for the mirror reactors FPD-I, FPD-II, and FPD-III

International Nuclear Information System (INIS)

Finn, P.A.

1985-01-01

The tritium system design for the Fusion Power Demonstration Reactor (FPD-I, II, and III) is described. The device operates at 25% availability. For FPD-II, an engineering mode using tritium neutral beams is part of the design

Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

International Nuclear Information System (INIS)

Ergun, E; Kadiroglu, O.S.

1999-01-01

The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

Directions for attractive tokamak reactors: The ARIES-II and ARIES-IV second-stability designs

International Nuclear Information System (INIS)

Najmabadi, F.; Conn, R.W.

1993-01-01

ARIES is a research program to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The ARIES study has developed four visions for tokamaks. All four designs are steady-state, 1000-MWe (net) power reactors. The ARIES-II and ARIES-IV designs assume potential advances in plasma physics (such as second-stability operation) predicted by theory but not yet established experimentally. The two designs have the same fusion plasma but different fusion-power-core. There are only minor differences between the ARIES-II and ARIES-IV plasma parameters. ARIES-IV is a 1000-MWe reactor with an average neutron wall loading of 3 MW/m 2 , and a mass power density of about 120 kWe/tonne of fusion power core. The reactor major radius is 6.1 m, the plasma minor radius is 1.5 m and the plasma elongation is 2, and the plasma triangularity is 0.67. The plasma current is low (6.8 MA), B on-axis is 7.7 T (corresponding to a maximum field at the coil of 16T), and the toroidal beta is 3.4% (Troyon coefficient = 6). The operating regime is optimized such that most of the plasma current (∼ 90%) is provided by the bootstrap current. ARIES-II uses liquid lithium as the coolant and tritium breeder. V-5Cr-5Ti is used as the structural material so that the potential of low-activation metallic blankets can be studied. ARIES-IV uses helium as the coolant, a solid tritium-breeding material (Li 2 O), and silicon carbide composite as structural material. The waste produced by neutron activation in both designs is found to meet the criteria allowing shallow-land burial under U.S. regulations. The cost of electricity for the ARIES-II-IV class of reactors is estimated to be about 20% lower than comparable, steady-state first-stability reactors (e.g. ARIES-I). 25 refs, 2 figs, 1 tab

Considerations for advanced reactor design based on EBR-II experience

International Nuclear Information System (INIS)

King, R. W.

1999-01-01

The long-term success of the Experimental Breeder Reactor-II (EBR-II) provides several insights into fundamental characteristics and design features of a nuclear generating station that enhance safety, operability, and maintainability. Some of these same characteristics, together with other features, offer the potential for operational lifetimes well beyond the current licensing time frame, and improved reliability that could potentially reduce amortized capital costs as well as overall operation and maintenance costs if incorporated into advanced plant designs. These features and characteristics are described and the associated benefits are discussed

Bulgarian Emergency Response System (BERS) in case of nuclear accident with exposure doses’estimation

Energy Technology Data Exchange (ETDEWEB)

Syrakov, M.; Prodanova, M.; Slavov, K.; Veleva, B.

2015-07-01

A PC-oriented Emergency Response System in case of nuclear accident (BERS) is developed and works operationally in the National Institute of Meteorology and Hydrology (NIMH). The creation and development of BERS was highly stimulated by the ETEX (European Tracer Experiment) project. BERS comprises two main parts - the operational and the accidental ones. The operational part, run automatically every 12 hours, prepares the input meteorological file used by both trajectory and dispersion models, runs the trajectory models, visualizes the results and uploads the maps of trajectories to a dedicated web-site. The accidental part is activated manually when a real radioactive releases occur or during emergency exercises. Its core is the Bulgarian dispersion models EMAP. Outputs are concentration, accumulated deposition and selected doses fields. In the paper, the BERS overall structure is described and examples of its products are presented. (Author)

Perturbation analysis of the TRIGA Mark II reactor Vienna

Energy Technology Data Exchange (ETDEWEB)

Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan); Villa, M.; Stummer, T.; Boeck, H. [Vienna Univ. of Technology (Austria). Atominstitut; Saeedbadshah [International Islamic Univ., Islamabad (Pakistan)

2013-04-15

The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Gilleland, J.R.

1988-01-01

The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

Operating experience and maintenance at the TRIGA Mark II LENA reactor

International Nuclear Information System (INIS)

Cingoli, F.; Altieri, S.; Lana, F.; Rosti, G.; Alloni, L.; Meloni, S.

1988-01-01

The last two years at the Trigs Mark II LENA plant were characterized by the running of the n-n-bar oscillation NADIR experiment. Consequently reactor operation was positively affected and the running hours rose again above 1000 hours per year. The LENA team was also deeply involved in the procedures for the renewal of the reactor operation license. The new requirements set by the Nuclear Energy Licensing Authority (ENEA for Italy) most of which concerning radiation protection and environmental impact, have been already fulfilled. In some cases the installation of new apparatus is underway

Purification, crystallization and initial crystallographic characterization of brazil-nut allergen Ber e 2

Energy Technology Data Exchange (ETDEWEB)

Guo, Feng; Jin, Tengchuan; Howard, Andrew; Zhang, Yu-Zhu, E-mail: zhangy@iit.edu [Department of Biology, Illinois Institute of Technology, Chicago, IL 60616 (United States)

2007-11-01

The crystallization of the brazil nut allergen Ber e 2 is reported. Peanut and tree-nut allergies have attracted considerable attention because of their frequency and their lifelong persistence. Brazil-nut (Bertholletia excelsa) allergies have been well documented and the 11S legumin-like seed storage protein Ber e 2 (excelsin) is one of the two known brazil-nut allergens. In this study, Ber e 2 was extracted from brazil-nut kernels and purified to high purity by crystalline precipitation and gel-filtration chromatography. Well diffracting single crystals were obtained using the hanging-drop vapour-diffusion method. A molecular-replacement structural solution has been obtained. Refinement of the structure is currently under way.

Damage analysis of TRIGA MARK II Bandung reactor tank material structure

International Nuclear Information System (INIS)

Soedardjo; Sumijanto

2000-01-01

Damage of Triga Mark II Bandung reactor tank material structure has been analyzed. The analysis carried out was based on ultrasonic inspection result in 1996 and the monthly reports of reactor operation by random data during 1988 up to 1995. Ultrasonic test data had shown that thinning processes on south and west region of reactor out side wall at upper part of water level had happened. Reactor operation data had shown the demineralized water should be added monthly to the reactor and bulk shielding water tank. Both reactor and bulk shielding tank are shielded by concrete of Portland type I cement consisting of CaO content about 58-68 %. The analysis result shows that the reaction between CaO and seepage water from bulk shielding wall had taken place and consequently the reactor out sidewall surroundings became alkaline. Based on Pourbaix diagram, the aluminum reactor tank made of aluminum alloy 6061 T6 would be corroded easily at pH equal an greater than 8.6. The passive layer AI 2 O 3 aluminum metal surface would be broken due to water reaction taken place continuously at high pH and produces hydrogen gas. The light hydrogen gas would expand the concrete cement and its expanding power would open the passive layer of aluminum metal upper tank. The water sea pages from adding water into reactor tank could indicate the upper water level tank corrosion is worse than the lower water level tank. (author)

Study of a new automatic reactor power control for the TRIGA Mark II reactor at University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Borio Di Tigliole, A.; Magrotti, G. [Laboratorio Energia Nucleare Applicata (L.E.N.A.), University of Pavia, Via Aselli 41, 27100 (Italy); Cammi, A.; Memoli, V. [Politecnico di Milano, Department of Energy, Nuclear Engineering Division (CeSNEF), Via Ponzio 34/3, 20133 Milano (Italy); Gadan, M. A. [Instrumentation and Control Department, National Atomic Energy Comission of Argentina, University of Pavia (Italy)

2009-07-01

The installation of a new Instrumentation and Control (IC) system for the TRIGA Mark-II reactor at University of Pavia has recently been completed in order to assure a safe and continuous reactor operation for the future. The intervention involved nearly the whole IC system and required a channel-by-channel component substitution. One of the most sensitive part of the intervention concerned the Automatic Reactor Power Controller (ARPC) which permits to keep the reactor at an operator-selected power level acting on the control rod devoted to the fine regulation of system reactivity. This controller installed can be set up using different control logics: currently the system is working in relay mode. The main goal of the work presented in this paper is to set up a Proportional-Integral-Derivative (PID) configuration of the new controller installed on the TRIGA reactor of Pavia so as to optimize the response to system perturbations. The analysis have shown that a continuous PID offers generally better results than the relay mode which causes power oscillations with an amplitude of 3% of the nominal power

Bulgarian Emergency Response System (BERS) in case of nuclear accident with exposure doses estimation

Energy Technology Data Exchange (ETDEWEB)

Syrakov, D.; Prodanova, M.; Slavov, K.; Veleva, B.

2015-07-01

A PC-oriented Emergency Response System in case of nuclear accident (BERS) is developed and works operationally in the National Institute of Meteorology and Hydrology (NIMH). The creation and development of BERS was highly stimulated by the ETEX (European Tracer Experiment) project. BERS comprises two main parts - the operational and the accidental ones. The operational part, run automatically every 12 hours, prepares the input meteorological file used by both trajectory and dispersion models, runs the trajectory models, visualizes the results and uploads the maps of trajectories to a dedicated web-site. The accidental part is activated manually when a real radioactive releases occur or during emergency exercises. Its core is the Bulgarian dispersion models EMAP. Outputs are concentration, accumulated deposition and selected doses fields. In the paper, the BERS overall structure is described and examples of its products are presented. Key words: nuclear accident, emergency response, early warning system, air dispersion models, radioactive exposure dose. (Author)

Reliability analysis for Atucha II reactor protection system signals

International Nuclear Information System (INIS)

Roca, Jose Luis

1996-01-01

Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

Reliability analysis for Atucha II reactor protection system signals

International Nuclear Information System (INIS)

Roca, Jose L.

2000-01-01

Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

Analysis Of Impact Of Various Parameters On BER Performance For IEEE 802.11b

Directory of Open Access Journals (Sweden)

Nilesh B. Kalani

2015-08-01

Full Text Available Abstract This paper discusses about IEEE 802.11b simulation model implemented using LabVIEW software and its analyses for impact on bit error rate BER for different parameters as channel type channel number data transmission rate and packet size. Audio file is being transmitted processed and analyzed using the model for various parameters. This paper gives analysis of BER verses ESN0 for various parameter like data rate packet size and communication channel for the IEEE 802.11b simulation model generated using LabVIEW. It is proved that BER can be optimized by tweaking different parameters of wireless communication system.

Modernization design of neutron radiography of ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Tugrul, B.; Bilge, A.N.

1988-01-01

ITU TRIGA MARK-II Research and Training Reactor has a power of 250 KW and has three beam tubes. One of them is tangential beam tube used for neutron radiography. In this study, the neutron radiography set in the tangential beam tube is described with its problems for ITU TRIGA Reactor. After that modernization of the system is designed and the applicability of the direct and indirect methods is evaluated. Improving the ratio of length to diameter for the beam tube, elimination the fogging on the film and constructive design for practice and secure application of the technique is developed. (author)

An analysis of water reactor burnup data with the METHUSELAH II code

International Nuclear Information System (INIS)

Floyd, M.; Hicks, D.

1964-10-01

The METHUSELAH II code has been used to predict long term reactivity and isotopic changes in the YANKEE, Dresden and NRX reactors. In general it is shown that there is a satisfactory measure of agreement and the first core lives of YANKEE and Dresden appear well predicted. However there are discrepancies in the isotopic composition of the plutonium formed which appear to be correlated with the degree of hardness of the reactor spectrum. It is demonstrated that plausible changes in nuclear data could reduce the discrepancies. (author)

The control equipment of the Melusine II reactor

International Nuclear Information System (INIS)

Cordelle, M.; Delcroix, V.; Denis, P.; Gariod, R.

1963-01-01

Melusine II, low-power reactor, used for the study of Siloe core has diverged at the CEA Grenoble, the 23. May 1962; its monitoring board studied and carried out in this center is the first in France to be entirely transistorized. The first months of running have justified the hope put in the new electronics to improve the stability and the safety of running. The article describes the design of the control and gives the main characteristics of the measurement chains and of the actions on reactivity. (O.M.) [fr

An analytical approach to the positive reactivity void coefficient of TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Edgue, Erdinc; Yarman, Tolga

1988-01-01

Previous calculations of reactivity void coefficient of I.T.U. TRIGA Mark-II Reactor was done by the second author et al. The theoretical predictions were afterwards, checked in this reactor experimentally. In this work an analytical approach is developed to evaluate rather quickly the reactivity void coefficient of I.T.U. TRIGA Mark-II, versus the size of the void inserted into the reactor. It is thus assumed that the reactor is a cylindrical, bare nuclear system. Next a belt of water of 2πrΔrH is introduced axially at a distance r from the center line of the system. r here, is the thickness of the belt, and H is the height of the reactor. The void is described by decreasing the water density in the belt region. A two group diffusion theory is adopted to determine the criticality of our configuration. The space dependency of the group fluxes are, thereby, assumed to be J 0 (2.405 r / R) cos (π Z / H), the same as that associated with the original bare reactor uniformly loaded prior to the change. A perturbation type of approach, thence, furnishes the effect of introducing a void in the belt region. The reactivity void coefficient can, rather surprisingly, be indeed positive. To our knowledge, this fact had not been established, by the supplier. The agreement of our predictions with the experimental results is good. (author)

Nachrichtenberichterstattung über Terrorismus. Eine Analyse der TV-Nachrichten über die Terroranschläge in Kenia 2002

Directory of Open Access Journals (Sweden)

Nicole Haußecker

2007-04-01

Full Text Available Nach der starken Kritik der Medienberichterstattung über Terrorismus, besonders nach dem 11.09.2001, werden die theoretischen Hintergründe ausgewählter Kritikpunkte betrachtet und ein weiteres terroristisches Ereignis, die Anschläge in Kenia am 28.11.2002, inhaltsanalytisch untersucht. Ziel dabei ist es, Aussagen bezüglich inhaltlicher und formaler Merkmale der Fernseh-berichterstattung über ein terroristisches Ereignis zu treffen. Deshalb liegt der Fokus auf den folgenden drei Forschungsfragen: 1. Welche Nachrichtenfaktoren spielen für die Selektion und Intensität der Berichterstattung über das terroristische Ereignis eine Rolle? 2. Wird mit Emotionalisierung gearbeitet? 3. Liegen negative Stereotype und/oder Feindbilder bezüglich der islamischen und arabischen Welt vor? Die Ergebnisse der Untersuchung zeigen erwartete Tendenzen sowohl die Merkmale der Terrorismusberichterstattung als auch die Senderkonvergenzen und -divergenzen im dualen System betreffend. Der Nachrichtenwert eines terroristischen Ereignisses ist anfänglich sehr hoch, jedoch verliert das Ereignis ab dem dritten Tag aufgrund diverser Nachrichtenfaktoren deutlich an Beachtung in der Berichterstattung. Das Vorkommen emotionalisierender Mittel wird empirisch bestätigt. Nicht nur emotionale Sprache und Sprechweise, sondern vor allem Formen der expliziten Emotionalisierung sind vertreten. Damit wird die Vermutung bekräftigt, dass die Medien die mit den terroristischen Ereignissen verbundene beängstigende Stimmung senderabhängig unterschiedlich stark aufgreifen. Ein direkter islamischer Feindbildaufbau ist in der Berichterstattung nicht zu verzeichnen. Allerdings liegen latente negative Bewertungstendenzen sowie negative Stereotype bezüglich der arabischen und islamischen Welt vor, die das seit dem 11.09.2001 negativ geprägte Bild festigen. Die meist narrativ inszenierte Fixierung auf Bin Laden und al Qaida erfolgt oberflächlich und vernachlässigt m

Grossforschung in neuen Dimensionen : Denker unserer Zeit über die aktuelle Elementarteilchenphysik am CERN

CERN Document Server

Blanchard, Philippe; Kommer, Christoph; ZiF-Konferenz 2013; CERN : Large-scale Research in New Dimension

2015-01-01

Der 4. Juli 2012 stellt ein historisches Datum für das Europäische Kernforschungszentrum CERN in Genf dar: die Verkündung der Entdeckung des letzten, vorhergesagten, noch fehlenden und lange gesuchten Elementarteilchens, des Higgs-Bosons. Ein Jahr später kommen im Rahmen der Konferenz des Zentrums für interdisziplinäre Forschung (ZiF) in Bielefeld eine Vielzahl der mit dem Großprojekt am CERN verbundenen Wissenschaftler zusammen, um nicht nur über die Entdeckung des Teilchens, sondern vor allem über die Realisierung und Bedeutung eines derart ungeheuren Unterfangens für solch ein wissenschaftliches Großprojekt zu berichten. Der Generaldirektor des CERN, Prof. Dr. Rolf-Dieter Heuer, liefert in seinem Eröffnungsvortrag die Grundlage für die Diskussion der vielerlei Aspekte, die dabei ins Spiel kommen. So finden sich im vorliegenden Band zum einen Berichte über die Großforschung selbst: von der Entdeckung des Higgs-Teilchens über die Erforschung des Quark-Gluon-Plasmas und die Bedeutung von Th...

A digital data acquisition and display system for ITU TRIGA Mark II reactor

International Nuclear Information System (INIS)

Can, B.; Omuz, S.

2008-01-01

Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

A New Closed Form Approximation for BER for Optical Wireless Systems in Weak Atmospheric Turbulence

Science.gov (United States)

Kaushik, Rahul; Khandelwal, Vineet; Jain, R. C.

2018-04-01

Weak atmospheric turbulence condition in an optical wireless communication (OWC) is captured by log-normal distribution. The analytical evaluation of average bit error rate (BER) of an OWC system under weak turbulence is intractable as it involves the statistical averaging of Gaussian Q-function over log-normal distribution. In this paper, a simple closed form approximation for BER of OWC system under weak turbulence is given. Computation of BER for various modulation schemes is carried out using proposed expression. The results obtained using proposed expression compare favorably with those obtained using Gauss-Hermite quadrature approximation and Monte Carlo Simulations.

Gõtter, Grãber und Gelehrte

Directory of Open Access Journals (Sweden)

Pedro Moacyr Campos

1953-09-01

Full Text Available CERAM, (C. W.. — Gõtter, Grãber und Gelehrte, Rowohlt Verlag, Hamburg.,  1951, 494 págs. (Primeiro Parágrafo do Artigo Bem merecido foi o sucesso encontrado pelo livro de Ceram, Gotter, Graber und Gelehrte, ("Deuses, túmulos e eruditos", que alcançou a cifra de 106.000 exemplares entre novembro de 1949 e março de 1951, continuando a ser reimpresso após esta data.

Computational analysis of neutronic parameters of CENM TRIGA Mark II research reactor

International Nuclear Information System (INIS)

El Younoussi, C.; El Bakkari, B.; Boulaich, Y.; Riyach, D.; Otmani, S.; Marrhich, I.; Badri, H.; Htet, A.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Zoubair, M.; Ossama, M.; Chakir, E.

2010-01-01

The CENM TRIGA MARK II reactor is part of the National Center for Energy, Sciences and Nuclear Techniques (CNESTEN). It's a standard design 2MW, natural-convection-cooled reactor with a graphite reflector containing 4 beam tubes and a thermal column. The reactor has several applications in different fields as industry, agriculture, medicine, training and education. In the present work a computational study has been carried out in the framework of neutronic parameters studies of the reactor. A detailed MCNP model that include all elements of the core and surrounding structures has been developed to calculate different parameters of the core (The effective multiplication factor, reactivity experiments comprising control rods worth, excess reactivity and shutdown margin). Further calculations have been carried out to calculate the neutron flux profiles at different locations of the reactor core. The cross sections used are processed from the library provided with MCNP5 and based on the ENDF/B-VII with continuous dependence in energy and special treatment of thermal neutrons in lightweight materials. (author)

Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

Energy Technology Data Exchange (ETDEWEB)

Tigliole, A. Borio Di; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Provasi, M.C.; Salvini, A.; Scian, G.; Vinciguerra, G. [University of Pavia, Laboratory of Applied Nuclear Energy (L.E.N.A), Via Aselli 41, 27100 Pavia (Italy)

2011-07-01

The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

Conceptual design of ICF reactor SENRI, Part II. Advances in design and pellet gain scaling

International Nuclear Information System (INIS)

Ido, S.; Mima, K.; Nakai, S.; Tsuji, R.; Yamanaka, C.

1984-01-01

This chapter reviews the recent design studies on reactor concepts with magnetically guided lithium flow, SENRI-I, SENRI-IA and SENRI-II. The routes from the present status to power reactors and an advanced fuel pellet concept is also discussed. Topics covered include pellet design, magnetohydrodynamic design of liquid lithium flow; reactor cavity concepts with magnetically guided lithium flow, a thermo-hydraulic analysis, a tritium recovery system; and an advanced fuel pellet concept for an inertial confinement fusion (ICF) reactor without a tritium breeding blanket. An advanced fuel pellet for an ICF reactor without a T breeder was studied in the model calculations, which showed sufficiently high values of pellet gain. Includes a table and 8 diagrams

Visual examination program of the TRIGA Mark II reactor Vienna with the nuclear underwater telescope

International Nuclear Information System (INIS)

Boeck, H.; Hammer, J.; Varga, K.

1985-12-01

The visual inspection programm carried out during a three month shut-period at the TRIGA Mark II reactor Vienna is described. Optical inspection of all welds inside the reactor tank was carried out with an underwater telescope developed by the Central Research Institute of Physics, Budapest, Hungary. It is shown that even after 23 years of reactor operation all tank internals were found to be in good condition and minor defects can be easily repaired by remote handling tools. (Author)

BER Analysis of Coherent Free-Space Optical Communication Systems with a Focal-Plane-Based Wavefront Sensor

Science.gov (United States)

Cao, Jingtai; Zhao, Xiaohui; Liu, Wei; Gu, Haijun

2018-03-01

A wavefront sensor is one of most important units for an adaptive optics system. Based on our previous works, in this paper, we discuss the bit-error-rate (BER) performance of coherent free space optical communication systems with a focal-plane-based wavefront sensor. Firstly, the theory of a focal-plane-based wavefront sensor is given. Then the relationship between the BER and the mixing efficiency with a homodyne receiver is discussed on the basis of binary-phase-shift-keying (BPSK) modulation. Finally, the numerical simulation results are shown that the BER will be decreased obviously after aberrations correction with the focal-plane-based wavefront sensor. In addition, the BER will decrease along with increasing number of photons received within a single bit. These analysis results will provide a reference for the design of the coherent Free space optical communication (FSOC) system.

Nine years of operation of ITU-TRR TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Yavuz, H.; Bayuelken, A.R.; Yavuz, M.A.

1988-01-01

ITU-TRR TRIGA Mark-II reactor in Istanbul with a steady state power of 250 kW and a pulsing capability up to 1200 MW has been operating since March 11,1979 with an energy release of 107.5 MWh and a total of 72 pulses. During this nearly nine years, the reactor was in operation without any major undesired shut down. One of the major problems was faced when the instrumented fuel element in position 9 of the F ring went totally out of order. Secondly, the cooling tower of the secondary cooling system could not be operated properly during the hot summer days, and also we had a tar leakage problem with the radial beam port connection to the tank. During the regular maintenance work in this summer, the measurements of changes in nuclear and physical parameters of the reactor fuel and dummy elements have also proceeded. (author)

MOTHER MK II: An advanced direct cycle high temperature gas reactor

International Nuclear Information System (INIS)

Hart, R.S.; Kendall, J.M.; Marsden, B.J.

2003-01-01

-power refueling feature of the Pebble Bed reactor core concept is attractive in many situations, the MOTHER MK II conceptual design adopts a Pebble Bed core configuration. The power conversion systems of MOTHER MKI are utilized. In an effort to overcome the disadvantages of current graphite pebble annular Pebble Bed core designs, MOTHER MK II introduces a novel split core configuration. The MOTHER concepts were developed with an objective of minimizing technical risk and the need for technology development. A principal purpose of this paper is to inform other designers currently working on direct cycle HTGR concepts of the work undertaken in defining the designs for the MOTHER nuclear power plants, and of the many novel technical features adopted. (author)

KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

International Nuclear Information System (INIS)

1978-01-01

The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

Lebensdauer von Stahlseilen beim kombinierten Lauf über Treib- und Ablenkscheiben unterschiedlicher Rillenform

OpenAIRE

Berner, Oliver Roger

2011-01-01

Die Lebensdauer von Stahldrahtseilen in Treibscheibenaufzügen wird von unterschiedlichen Faktoren beeinflusst. Bei allen Anwendungen, bei denen Seile über Rollen laufen, erfahren die Seile eine Zug- und Biegebeanspruchung, die zu einer Materialermüdung und zu einem Verschleiß der Drähte an der Seiloberfläche und im Inneren der Seile führt. Darüber hinaus findet man bei Treibscheibenaufzügen systembedingt weitere anwendungsspezifische Beanspruchungen, die zu einer zusätzlichen Lebensdauerminde...

Impact of Mutual Coupling and Polarization of Antennas on BER Performances of Spatial Multiplexing MIMO Systems

Directory of Open Access Journals (Sweden)

Jianfeng Zheng

2012-01-01

Full Text Available This paper is aimed at studying the impacts of mutual coupling, matching networks, and polarization of antennas on performances of Multiple-Input Multiple-Output (MIMO systems employing Spatial Multiplexing (SM. In particular, the uncoded average Bit Error Rate (BER of MIMO systems is investigated. An accurate signal analysis framework based on circuit network parameters is presented to describe the transmit/receive characteristics of the matched/unmatched antenna array. The studied arrays consist of matched/unmatched compact copolarization and polarization diversity antenna array. Monte-Carlo numerical simulations are used to study the BER performances of the SM MIMO systems using maximum-likelihood and/or zero-forcing detection schemes. The simulation results demonstrate that the use of matching networks can improve the BER performance of SM MIMO systems significantly, and the BER performance deterioration due to antenna orientation randomness can be compensated by use of polarization diversity antenna arrays.

Time constants and feedback transfer functions of EBR-II [Experimental Breeder Reactor] subassembly types

International Nuclear Information System (INIS)

Grimm, K.N.; Meneghetti, D.

1986-09-01

Time constants, feedback reactivity transfer functions and power coefficients are calculated for stereotypical subassemblies in the EBR-II reactor. These quantities are calculated from nodal reactivities obtained from a reactor kinetic code analysis for a step change in power. Due to the multiplicity of eigenvalues, there are several time constants for each nodal position in a subassembly. Compared with these calculated values are analytically derived values for the initial node of a given channel

Expert's statement on the research reactor Munich II (FRM-II); Gutachterliche Stellungnahme zum Forschungsreaktor Muenchen II (FRM-II)

Energy Technology Data Exchange (ETDEWEB)

Liebert, Wolfgang; Friess, Friederike; Gufler, Klaus; Arnold, Nikolaus [Univ. fuer Bodenkultur (BOKU), Wien (Austria). Inst. fuer Sicherheits- und Risikowissenschaften (ISR)

2017-12-15

The Expert's statement on the research reactor FRM-II covers the following issues: The situation in Germany with respect to HEU (highly enriched uranium) fuel elements, the proliferation problems related to HEU fuel and the generated high-level radioactive wastes, possible safety hazards of an interim storage of HEU containing wastes, for instance in the interim storage facility Ahaus, possible safety hazards of final disposal of HEU containing radioactive wastes, possibilities to avoid the use of HEU fuel in order to prevent further production of these wastes, requirement of processing spent HEU containing fuel elements for final disposal.

Effects of Electrical Stimulation on the Degradation of Azo Dye in Three-Dimensional Biofilm Electrode Reactors

Directory of Open Access Journals (Sweden)

Xian Cao

2017-04-01

Full Text Available Three-dimensional biofilm electrode reactors (3D-BERs were constructed to degrade the azo dye Reactive Brilliant Red (RBR X-3B. The 3D-BERs with different influent concentrations and external voltages were individually studied to investigate their influence on the removal of X-3B. Experimental results showed that 3D-BERs have good X-3B removal efficiency; even when the influent concentration was 800 mg/L, removal efficiency of 73.4% was still achieved. In addition, the X-3B removal efficiency stabilized shortly after the influent concentration increased. In 3D-BERs, the average X-3B removal efficiency increased from 52.8% to 85.4% when the external voltage rose from 0 to 2 V. We further identified the intermediate products via UV-Vis and gas chromatography-mass spectrometry (GC-MS analyses, and discussed the potential mechanism of degradation. After the conjugate structure of X-3B was destroyed, all of the substances generated mainly consisted of lower-molecular-weight organics.

Real-time dynamic simulator for the Topaz II reactor power system

International Nuclear Information System (INIS)

Kwok, K.S.

1994-01-01

A dynamic simulator of the TOPAZ II reactor system has been developed for the Nuclear Electric Propulsion Space Test Program. The simulator is a self-contained IBM-PC compatible based system that executes at a speed faster than real-time. The simulator combines first-principle modeling and empirical correlations in its algorithm to attain the modeling accuracy and computational through-put that are required for real-time execution. The overall execution time of the simulator for each time step is 15 ms when no data is written to the disk, and 18 ms when nine double precision data points are written to the disk once in every time step. The simulation program has been tested and it is able to handle a step decrease of $8 worth of reactivity. It also provides simulation of fuel, emitter, collector, stainless steel, and ZrH moderator failures. Presented in this paper are the models used in the calculations, a sample simulation session, and a discussion of the performance and limitations of the simulator. The simulator has been found to provide realistic real-time dynamic response of the TOPAZ II reactor system under both normal and causality conditions

Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

International Nuclear Information System (INIS)

Muswema, J.L.; Ekoko, G.B.; Lukanda, V.M.; Lobo, J.K.-K.; Darko, E.O.; Boafo, E.K.

2015-01-01

Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

Energy Technology Data Exchange (ETDEWEB)

Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

2015-01-15

Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

International Nuclear Information System (INIS)

Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.

2006-01-01

A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

On the BER and capacity analysis of MIMO MRC systems with channel estimation error

KAUST Repository

Yang, Liang

2011-10-01

In this paper, we investigate the effect of channel estimation error on the capacity and bit-error rate (BER) of a multiple-input multiple-output (MIMO) transmit maximal ratio transmission (MRT) and receive maximal ratio combining (MRC) systems over uncorrelated Rayleigh fading channels. We first derive the ergodic (average) capacity expressions for such systems when power adaptation is applied at the transmitter. The exact capacity expression for the uniform power allocation case is also presented. Furthermore, to investigate the diversity order of MIMO MRT-MRC scheme, we derive the BER performance under a uniform power allocation policy. We also present an asymptotic BER performance analysis for the MIMO MRT-MRC system with multiuser diversity. The numerical results are given to illustrate the sensitivity of the main performance to the channel estimation error and the tightness of the approximate cutoff value. © 2011 IEEE.

Storage temperature affects fruit quality attributes of Ber ( Ziziphus ...

African Journals Online (AJOL)

Fruit utilization is affected by quality attributes and shelf life. The quality of Jujube or Ber (Ziziphus mauritiana Lamk.) fruits after harvest depends on storage conditions used. In this study, different storage temperatures and durations were evaluated to determine the appropriate storage conditions of fresh fruits of Z.

Decreased expression level of BER genes in Alzheimer's disease patients is not derivative of their DNA methylation status.

Science.gov (United States)

Sliwinska, Agnieszka; Sitarek, Przemysław; Toma, Monika; Czarny, Piotr; Synowiec, Ewelina; Krupa, Renata; Wigner, Paulina; Bialek, Katarzyna; Kwiatkowski, Dominik; Korycinska, Anna; Majsterek, Ireneusz; Szemraj, Janusz; Galecki, Piotr; Sliwinski, Tomasz

2017-10-03

Neurodegeneration in Alzheimer's disease can be caused by accumulation of oxidative DNA damage resulting from altered expression of genes involved in the base excision repair system (BER). Promoter methylation can affect the profile of BER genes expression. Decreased expression of BER genes was observed in the brains of AD patients. The aim of our study was to compare the expression and methylation profiles of six genes coding for proteins involved in BER, namely: hOGG1, APE1, MUTYH, NEIL1, PARP1 and XRCC1, in the peripheral blood cells of AD patients and healthy volunteers. The study consisted of 100 persons diagnosed with Alzheimer's disease according to DSM-IV criteria, and 110 healthy volunteers. DNA and total RNA were isolated from venous blood cells. Promoter methylation profiles were obtained by High Resolution Melting (HRM) analysis of bisulfide converted DNA samples. Real-time PCR with TaqMan probes was employed for gene expression analysis. APE1, hOGG1, MUTYH, PARP1 and NEIL1 were significantly (pgenes. The methylation status of promoters is not associated with downregulation of BER genes. Our results show that downregulation of BER genes detected in peripheral blood samples could reflect the changes occurring in the brain of patients with AD, and may be a useful biomarker of this disease. Copyright © 2017 Elsevier Inc. All rights reserved.

Present Services at the TRIGA Mark II Reactor of the JSI

International Nuclear Information System (INIS)

Smodiš, B.; Snoj, L.

2013-01-01

The TRIGA Mark II research reactor of the Jožef Stefan Institute has been continuously operating since the year 1966. The currently offered services include: (1) Neutron activation analysis in both instrumental and radiochemical modes; (2) neutron irradiation of various kinds of materials intended to be used for research and applicative purposes; (3) training and education of university students as well as on-job training of staff working in public and private institutions, (4) verification of computer codes and nuclear data, comprising primarily criticality calculations and neutron flux distribution studies and (5) testing and development of a digital reactivity meter. The paper briefly describes the aforementioned activities and shows that even such small reactors are still indispensable in nuclear science and technology. (author)

Culham Conceptual Tokamak Mark II. Design study of the layout of a twin-reactor fusion power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-07-01

This report describes the building layout and outline design for the nuclear complex of a fusion reactor power station incorporating two Culham Conceptual Tokamak Reactors Mk.II. The design incorporates equipment for steam generation, process services for the fusion reactors and all facilities for routine and non-routine servicing of the nuclear complex. The design includes provision of temporary facilities for on site construction of the major reactor components and shows that these facilities may be used for disassembly of the reactors either for major repair and/or decommissioning. Preliminary estimates are included, which indicate the cost benefits to be obtained from incorporating two reactors in one nuclear complex and from increased wall loading. (author)

Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

International Nuclear Information System (INIS)

Guthrie, J.A.S.

1980-04-01

In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

Effect of Pointing Error on the BER Performance of an Optical CDMA FSO Link with SIK Receiver

Science.gov (United States)

Nazrul Islam, A. K. M.; Majumder, S. P.

2017-12-01

An analytical approach is presented for an optical code division multiple access (OCDMA) system over free space optical (FSO) channel considering the effect of pointing error between the transmitter and the receiver. Analysis is carried out with an optical sequence inverse keying (SIK) correlator receiver with intensity modulation and direct detection (IM/DD) to find the bit error rate (BER) with pointing error. The results are evaluated numerically in terms of signal-to-noise plus multi-access interference (MAI) ratio, BER and power penalty due to pointing error. It is noticed that the OCDMA FSO system is highly affected by pointing error with significant power penalty at a BER of 10-6 and 10-9. For example, penalty at BER 10-9 is found to be 9 dB corresponding to normalized pointing error of 1.4 for 16 users with processing gain of 256 and is reduced to 6.9 dB when the processing gain is increased to 1,024.

Solution structure, copper binding and backbone dynamics of recombinant Ber e 1-the major allergen from Brazil nut.

Directory of Open Access Journals (Sweden)

Louise Rundqvist

Full Text Available BACKGROUND: The 2S albumin Ber e 1 is the major allergen in Brazil nuts. Previous findings indicated that the protein alone does not cause an allergenic response in mice, but the addition of components from a Brazil nut lipid fraction were required. Structural details of Ber e 1 may contribute to the understanding of the allergenic properties of the protein and its potential interaction partners. METHODOLOGY/PRINCIPAL FINDINGS: The solution structure of recombinant Ber e 1 was solved using NMR spectroscopy and measurements of the protein back bone dynamics at a residue-specific level were extracted using (15N-spin relaxation. A hydrophobic cavity was identified in the structure of Ber e 1. Using the paramagnetic relaxation enhancement property of Cu(2+ in conjunction with NMR, it was shown that Ber e 1 is able to specifically interact with the divalent copper ion and the binding site was modeled into the structure. The IgE binding region as well as the copper binding site show increased dynamics on both fast ps-ns timescale as well as slower µs-ms timescale. CONCLUSIONS/SIGNIFICANCE: The overall fold of Ber e 1 is similar to other 2S albumins, but the hydrophobic cavity resembles that of a homologous non-specific lipid transfer protein. Ber e 1 is the first 2S albumin shown to interact with Cu(2+ ions. This Cu(2+ binding has minimal effect on the electrostatic potential on the surface of the protein, but the charge distribution within the hydrophobic cavity is significantly altered. As the hydrophobic cavity is likely to be involved in a putative lipid interaction the Cu(2+ can in turn affect the interaction that is essential to provoke an allergenic response.

Lipman Bers, a life in mathematics

CERN Document Server

Keen, Linda; Rodríguez, Rubí E

2015-01-01

The book is part biography and part collection of mathematical essays that gives the reader a perspective on the evolution of an interesting mathematical life. It is all about Lipman Bers, a giant in the mathematical world who lived in turbulent and exciting times. It captures the essence of his mathematics, a development and transition from applied mathematics to complex analysis-quasiconformal mappings and moduli of Riemann surfaces-and the essence of his personality, a progression from a young revolutionary refugee to an elder statesman in the world of mathematics and a fighter for global h

Experimental Breeder Reactor-II automatic control-rod-drive system

International Nuclear Information System (INIS)

Christensen, L.J.

1983-01-01

A computer-controlled automatic control rod drive system (ACRDS) was designed and operated in EBR-II during reactor runs 121 and 122. The ACRDS was operated in a checkout mode during run 121 using a low worth control rod. During run 122 a high worth control rod was used to perform overpower transient tests as part of the LMFBR oxide fuels transient testing program. The testing program required an increase in power of 4 MW/s, a hold time of 12 minutes and a power decrease of 4 MW/s. During run 122, 13 power transients were performed

Paradigmenwechsel in der politischen Ökologie? Zur deutschen Diskussion über die Lokale Agenda 21

OpenAIRE

Pettenkofer, Andreas

2001-01-01

Der Text berichtet über die in Deutschland geführte sozialwissenschaftliche Diskussion über das umweltpolitische Programm Lokale Agenda 21; er versucht eine vorläufige Rekonstruktion der spezifischen Gestalt der Agenda-Prozesse und ihrer Abweichungen von der westdeutschen Tradition der Politischen Ökologie, und diskutiert die theoretischen Fragen, die sich angesichts dieser neuartigen Phänomene stellen. Dabei geht es vor allem um die Angemessenheit der modernisierungstheoretischen Perspektive...

Das Volkslied in Deutschland, Frankreich, Belgien und Holland. Untersuchungen über die Auffassung des Begriffes; Über die traditionellen Zeilen, die Zahlen-, Blumen- und Farbensymbolik

NARCIS (Netherlands)

Brouwer, Cornelis

1930-01-01

Wenn man die gewaltige Menge Publikationen über das Volkslied, die im Laufe des 19. und 20. Jahrhunderts in Deutschland, Frankreich, Belgien und Holland erschienen sind, vergleicht, springen zwei Punkte sofort ins Auge: 1. der Unterschied zwischen den verschiedenen Auffassungen vom Begriff Volkslied

Notizen über Thatcheria Angas, Clinura Bellardi und Clinuropsis Vincent

NARCIS (Netherlands)

Beets, C.

1942-01-01

1. Es sind bisher über die merkwürdige rezente marine Gastropoden-Gruppe Thatcheria Angas, 1877 sehr wenige Gegebenheiten bekannt, vergleichende Untersuchungen mit fossilen Arten fehlen (fast) vollkommen und auch die Frage ihrer systematischen Stellung ist nie in befriedigender Weise gelöst worden.

Regulation of Burkholderia cenocepacia biofilm formation by RpoN and the c-di-GMP effector BerB

DEFF Research Database (Denmark)

Fazli, Mustafa; Rybtke, Morten Levin; Steiner, Elisabeth

2017-01-01

Knowledge about the molecular mechanisms that are involved in the regulation of biofilm formation is essential for the development of biofilm-control measures. It is well established that the nucleotide second messenger cyclic diguanosine monophosphate (c-di-GMP) is a positive regulator of biofilm...... formation in many bacteria, but more knowledge about c-di-GMP effectors is needed. We provide evidence that c-di-GMP, the alternative sigma factor RpoN (σ54), and the enhancer-binding protein BerB play a role in biofilm formation of Burkholderia cenocepacia by regulating the production of a biofilm......-stabilizing exopolysaccharide. Our findings suggest that BerB binds c-di-GMP, and activates RpoN-dependent transcription of the berA gene coding for a c-di-GMP-responsive transcriptional regulator. An increased level of the BerA protein in turn induces the production of biofilm-stabilizing exopolysaccharide in response to high...

Implementation of strength pareto evolutionary algorithm II in the multiobjective burnable poison placement optimization of KWU pressurized water reactor

International Nuclear Information System (INIS)

Gharari, Rahman; Poursalehi, Navid; Abbasi, Mohmmadreza; Aghale, Mahdi

2016-01-01

In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor (K-e-f-f) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor

Implementation of strength pareto evolutionary algorithm II in the multiobjective burnable poison placement optimization of KWU pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Gharari, Rahman [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of); Poursalehi, Navid; Abbasi, Mohmmadreza; Aghale, Mahdi [Nuclear Engineering Dept, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

2016-10-15

In this research, for the first time, a new optimization method, i.e., strength Pareto evolutionary algorithm II (SPEA-II), is developed for the burnable poison placement (BPP) optimization of a nuclear reactor core. In the BPP problem, an optimized placement map of fuel assemblies with burnable poison is searched for a given core loading pattern according to defined objectives. In this work, SPEA-II coupled with a nodal expansion code is used for solving the BPP problem of Kraftwerk Union AG (KWU) pressurized water reactor. Our optimization goal for the BPP is to achieve a greater multiplication factor (K-e-f-f) for gaining possible longer operation cycles along with more flattening of fuel assembly relative power distribution, considering a safety constraint on the radial power peaking factor. For appraising the proposed methodology, the basic approach, i.e., SPEA, is also developed in order to compare obtained results. In general, results reveal the acceptance performance and high strength of SPEA, particularly its new version, i.e., SPEA-II, in achieving a semioptimized loading pattern for the BPP optimization of KWU pressurized water reactor.

Neutronics analysis of TRIGA Mark II research reactor

Directory of Open Access Journals (Sweden)

Haseebur Rehman

2018-02-01

Full Text Available This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4 and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE codes. Cores 133 and 134 were analyzed in 2-D (r, θ and 3-D (r, θ, z, using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0, Joint Evaluated Fission and Fusion File (JEFF-3.1, Japanese Evaluated Nuclear Data Library (JENDL-3.2, and Joint Evaluated File (JEF-2.2 nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

Identification of nuclear reactor characteristics by the reactor noise analysis

International Nuclear Information System (INIS)

Yashima, Hideyuki

1980-01-01

Reactor noise analysis method was applied to TRIGA II Research Reactor (Atomic Research Laboratory, Musashi Institute of Technology) and computed power spectral density (PSD) from the CIC current record. PSD has provided many valuable informations regarding to the reactor kinetics, including the effect of control rods vibration. Another information of neutron physics parameters were obtained and this result was compared with the parameter which was formerly measured by the Feynman-α experiment. Through these experiments we could find overall frequency characteristics of TRIGA II Reactor. (author)

Twenty years of operation of Ljubljana's TRIGA Mark II reactor

International Nuclear Information System (INIS)

Dimic, V.

1986-01-01

Twenty years have now passed since the start of the TRIGA Mark II reactor in Ljubljana. The reactor was critical on May 31, 1966. The total energy produced until the end of May 1986 was 14.048 MWh or 585 MWd. For the first 14 years (until 1981) the yearly energy produced was about 600 MWh, since 1981 the yearly energy produced was 1000 MWh when a routine radioactive isotopes production started for medical use as well as other industrial applications, such as doping and irradiation with fast neutrons of silicon monocrystals, production of level indicators (irradiated cobalt wire), production of radioactive iridium for gamma-radiography, leak detection in pipes by sodium, etc. Besides these, applied research around the reactor is being conducted in the following main fields, where- many unique methods have been developed or have found their way into the local industry or hospitals: neutron radiography, neutron induced auto-radiography using solid state nuclear track detectors, nondestructive methods for assessment of nuclear burn-up, neutron dosimetry, calculation of core burn-up for the optimal in-core fuel management strategy. The solvent extraction method was developed for the everyday production of 99m Tc, which is the most widely used radionuclide in diagnostic nuclear medicine. The methods were developed for the production of the following isotopes: 18 F, 85m Kr, 24 Na, 82 Br, 64 Zn, 125 I. Neutron activation analysis represents one of the major usages for the TRIGA reactor. Basic research is being conducted in the following main fields: solid state physics (elastic and inelastic scattering of the neutrons), neutron dosimetry, neutron radiography, reactor physics and neutron activation analysis. The reactor is used very extensively as a main instrument in the Reactor Training Centre in Ljubljana where manpower training for our nuclear power plant and other organisations has been performed. Although the reactor was designed very carefully in order to be used for

Implementation of MOAS II diagnosis system at the OECD Halden Reactor project

International Nuclear Information System (INIS)

Kim, I.S.; Grini, R.E.; Nilsen, S.

1995-01-01

MOAS II is a surveillance and diagnosis system that uses several techniques for knowledge acquisition and diagnostic reasoning, e.g., goal tree-success tree, simplified directed graphs, diagnosis trees, and detailed knowledge of the process, such as mass or energy balance. This new approach was used at the Halden Man-Machine Laboratory of the OECD Halden Reactor Project. The performance of MOAS II, developed in G2 real-time expert system shell for the high-pressure preheaters of the NORS process, was tested against a variety of transient scenarios, including failures of control valves and sensors, and leakage of tubes of the preheaters. These tests showed that MOAS II successfully carried out its intended functions, i.e., quickly recognizing an occurring disturbance, correctly diagnosing its cause, and presenting advice on its control to the operator. The insights gained during the implementation are discussed

Investigation of the transition from forced to natural convection in the research reactor Munich II

International Nuclear Information System (INIS)

Skreba, S.; Adamek, J.; Unger, H.

1999-01-01

The new research reactor Munich II (FRM-II), which is under construction at the Technical University Munich, Germany, makes use of a newly developed compact reactor core consisting of a single fuel element, which is assembled of two concentric pipes. Between the fuel element's inner and outer pipe 113 involutely bent fuel plates are placed rotationally symmetric, forming 113 cooling channels of a constant width of 2.2 mm. After a shut down of the reactor, battery supported cooling pumps are started by the reactor safety system in order to remove the decay heat by a downwards directed forced flow. Three hours after they have been started, the cooling pumps are shut down and so-called 'natural convection flaps' are opened by their own weight. Through a flow path, which is provided by the opening of the natural convection flaps, the decay heat is given off to the water in the reactor pool after the direction of the flow has changed and an upwards directed natural convection flow has developed. At the Department for Nuclear and New Energy Systems of the Ruhr-University Bochum, Germany, a test facility has been built in order to confirm the concept of the decay heat removal in the FRM-II, to acquire data of single and two phase natural convection flows and to detect the dry out in a narrow channel. The thermohydraulics of the FRM-II are simulated by an electrically heated test section, which represents one cooling channel of the fuel element. At first experiments have been performed, which simulated the transition from forced to natural convection in the core of the FRM-II, both at normal operation and at a complete loss of the decay heat removal pumps. In case of normal operation, the transition from forced to natural convection takes place single phased. If a complete loss of the active decay heat removal system occurs, the decay heat removal is ensured by a quasi-steady two phase flow. In a second test series minimum heat flux densities leading to pressure pulsations

H Reactor

Data.gov (United States)

Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

Die Tischglocke Rudolfs II: über ihren Urheber und ihre Bedeutung

Czech Academy of Sciences Publication Activity Database

Bukovinská, Beket; Purš, Ivo

-, č. 10 (2010), s. 89-104 ISSN 1213-5372 Institutional research plan: CEZ:AV0Z80330511 Keywords : Rudolphinian art * Rudolf II * magic * hermetic philosophy Subject RIV: AL - Art, Architecture, Cultural Heritage

Importance of reduced sulfur for the equilibrium chemistry and kinetics of Fe(II), Co(II) and Ni(II) supplemented to semi-continuous stirred tank biogas reactors fed with stillage

Energy Technology Data Exchange (ETDEWEB)

Shakeri Yekta, Sepehr, E-mail: sepehr.shakeri.yekta@liu.se [Department of Thematic Studies – Water and Environmental Studies, Linköping University, SE-581 83 Linköping (Sweden); Lindmark, Amanda [Department of Thematic Studies – Water and Environmental Studies, Linköping University, SE-581 83 Linköping (Sweden); Skyllberg, Ulf [Department of Forest Ecology and Management, Swedish University of Agricultural Sciences, SE-901 83 Umeå (Sweden); Danielsson, Åsa; Svensson, Bo H. [Department of Thematic Studies – Water and Environmental Studies, Linköping University, SE-581 83 Linköping (Sweden)

2014-03-01

Highlights: • Thermodynamics and kinetics of Fe, Co and Ni added to biogas reactors were studied. • Formation of Fe-sulfide and Fe-thiol aqueous complexes controlled the Fe solubility. • Cobalt solubility was controlled by processes independent of Co-sulfide interaction. • Iron added to the biogas reactors effected the Ni speciation and solubility. - Abstract: The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes.

Importance of reduced sulfur for the equilibrium chemistry and kinetics of Fe(II), Co(II) and Ni(II) supplemented to semi-continuous stirred tank biogas reactors fed with stillage

International Nuclear Information System (INIS)

Shakeri Yekta, Sepehr; Lindmark, Amanda; Skyllberg, Ulf; Danielsson, Åsa; Svensson, Bo H.

2014-01-01

Highlights: • Thermodynamics and kinetics of Fe, Co and Ni added to biogas reactors were studied. • Formation of Fe-sulfide and Fe-thiol aqueous complexes controlled the Fe solubility. • Cobalt solubility was controlled by processes independent of Co-sulfide interaction. • Iron added to the biogas reactors effected the Ni speciation and solubility. - Abstract: The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes

Plans for the utilization of a new research reactor FRM II

International Nuclear Information System (INIS)

Glaeser, W.

1999-01-01

The construction of the new research reactor FRM II is close to completion. The start of nuclear operation is planned for the year 2001. After a short description of the concept and figures of merit of the facility, the scientific instrumentation and user installations for basic and applied research (worked out largely by the German user community) being under construction will be summarized. Besides the introduction of several new techniques considerable progress in the performance of standard neutron techniques is envisaged. (author)

Adaptive control method for core power control in TRIGA Mark II reactor

Science.gov (United States)

Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

2018-01-01

The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

BER analysis of DS-UWB system employing a laplace distribution model

KAUST Repository

Mehbodniya, Abolfazl; Aissa, Sonia; Adachi, Fumiyuki

2011-01-01

This letter takes a new approach to extract a closed-form expression for the bit error rate (BER) of direct-sequence ultra wideband (DS-UWB) system. In the analysis, the main signal is impaired by multi-user interference (MUI) and an external source

Feasibility study on commercialized fast reactor cycle systems. Phase II final report

International Nuclear Information System (INIS)

Ieda, Yoshiaki; Uchikawa, Sadao; Okubo, Tsutomu; Ono, Kiyoshi; Kato, Atsushi; Kurisaka, Kenichi; Sakamoto, Yoshihiko; Sato, Kazujiro; Sato, Koji; Chikazawa, Yoshitaka; Nakai, Ryodai; Nakabayashi, Hiroki; Nakamura, Hirofumi; Namekawa, Takashi; Niwa, Hajime; Nomura, Kazunori; Hayashi, Hideyuki; Hayafune, Hiroki; Hirao, Kazunori; Mizuno, Tomoyasu; Muramatsu, Toshiharu; Ando, Masato; Ono, Katsumi; Ogata, Takanari; Kubo, Shigenobu; Kotake, Shoji; Sagayama, Yutaka; Takakuma, Katsuyuki; Tanaka, Toshihiko; Namba, Takashi; Fujii, Sumio; Muramatsu, Kazuyoshi

2006-06-01

A joint project team of Japan Atomic Energy Agency and the Japan Atomic Power Company (as the representative of the electric utilities) started the feasibility study on commercialized fast reactor cycle systems (F/S) in July 1999 in cooperation with Central Research Institute of Electric Power Industry and vendors. On the major premise of safety assurance, F/S aims to present an appropriate picture of commercialization of fast reactor (FR) cycle system which has economic competitiveness with light water reactor cycle systems and other electricity base load systems, and to establish FR cycle technologies for the future major energy supply. In the period from Japanese fiscal year (JFY) 1999 to 2000, the phase-I of F/S was carried out to screen our representative FR, reprocessing and fuel fabrication technologies. In the phase-II (JFY 2001-2005), the design study of FR cycle concepts, the development of significant technologies necessary for the feasibility evaluation, and the confirmation of key technical issues were performed to clarify the promising candidate concepts toward the commercialization. In this final phase-II report clarified the most promising concept, the R and D plan until around 2015, and the key issues for the commercialization. Based on the comprehensive evaluation in F/S, the combination of the sodium-cooled FR with MOX fuel core, the advanced-aqueous reprocessing process and the simplified-pelletizing fuel fabrication process was recommended as the mainline choice for the most promising concept. The concept exceeds in technical advancement, and the conformity to the development targets was higher compared with that of the others. Alternative technologies are prepared to be decrease the development risk of innovative technologies in the mainline choice. (author)

BER Science Network Requirements Workshop -- July 26-27,2007

Energy Technology Data Exchange (ETDEWEB)

Tierney, Brian L.; Dart, Eli

2008-02-01

The Energy Sciences Network (ESnet) is the primary provider of network connectivity for the US Department of Energy Office of Science, the single largest supporter of basic research in the physical sciences in the United States of America. In support of the Office of Science programs, ESnet regularly updates and refreshes its understanding of the networking requirements of the instruments, facilities, scientists, and science programs that it serves. This focus has helped ESnet to be a highly successful enabler of scientific discovery for over 20 years. In July 2007, ESnet and the Biological and Environmental Research (BER) Program Office of the DOE Office of Science organized a workshop to characterize the networking requirements of the science programs funded by the BER Program Office. These included several large programs and facilities, including Atmospheric Radiation Measurement (ARM) Program and the ARM Climate Research Facility (ACRF), Bioinformatics and Life Sciences Programs, Climate Sciences Programs, the Environmental Molecular Sciences Laboratory at PNNL, the Joint Genome Institute (JGI). National Center for Atmospheric Research (NCAR) also participated in the workshop and contributed a section to this report due to the fact that a large distributed data repository for climate data will be established at NERSC, ORNL and NCAR, and this will have an effect on ESnet. Workshop participants were asked to codify their requirements in a 'case study' format, which summarizes the instruments and facilities necessary for the science and the process by which the science is done, with emphasis on the network services needed and the way in which the network is used. Participants were asked to consider three time scales in their case studies--the near term (immediately and up to 12 months in the future), the medium term (3-5 years in the future), and the long term (greater than 5 years in the future). In addition to achieving its goal of collecting and

Effect of Post-Harvest Treatments on Enzyme Activity and Quality of Cold Stored Ber Fruit

Directory of Open Access Journals (Sweden)

Sukhjit K. JAWANDHA

2012-11-01

Full Text Available Ber (Zizyphus mauritiana Lamk fruits are perishable in nature and have poor shelf- life. To extend the storage life of ber fruits different post-harvest treatments, like CaCl2 (0.5, 1.0 and 2.0%, Ca(NO32 (0.5, 1.0 and 2.0%, gibberellic acid (GA3 (20, 40 and 60 ppm and Bavistin (0.1% were given to fruits and their effect was studied on the storage life and quality of Umran ber (Zizyphus mauritiana Lamk fruits under cold storage conditions. Fruits of uniform size and colour were harvested, from healthy plants, and subjected to post harvest dip of different chemicals, before packing in CFB boxes and placed in cold storage (3-5C and 85-90% RH They have been evaluated after 10, 20 and 30 days of storage for physiological loss in weight (PLW, firmness, pectin methyl esterase (PME activity and palatability rating. The PLW increased and fruit firmness decreased during storage at each treatment. The sensory rating increased up to 10 days at all treatments, except control but subsequently it decreased during storage. The increase rate in PME activity was faster, up to 10 days, in untreated fruits, while treated fruits showed slower rise in PME activity, up to 20 days with a decline afterwards. It can be concluded that ber fruits can be stored up to 20 days by post-harvest treatment using GA3 at 60 ppm, with acceptable quality.

Effect of Post-Harvest Treatments on Enzyme Activity and Quality of Cold Stored Ber Fruit

Directory of Open Access Journals (Sweden)

Sukhjit K. JAWANDHA

2012-11-01

Full Text Available Ber (Zizyphus mauritiana Lamk fruits are perishable in nature and have poor shelf- life. To extend the storage life of ber fruits different post-harvest treatments, like CaCl2 (0.5, 1.0 and 2.0%, Ca(NO32 (0.5, 1.0 and 2.0%, gibberellic acid (GA3 (20, 40 and 60 ppm and Bavistin (0.1% were given to fruits and their effect was studied on the storage life and quality of �Umran� ber (Zizyphus mauritiana Lamk fruits under cold storage conditions. Fruits of uniform size and colour were harvested, from healthy plants, and subjected to post harvest dip of different chemicals, before packing in CFB boxes and placed in cold storage (3-5�C and 85-90% RH They have been evaluated after 10, 20 and 30 days of storage for physiological loss in weight (PLW, firmness, pectin methyl esterase (PME activity and palatability rating. The PLW increased and fruit firmness decreased during storage at each treatment. The sensory rating increased up to 10 days at all treatments, except control but subsequently it decreased during storage. The increase rate in PME activity was faster, up to 10 days, in untreated fruits, while treated fruits showed slower rise in PME activity, up to 20 days with a decline afterwards. It can be concluded that ber fruits can be stored up to 20 days by post-harvest treatment using GA3 at 60 ppm, with acceptable quality.

BER Reduction for mQAM High-Speed Tranmission in Optical Fiber ...

African Journals Online (AJOL)

The study was conducted in a software cosimulation environment with VPITransmissionMaker and Matlab software. The simulation results showed that error correcting codes and particularly LDPC codes are effective and provide satisfactory solutions to reduce the BER by fighting against optical transmission channel ...

Different microprocessor controlled devices for ITU TRIGA Mark II reactor

International Nuclear Information System (INIS)

Can, B.; Omuz, S.; Uzun, S.; Apan, H.

1990-01-01

In this paper the design of a period meter and multichannel thermometer, which are controlled by a microprocessor, in order to be used at ITU TRIGA Mark-II Reactor is presented. The system works as a simple microcomputer, which includes a CPU, a EPROM, a RAM, a CTC, a PIO, a PIA a keyboard and displays, using the assembly language. The period meter can work either with pulse signal or with analog signal depending on demand of the user. The period is calculated by software and its range is -99,9 sec, to +2.1 sec. When the period drops +3 sec, the system gives alarm illuminating a LED. The multichannel thermometer has eight temperature channels. Temperature channels can manually or automatically be selected. The channel selection time can be adjusted. The thermometer gives alarm illuminating a LED, when the temperature rises to 600 C. Temperature data is stored in the RAM and is shown on a display. This system provides us to use four spare thermocouples in the reactor. (orig.)

Untersuchungen über Veronica triloba (Opiz) Kerner in Rheinhessen

OpenAIRE

Oesau, Albert

1981-01-01

Untersuchungen in den Jahren 1977-1979 ergaben, daß Veronica triloba (Opiz) Kerner, eine Sippe aus dem Veronica hederifolia-Aggregat, in Rheinhessen überraschenderweise weit verbreitet ist und bisher übersehen wurde. Sie siedelt vor allem in Ackerunkraut- und Felsgrus-Gesellschaften sowie lückigen Trockenrasen, ohne eine gesellschaftsspezifische Bindung zu zeigen. Gut ausgeprägte morphologische Merkmale gestatten eine sichere Differenzierung gegenüber den übrigen Sippenvertretern V. hederifol...

Clustered DNA lesions containing 5-formyluracil and AP site: repair via the BER system.

Directory of Open Access Journals (Sweden)

Ekaterina A Belousova

Full Text Available Lesions in the DNA arise under ionizing irradiation conditions or various chemical oxidants as a single damage or as part of a multiply damaged site within 1-2 helical turns (clustered lesion. Here, we explored the repair opportunity of the apurinic/apyrimidinic site (AP site composed of the clustered lesion with 5-formyluracil (5-foU by the base excision repair (BER proteins. We found, that if the AP site is shifted relative to the 5-foU of the opposite strand, it could be repaired primarily via the short-patch BER pathway. In this case, the cleavage efficiency of the AP site-containing DNA strand catalyzed by human apurinic/apyrimidinic endonuclease 1 (hAPE1 decreased under AP site excursion to the 3'-side relative to the lesion in the other DNA strand. DNA synthesis catalyzed by DNA polymerase lambda was more accurate in comparison to the one catalyzed by DNA polymerase beta. If the AP site was located exactly opposite 5-foU it was expected to switch the repair to the long-patch BER pathway. In this situation, human processivity factor hPCNA stimulates the process.

Erstellung eines multimedialen Lernprogramms über die Grundlagen der EKG-Aufzeichnung und Auswertung bei Hund und Katze

OpenAIRE

Weigel-Ossiander, Fiona Sarah

2016-01-01

Mit EKG Basics wurde ein detailliertes multimediales Lernprogramm über die Grundlagen der Elektrokardiographie bei Hund und Katze erstellt unter besonderer Berücksichtigung von didaktischen und ergonomischen Gesichtspunkten. Da vielen Veterinärmedizinerinnen/Veterinärmedizinern das Erlernen, Verstehen und Interpretieren der Elektrokardiographie schwer fällt, war das Ziel der vorliegenden Arbeit, sowohl Studierenden der Tiermedizin als auch interessierten Tierärztinnen/Tierärzten die Thema...

NSGA-II Algorithm with a Local Search Strategy for Multiobjective Optimal Design of Dry-Type Air-Core Reactor

Directory of Open Access Journals (Sweden)

Chengfen Zhang

2015-01-01

Full Text Available Dry-type air-core reactor is now widely applied in electrical power distribution systems, for which the optimization design is a crucial issue. In the optimization design problem of dry-type air-core reactor, the objectives of minimizing the production cost and minimizing the operation cost are both important. In this paper, a multiobjective optimal model is established considering simultaneously the two objectives of minimizing the production cost and minimizing the operation cost. To solve the multi-objective optimization problem, a memetic evolutionary algorithm is proposed, which combines elitist nondominated sorting genetic algorithm version II (NSGA-II with a local search strategy based on the covariance matrix adaptation evolution strategy (CMA-ES. NSGA-II can provide decision maker with flexible choices among the different trade-off solutions, while the local-search strategy, which is applied to nondominated individuals randomly selected from the current population in a given generation and quantity, can accelerate the convergence speed. Furthermore, another modification is that an external archive is set in the proposed algorithm for increasing the evolutionary efficiency. The proposed algorithm is tested on a dry-type air-core reactor made of rectangular cross-section litz-wire. Simulation results show that the proposed algorithm has high efficiency and it converges to a better Pareto front.

The BER necessities: the repair of DNA damage in human-adapted bacterial pathogens.

Science.gov (United States)

van der Veen, Stijn; Tang, Christoph M

2015-02-01

During colonization and disease, bacterial pathogens must survive the onslaught of the host immune system. A key component of the innate immune response is the generation of reactive oxygen and nitrogen species by phagocytic cells, which target and disrupt pathogen molecules, particularly DNA, and the base excision repair (BER) pathway is the most important mechanism for the repair of such oxidative DNA damage. In this Review, we discuss how the human-specific pathogens Mycobacterium tuberculosis, Helicobacter pylori and Neisseria meningitidis have evolved specialized mechanisms of DNA repair, particularly their BER pathways, compared with model organisms such as Escherichia coli. This specialization in DNA repair is likely to reflect the distinct niches occupied by these important human pathogens in the host.

The new 'BerSANS-PC' software for reduction and treatment of small angle neutron scattering data

International Nuclear Information System (INIS)

Keiderling, U.

2002-01-01

Measurements on small angle neutron scattering (SANS) instruments are typically characterized by a large number of samples, short measurement times for the individual samples, and a frequent change of visiting scientist groups. Besides this, recent advances in instrumentation have led to more frequent measurements of kinetic sequences and a growing interest in analyzing two-dimensional scattering data, these requiring special software tools that enable the users to extract physically relevant information from the scattering data with a minimum of effort. The new 'BerSANS-PC' data-processing software has been developed at the Hahn-Meitner-Institut (HMI) in Berlin, Germany, to meet these requirements and to support an efficiently working guest-user service. Comprising some basic functions of the 'BerSANS' program available at the HMI and other institutes in the past, BerSANS-PC is a completely new development for network-independent use on local PCs with a full-feature graphical interface. (orig.)

Exxon nuclear power distribution control for pressurized water reactors: Phase II

International Nuclear Information System (INIS)

Holm, J.S.; Burnside, R.J.

1978-01-01

The power distribution control procedure, denoted PDC-II, described in this report enables nuclear plants to manage core power distributions such that Technical Specification Limits on F/sub Q//sup T/ are not violated during normal operation and limits on MDNBR are not violated during steady-state, load-follow, and anticipated transients. The PDC-II data base described provides the means for predicting the maximum F/sub Q//sup T/(z) distribution anticipated during operation under the PDC-II procedure taking into account the incore measured equilibrium power distribution data for the reactor in question. A comparison of this distribution with the Technical Specification limit curve determines whether the Technical Specification limit can be protected by PDC-II procedure. If such protection can be confirmed for a given operating cycle interval, APDMS monitoring is not necessary over this interval and the excore monitored constant axial offset limits will protect the Technical Specification F/sub Q//sup T/ limits. This document describes the maximum possible variation in F/sub Q//sup T/(z) which can occur during operation when following the PDC-II procedures. This bounding variation in F/sub Q//sup T/(z) is referred to as V(z). This V(z) distribution represents the maximun variation in F/sub Q//sup T/(z) when the axial offset is maintained within the range defined in this report [+- 5% at full power condition

Neuronale Korrelate sozialer Interaktion beim Menschen und deren genetische Modulation unter besonderer Berücksichtigung des Spiegelneuronensystems

OpenAIRE

Cordes, Andrea

2013-01-01

Die in vielen Lebenssituationen über Erfolg und Misserfolg, Sympathie und Antipathie entscheidende, unterschiedlich ausgeprägte Fähigkeit zur sozialen Interaktion ist Gegenstand neurowissenschaftlicher Untersuchungen. Die Versuche einer kortikalen Lokalisierung führten zur Beschreibung des Spiegelneuronensystems (SNS) in Hirnregionen, die sowohl beim Ausführen einer Aktion Aktivität zeigten als auch beim Beobachten derselben beim Gegenüber. Die stärkste Aktivierung dieser Areale wurde bei sch...

Partial oxidation of Raffinate II and other mixtures of n-Butane and n-Butenes to maleic anhydride in a fixed-bed reactor

OpenAIRE

Brandstädter, Willi Michael

2008-01-01

The utilisation of the C4 streams of steamcrackers by converting raffinate II to maleic anhydride was studied. The oxidation reactions were investigated in a laboratory-scale fixed-bed reactor to determine reaction kinetics. The effects of pore diffusional resistance were investigated and explained. A two-dimensional pseudo-homogeneous reactor model was used for the simulation of a production-scale fixed-bed reactor. A flow scheme of the reactor section including a recycle was proposed.

MASS TRANSPORT PROPERTIES OF A FLOW-THROUGH ELECTROLYTIC REACTOR USING A POROUS ELECTRODE: PERFORMANCE AND FIGURES OF MERIT FOR Pb(II REMOVAL

Directory of Open Access Journals (Sweden)

Bertazzoli R.

1998-01-01

Full Text Available The removal of lead from an acid borate-nitrate solution containing Pb(II was used to characterize the mass transport properties of an electrolytic reactor with reticulated vitreous carbon cathodes, operated in the flow-through mode. Current potential curves recorded at a rotating vitreous carbon disc electrode were used to determine the diffusion coefficient for Pb(II under the conditions of the experiments. The performance and figures of merit of the electrolytic reactor were investigated by using different flowrates and cathode porosities. Dimensionless Sherwood and Reynolds numbers were correlated to characterize the mass transport properties of the reactor, and they were fitted to the equation Sh=24Re0.32Sc0.33.

Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

International Nuclear Information System (INIS)

1993-09-01

The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses

Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

2015-01-01

Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

Utilization of boron irradiation filters in reactor neutron activation via epithermal (n,γ) and fast neutron reactions

International Nuclear Information System (INIS)

Chisela, F.

1986-01-01

The technique of instrumental neutron activation analysis based on irradiation with reactor epithermal and fast neutrons has been described and evaluated. Important characteristics of boron neutron absorbers used to remove thermal neutrons from the reactor neutron spectrum have been examined and compared with those of cadmium. Three boron compound shields, have been designed and constructed at the BER II 5MW reactor for use in epithermal neutron activation analysis of biological materials. The major advantages offered by these filters in this application include the flexibility of varying the filter thickness, the low radioactivity induced in the filters during irradiation, ease of fabrication and the relatively low cost of the filter materials. The radiation heating due to the 10 B(n,α) 7 Li-reaction has been experimentally investigated for the filters used and the results obtained confirm the necessity for efficient cooling of these filters during irradiation. Three irradiation facilities have been characterized with respect to the neutron flux density and the flux spatial distribution. An experiment has been designed and carried out to compensate the flux inhomogeneity in two irradiation positions of the DBV facility caused by the reactor geometry. Several biological samples including well characterized reference materials have been analysed after epithermal activation and the results compared with those obtained with the classical thermal neutron activation method. Improved sensitivity of determination has been found for elements with high resonance integral to thermal neutron cross section ratios (RI/σ 0 ). The range of elements that can be determined instrumentally is extended and the time scale of analysis is considerably reduced. (orig.) [de

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Reactor building design of nuclear power plant ATUCHA II, Argentina

International Nuclear Information System (INIS)

Rufino, R.E.; Hermann, E.R.; Richter, E.

1984-01-01

It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-06-28

The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the ²³⁵U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

The control equipment of the Melusine II reactor; L'equipement de controle du reacteur Melusine II

Energy Technology Data Exchange (ETDEWEB)

Cordelle, M; Delcroix, V; Denis, P; Gariod, R

1963-07-01

Melusine II, low-power reactor, used for the study of Siloe core has diverged at the CEA Grenoble, the 23. May 1962; its monitoring board studied and carried out in this center is the first in France to be entirely transistorized. The first months of running have justified the hope put in the new electronics to improve the stability and the safety of running. The article describes the design of the control and gives the main characteristics of the measurement chains and of the actions on reactivity. (O.M.) [French] Melusine II, reacteur de faible puissance destine a l'etude du coeur de Siloe a diverge au Centre d'Etudes Nucleaires de Grenoble, le 23 mai 1962, son tableau de controle etudie et realise dans ce Centre est le premier en France a etre entierement transistorise. Les premiers mois de fonctionnement ont justifie l'espoir mis dans la nouvelle electronique pour ameliorer la stabilite et la surete de fonctionnement. L'article decrit la conception du controle et donne les principales caracteristiques des chaines de mesure et des actions sur la reactivite. (auteurs)

BER Performance Simulation of Generalized MC DS-CDMA System with Time-Limited Blackman Chip Waveform

Directory of Open Access Journals (Sweden)

I. Develi

2010-09-01

Full Text Available Multiple access interference encountered in multicarrier direct sequence-code division multiple access (MC DS-CDMA is the most important difficulty that depends mainly on the correlation properties of the spreading sequences as well as the shape of the chip waveforms employed. In this paper, bit error rate (BER performance of the generalized MC DS-CDMA system that employs time-limited Blackman chip waveform is presented for Nakagami-m fading channels. Simulation results show that the use of Blackman chip waveform can improve the BER performance of the generalized MC DS-CDMA system, as compared to the performances achieved by using timelimited chip waveforms in the literature.

The physics design of EBR-II

International Nuclear Information System (INIS)

Loewenstein, W.B.

1962-01-01

The physics design oi EBR-II. Calculations of the static, dynamic and long-term reactivity behaviour of EBR-II are reported together with results and analysis of EBR-II dry critical and ZPR-III mock-up experiments. Particular emphasis is given to reactor-physics design problems which arise after the conceptual design is established and before the reactor is built or placed into operation. Reactor-safety analyses and hazards-evaluation considerations are described with their influence on the reactor design. The manner of utilizing the EBR-II mock-up on ZPR-III data and the EBR-II dry critical data is described. These experiments, their analysis and theoretical predictions are the basis for predetermining the physics behaviour of the reactor system. The limitations inherent in applying the experimental data to the performance of the power-reactor system are explored in some detail. This includes the specification of reactor core size and/or fuel-alloy enrichment, provisions for adequate operating and shut-down reactivity, determination of operative temperature and power coefficients of reactivity, and details of power- and flux-distribution as a function of position within the reactor structure. The overall problem of transferring information from simple idealized analytical or experimental geometry to actual hexagonal reactor geometry is described. Nuclear performance, including breeding, of the actual reactor system is compared with that of the idealized conceptual system. The long-term reactivity and power behaviour of the reactor blanket is described within the framework of the proposed cycling of the fuel and blanket alloy. Safety considerations, including normal and abnormal rates of reactivity-insertion, the implication of postulated reactivity effects based on the physical behaviour of the fuel alloy and reactor structure as well as extrapolation of TREAT experiments to the EBR-II system are analysed. The EBR-II core melt-down problem is reviewed. (author

BER and optimal power allocation for amplify-and-forward relaying using pilot-aided maximum likelihood estimation

KAUST Repository

Wang, Kezhi

2014-10-01

Bit error rate (BER) and outage probability for amplify-and-forward (AF) relaying systems with two different channel estimation methods, disintegrated channel estimation and cascaded channel estimation, using pilot-aided maximum likelihood method in slowly fading Rayleigh channels are derived. Based on the BERs, the optimal values of pilot power under the total transmitting power constraints at the source and the optimal values of pilot power under the total transmitting power constraints at the relay are obtained, separately. Moreover, the optimal power allocation between the pilot power at the source, the pilot power at the relay, the data power at the source and the data power at the relay are obtained when their total transmitting power is fixed. Numerical results show that the derived BER expressions match with the simulation results. They also show that the proposed systems with optimal power allocation outperform the conventional systems without power allocation under the same other conditions. In some cases, the gain could be as large as several dB\\'s in effective signal-to-noise ratio.

BER and optimal power allocation for amplify-and-forward relaying using pilot-aided maximum likelihood estimation

KAUST Repository

Wang, Kezhi; Chen, Yunfei; Alouini, Mohamed-Slim; Xu, Feng

2014-01-01

Bit error rate (BER) and outage probability for amplify-and-forward (AF) relaying systems with two different channel estimation methods, disintegrated channel estimation and cascaded channel estimation, using pilot-aided maximum likelihood method in slowly fading Rayleigh channels are derived. Based on the BERs, the optimal values of pilot power under the total transmitting power constraints at the source and the optimal values of pilot power under the total transmitting power constraints at the relay are obtained, separately. Moreover, the optimal power allocation between the pilot power at the source, the pilot power at the relay, the data power at the source and the data power at the relay are obtained when their total transmitting power is fixed. Numerical results show that the derived BER expressions match with the simulation results. They also show that the proposed systems with optimal power allocation outperform the conventional systems without power allocation under the same other conditions. In some cases, the gain could be as large as several dB's in effective signal-to-noise ratio.

Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

International Nuclear Information System (INIS)

Mohd Rafi Mohd Solleh; Abdul Aziz Tajuddin; Abdul Aziz Mohamed; Eid Mahmoud Eid Abdel Munem; Mohamad Hairie Rabir; Julia Abdul Karim; Yoshiaki, Kiyanagi

2011-01-01

The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 10 8 n/ cm 2 / s. According to IAEA (2001) flux of 1.00 x 10 9 n/ cm 2 / s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

International Working Group on Fast Reactors Thirteenth Annual Meeting. Summary Report. Part II

International Nuclear Information System (INIS)

1980-10-01

The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

International Nuclear Information System (INIS)

DeVault, G.P.; Bell, C.R.

1985-01-01

Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

BER and FER Prediction of Control and Traffic Channels for a GSM type of interface

DEFF Research Database (Denmark)

Wigard, Jeroen; Nielsen, Thomas Toftegaard; Michaelsen, Per Henrik

1998-01-01

in a network simulator, but without having to simulate every single link, since this would be to time consuming. In this paper a method is presented to find the BER and FER from the signal to interference (C/I) values for a GSM type of air-interface, which can be used for integration of link aspects...... in a network simulator. The accuracy is within 0.2 dB in case of the BER and 0.5 for the FER. Both traffic and control channels are studied and the method is independent of hopping sequences and speed...

Outage and BER analysis for ultrawideband-based WPAN in Nakagami-m fading channels

KAUST Repository

Mehbodniya, Abolfazl; Aissa, Sonia

2011-01-01

layer. Approximate expressions for the outage probability and average bit error rate (BER) are derived in closed form for the MB-OFDM target receiver, taking into account multi-user interference (MUI), as well as external interference in the form of time

Berührt sein – Ein Fallbericht zur Rolle von WATSU (WasserShiatsu) in der Rehabilitation nach Polytrauma

OpenAIRE

Schitter, Agnes Maria; Fleckenstein, Johannes

2017-01-01

Hintergrund: Die herausragende Bedeutung von Körperkontakt für den Menschen ist bekannt und nachgewiesen. Während aber intensiver Körperkontakt in der medizinischen Versorgung von Frühgeborenen etabliert ist, wird Berührung in schulmedizinische Therapiekonzepte für Erwachsene bisher nur marginal integriert. Massnahmen aus der Komplementärmedizin berücksichtigen den individuellen Bedarf an körperlicher Nähe und leisten einen klinischen Mehrwert für den Gesundungsprozess. In besonderer Weise...

Untersuchungen zum Einfluss einer Air-Liquid-Interphase auf die Differenzierung von Alveolarepithelzellen Typ II

OpenAIRE

Sinis, Nektarios

2003-01-01

Die Dedifferenzierung der Alveolarepithelzellen Typ II stellt in vitro ein großes Problem dar. Zellen, die länger als 48 Stunden in vitro kultiviert werden tendieren zu einer Dedifferenzierung. Die vorliegende Arbeit stellt ein Air-Liquid-Interphasesystem zur Kultivierung von TypII-Zellen vor. Hierdurch ist es uns gelungen TypII-Zellen über einen Zeitraum von bis zu 14 Tagen in vitro auf einem hohen Grad der Differenzierung zu konservieren. Zellen die nicht so kultiviert wurden zeigten fr...

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

Energy Technology Data Exchange (ETDEWEB)

Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity

Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

International Nuclear Information System (INIS)

Kang, S. H.; Ha, K. S.

2013-01-01

The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity feedback (Doppler and

Richard Wollheim über die Metapher in der Malerei

OpenAIRE

Heinrich, Richard

1993-01-01

Es handelt sich um den Text eines Vortrages, den ich Jänner 1993 im Rahmen einer Vortragsreihe am Institut für Kunstgeschichte der Universität Wien gehalten habe. Er ist bisher unpubliziert und nicht redigiert. Das worüber ich spreche ist ein Artikel von Richard Wollheim mit dem Titel "Die Metapher in der Malerei". Diesen Artikel habe ich herausgegeben, gemeinsam mit meinem Kollegen Helmuth Vetter, 1991 in dem Band "Bilder der Philosophie". Ursprünglich handelt es sich um einen Vortrag, d...

Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

International Nuclear Information System (INIS)

Ward, D.R.; Holland, L.B.

1979-09-01

The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Scientific results - report 1981

International Nuclear Information System (INIS)

1982-01-01

During the year under report 1981, five working groups were active in the field of nuclear chemistry: neutron diffraction, radiation damage in solids, reactor chemistry, trace element research in bio-medicine and geo-chemistry. The objectives of the R+D projects ranged from the more basic research to the development of technological processes. Nuclear inspection methodes that have already been developed (e.g. neutron diffraction, trace element analysis) have increasingly been used in an interdisciplinary way. Besides these R+D projects the project of increase of power of the BER II was pursued also in 1981, and further planning documents on the extension of the BER II have been established. The report informs about the most important results of the single sections. A list of the publications (with abstracts) and lectures, also by guest scientists, is attached. (RB) [de

Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

International Nuclear Information System (INIS)

Salam, M. A.

2013-01-01

The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

Energy Technology Data Exchange (ETDEWEB)

Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

2013-07-01

The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

Monitoring of the spatial and temporal dynamics of BER/SSBR pathway proteins, including MYH, UNG2, MPG, NTH1 and NEIL1-3, during DNA replication.

Science.gov (United States)

Bj Rås, Karine Ø; Sousa, Mirta M L; Sharma, Animesh; Fonseca, Davi M; S Gaard, Caroline K; Bj Rås, Magnar; Otterlei, Marit

2017-08-21

Base lesions in DNA can stall the replication machinery or induce mutations if bypassed. Consequently, lesions must be repaired before replication or in a post-replicative process to maintain genomic stability. Base excision repair (BER) is the main pathway for repair of base lesions and is known to be associated with DNA replication, but how BER is organized during replication is unclear. Here we coupled the iPOND (isolation of proteins on nascent DNA) technique with targeted mass-spectrometry analysis, which enabled us to detect all proteins required for BER on nascent DNA and to monitor their spatiotemporal orchestration at replication forks. We demonstrate that XRCC1 and other BER/single-strand break repair (SSBR) proteins are enriched in replisomes in unstressed cells, supporting a cellular capacity of post-replicative BER/SSBR. Importantly, we identify for the first time the DNA glycosylases MYH, UNG2, MPG, NTH1, NEIL1, 2 and 3 on nascent DNA. Our findings suggest that a broad spectrum of DNA base lesions are recognized and repaired by BER in a post-replicative process. © The Author(s) 2017. Published by Oxford University Press on behalf of Nucleic Acids Research.

Rehabilitierungs-Szenarien, Geschichten über Hexen, Verdingkinder, Flüchtlingshelfer usw.

OpenAIRE

Rüegger, Vanessa

2015-01-01

In den vergangenen Jahren haben Parlamente, Regierungen und Gerichte unterschiedliche Personen und Personengruppen rehabilitiert. Der vorliegende Beitrag bietet einen Überblick über die jüngst in der Schweiz ausgesprochenen Rehabilitierungen, ordnet diese rechtsdogmatisch ein und zeigt weiter die Funktionen und Grenzen der Rehabilitierung als Form staatlicher ,Vergangenheitsbewältigung‘ auf. Der Beitrag schliesst mit der Erkenntnis, dass Flüchtlingshelfer, Spanienkämpfer, administrativ Versor...

The sensitivity of bit error rate (BER) performance in multi-carrier (OFDM) and single-carrier

Science.gov (United States)

Albdran, Saleh; Alshammari, Ahmed; Matin, Mohammad

2012-10-01

Recently, the single-carrier and multi-carrier transmissions have grabbed the attention of industrial systems. Theoretically, OFDM as a Multicarrier has more advantages over the Single-Carrier especially for high data rate. In this paper we will show which one of the two techniques outperforms the other. We will study and compare the performance of BER for both techniques for a given channel. As a function of signal to noise ratio SNR, the BER will be measure and studied. Also, Peak-to-Average Power Ratio (PAPR) is going to be examined and presented as a drawback of using OFDM. To make a reasonable comparison between the both techniques, we will use additive white Gaussian noise (AWGN) as a communication channel.

Neutron radiography applications in I.T.U. TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Tugrul, A. B.

2002-01-01

Neutron radiography is an important radiographic technique which is supplied different and advanced information according to the X or gamma ray radiography. However, it has a trouble for supplying the convenient neutron sources. Tangential beam tube of Istanbul Technical University (ITU) TRIGA Mark-II Training and Research Reactor has been arranged for using neutron radiography. The neutron radiography set defined as detailed for the application of the technique. Two different techniques for neutron radiography are defined as namely, transfer method and direct method. For the transfer method dysprosium and indium screens are used in the study. But, dysprosium generally was preferred in many studies in the point of view nuclear safety. Gadolinium was used for direct method. Two techniques are compared and explained the preferring of the transfer technique. Firstly, reference composition is prepared for seeing the differences between neutron and X-ray or gamma radiography. In addition of it, some radiograph samples are given neutron and X-ray radiography which shows the different image characters. Lastly, some examples are given from archaeometric studies. One of them the brass plates of Great Mosque door in Cizre. After the neutron radiography application, organic dye traces are noticed. Other study is on a sword that belong to Urartu period at the first millennium B.C. It is seen that some wooden part on it. Some different artefacts are examined with neutron radiography from the Ikiztepe excavation site, then some animal post parts are recognized on them. One of them is sword and sheath which are corroded together. After the neutron radiography application, it can be noticed that there are a cloth between the sword and its sheath. By using neutron radiography, many interesting and detailed results are observed in ITU TRIGA Mark-II Training and Research Reactor. Some of them shouldn't be recognised by using any other technique

Categorization of ber varieties in relation to blooming period, fruit setting and harvesting time

International Nuclear Information System (INIS)

Sharif, N.; Abbas, M.M.; Ishfaq, M.; Memon, N.U.N.

2013-01-01

Thirty four local Ber varieties were evaluated at Horticultural Research Institute AARI, Faisalabad, Horticultural Research Station Bahawalpur (Punjab) and Jujube Research Station, Tandojam (Sindh). Traits viz. total period of blooming (dates), peak period of blooming (dates), total period of fruit set (dates), peak period of fruit set (dates), total period of fruit harvest (dates), peak period of fruit harvest (dates), total flowering days, peak flowering days, total fruit setting days, peak fruit setting days, total harvesting days and peak harvesting days were studied. The results revealed significant differences in parameters studied except total period of blooming under Tandojam, Sindh conditions. Varieties were classified as early, mid and late season for both provinces. Local varieties had potential for further manipulation in terms of variety improvement to attract growers for extensive ber cultivations under changing global climatic scenario. (author)

Experience in operation and maintenance of the TRIGA Mark II reactor at the University of Pavia in the time period July 1974 - June 1976

International Nuclear Information System (INIS)

Cambieri, A.; Cingoli, F.; Genova, N.; Meloni, S.; Perlini, G.

1976-01-01

The operation of the 250 kW steady state/250 MW pulsed TRIGA Mark II Reactor of the University of Pavia over the past two years is presented and discussed. Reactor maintenance activity is presented as well. Data for reactor utilization and a summary of the health physics procedures are also given. Since the third European Conference of TRIGA Reactor Users in 1974, reactor operation took place smoothly without major troubles. Because of the core excess decrease due to reactor operation and consequent burn-up, ten new stainless steel clad fuel elements were bought from General Atomic. Reactor operation license expired at the end of 1975 and it is now under way the bureaucratic work to get its renewal. The aging of the electronic equipment raises minor problems and the predicted switching to a new solid state equipment started by changing the old electromechanical rod position indicators with new digital ones. The installation of the Euracos II facility (Enriched Uranium Converter Source), described at the last TRIGA Users Conference, began at the end of 1975 and it is still under way. The first operation of the facility will take place at reduced power so that the removal of the graphite blocks from thermal column was not accomplished. The installation of the facility is described and the procedures of its operation in connection with reactor operation are presented as well. (author)

On the BER and capacity analysis of MIMO MRC systems with channel estimation error

KAUST Repository

Yang, Liang; Alouini, Mohamed-Slim

2011-01-01

In this paper, we investigate the effect of channel estimation error on the capacity and bit-error rate (BER) of a multiple-input multiple-output (MIMO) transmit maximal ratio transmission (MRT) and receive maximal ratio combining (MRC) systems over

BER and total throughput of asynchronous DS-OCDMA/WDM systems with multiple user interference

OpenAIRE

Ghiringhelli, F.; Zervas, M.N.

2003-01-01

The BER and throughput of Direct-Sequence OCDMA/WDM systems based on quadripolar codes and superstructured fiber Bragg gratings are statistically derived under asynchronous operation, intensity detection, and Multiple User Interference. Performance improvements with Forward Error Correction are included.

A report on the second specialists' meeting on reactor noise (SMORN II) Gatlinburg, Tennessee, September 19th-23rd 1977

International Nuclear Information System (INIS)

Greef, C.P.

1978-03-01

The current state of power reactor noise analysis as presented at the Second Specialists' Meeting on Reactor Noise (SMORN II) is reviewed. Practical successes include the analysis of PWR core support barrel motions and BWR instrument tube vibrations, leading to better understanding of the Physical processes involved. Future applications to LMFBRs are being actively pursued, but more information is required, notably on background noise sources in operating reactors. Theoretical understanding is improving in a number of specific areas, but general fundamental progress is hampered by the multiplicity of potential noise sources in a reactor. Past attempts to automate noise analysis so that simplified information may be presented to the reactor operator have not been totally successful. This has resulted from an over-hasty application of the techniques before all of the factors involved in physical phenomena were fully appreciated. (author)

Reentry safety for the Topaz II Space Reactor: Issues and analyses

International Nuclear Information System (INIS)

Connell, L.W.; Trost, L.C.

1994-03-01

This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

Chemical surveillance of commercial fast breeder reactors

International Nuclear Information System (INIS)

Stamm, H.H.; Stade, K.Ch.

1988-01-01

After BN-600 (USSR) and SUPERPHENIX (France) were started succesfully, the international development of LMFBRs is standing at the doorstep of commercial use. For commercial use of LMFBRs cost reductions for construction and operation are highly desirable and necessary. Several nations developing breeder reactors have joined in a common effort in order to reach this aim by standardization and harmonization. On the base of more than 20 years of operation experience of experimental reactors (EBR-II, FFTF, RAPSODIE, DFR, BR-5/BR-10, BOR-60, JOYO, KNK-II) and demonstration plants (PHENIX, PFR, BN-350), possibilities for standardization in chemical surveillance of commercial breeder reactors without any loss of availability, reliability and reactor safety will be discussed in the following chapters. Loop-type reactors will be considered as well as pool-type reactors, although all commercial plants under consideration so far (SUPERPHENIX II, BN-800, BN-1600, CFBR, SNR-2, EFR) include pool-type reactors only. Table 1 gives a comparison of the Na inventories of test reactors, prototype plants and commercial LMFBRs

The new area monitoring system and the fuel database of the TRIGA Mark II reactor in Vienna

International Nuclear Information System (INIS)

Villa, M.; Boeck, H.; Hofbauer, M.; Schwarz, V.

2004-01-01

The 250 kW TRIGA Mark-II reactor operates since March 1962 at the Atominstitut, Vienna, Austria. Its main tasks are nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection and dosimetry, and low temperature physics and fusion research. Academic research is carried out by students in the above mentioned fields coordinated and supervised by about 70 staff members with the aim of a masters- or PhD degree in one of the above mentioned areas. After 25 years of successful operation, it was necessary to exchange the old area monitoring system with a new digital one. The purpose of the new system is the permanent control of the reactor hall, the primary and secondary cooling system and the monitoring of the ventilation system. The paper describes the development and implementation of the new area monitoring system. The second topic in this paper describes the development of the new fuel database. Since March 7th, 1962, the TRIGA Mark II reactor Vienna operates with an average of 263 MWh per year, which corresponds to a uranium burn-up of 13.7 g per year. Presently we have 81 TRIGA fuel elements in the core, 55 of them are old aluminium clad elements from the initial criticality while the rest are stainless steel clad elements which had been added later to compensate the uranium consumption. Because 67 % of the elements are older than 40 years, it was necessary to put the history of every element in a database, to get an easy access to all the relevant data for every element in our facility. (author)

PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Masood, Z.

2016-01-01

The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

Archaeometric studies by using neutron radiography in ITU TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Tugrul, A. Beril

2008-01-01

Archaeometric many studies have been done by using neutron radiography in ITU TRIGA Mark-II Training and Research Reactor for over 15 years. Tangential beam tube has been arranged for using neutron radiography. Generally, transfer technique has been preferred with using dysprosium screen, but indium screen also is used. Some studies are described which are all on the Anatolian artefacts. The first study from 13th century AD deals with Seljukian period from south-east Anatolia. It investigated a plate from Great Mosque door in Cizre. With means of the neutron radiography painting traces are investigated on the plates. Organic dye traces are noticed on some of plates, which have generally animal figures. Other studies from Urartu period at the first millennium B.C, investigates artefacts found at the vicinity of Van on east Anatolia. An important one is a sword that was found in a grave. It has some corrosion defects. The neutron radiography was applied and shown that wooden parts are there. Other studies referred to samples from the Ikiztepe Excavation site on north Anatolia. Many artefacts were examined by neutron radiography. Some of them evidenced animal parts are recognised as covering parts. An interesting result was obtained to a sword and its sheath that were corroded together. After the neutron radiography applications, it was noticed that there are a cloth between the sword and its sheath. Hence, it was the cause of corrosion of the artefact. By using neutron radiography, many interesting and detailed results were observed by means of the neutron beam from the ITU TRIGA Mark-II Training and Research Reactor. Some of them could not be evidenced by means of any other technique

EBR-II Data Digitization

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su-Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sackett, John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-08-01

1. Objectives To produce a validation database out of those recorded signals it will be necessary also to identify the documents need to reconstruct the status of reactor at the time of the beginning of the recordings. This should comprehends the core loading specification (assemblies type and location and burn-up) along with this data the assemblies drawings and the core drawings will be identified. The first task of the project will be identify the location of the sensors, with respect the reactor plant layout, and the physical quantities recorded by the Experimental Breeder Reactor-II (EBR-II) data acquisition system. This first task will allow guiding and prioritizing the selection of drawings needed to numerically reproduce those signals. 1.1 Scopes and Deliverables The deliverables of this project are the list of sensors in EBR-II system, the identification of storing location of those sensors, identification of a core isotopic composition at the moment of the start of system recording. Information of the sensors in EBR-II reactor system was summarized from the EBR-II system design descriptions listed in Section 1.2.

Research reactor`s role in Korea

Energy Technology Data Exchange (ETDEWEB)

Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1996-12-31

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.

Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

Energy Technology Data Exchange (ETDEWEB)

Steven R. Sherman

2005-04-01

Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium

METHUSELAH II - A Fortran program and nuclear data library for the physics assessment of liquid-moderated reactors

International Nuclear Information System (INIS)

Brinkworth, M.J.; Griffiths, J.A.

1966-03-01

METHUSELAH II is a Fortran program with a nuclear data library, used to calculate cell reactivity and burn-up in liquid-moderated reactors. It has been developed from METHUSELAH I by revising the nuclear data library, and by introducing into the program improvements relating to nuclear data, improvements in efficiency and accuracy, and additional facilities which include a neutron balance edit, specialised outputs, fuel cycling, and fuel costing. These developments are described and information is given on the coding and usage of versions of METHUSELAH II for the IBM 7030 (STRETCH), IBM 7090, and KDF9 computers. (author)

Safety analysis and optimization of the core fuel reloading for the Moroccan TRIGA Mark-II reactor

International Nuclear Information System (INIS)

Nacir, B.; Boulaich, Y.; Chakir, E.; El Bardouni, T.; El Bakkari, B.; El Younoussi, C.

2014-01-01

Highlights: • Additional fresh fuel elements must be added to the reactor core. • TRIGA reactor could safely operate around 2 MW power with 12% fuel elements. • Thermal–hydraulic parameters were calculated and the safety margins are respected. • The 12% fuel elements will have no influence on the safety of the reactor. - Abstract: The Moroccan TRIGA MARK II reactor core is loaded with 8.5% in weight of uranium standard fuel elements. Additional fresh fuel elements must periodically be added to the core in order to remedy the observed low power and to return to the initial reactivity excess at the End Of Cycle. 12%-uranium fuel elements are available to relatively improve the short fuel lifetime associated with standard TRIGA elements. These elements have the same dimensions as standards elements, but with different uranium weight. The objective in this study is to demonstrate that the Moroccan TRIGA reactor could safely operate, around 2 MW power, with new configurations containing these 12% fuel elements. For this purpose, different safety related thermal–hydraulic parameters have been calculated in order to ensure that the safety margins are largely respected. Therefore, the PARET model for this TRIGA reactor that was previously developed and combined with the MCNP transport code in order to calculate the 3-D temperature distribution in the core and all the most important parameters like the axial distribution of DNBR (Departure from Nucleate Boiling Ratio) across the hottest channel. The most important conclusion is that the 12% fuel elements utilization will have no influence on the safety of the reactor while working around 2 MW power especially for configurations based on insertions in C and D-rings

A horizontal plug-flow baffled bioelectrocatalyzed reactor for the reductive decolorization of Alizarin Yellow R.

Science.gov (United States)

Sun, Qian; Li, Zhiling; Wang, Youzhao; Cui, Dan; Liang, Bin; Thangavel, Sangeetha; Chung, Jong Shik; Wang, Aijie

2015-11-01

An application-oriented membrane-free, continuous plug-flow baffled bioelectrocatalyzed reactor (PFB-BER), was designed and testified for the decolorization of Alizarin Yellow R. Decolorization efficiency (DE) with an external power source of 0.5 V was higher than without electrolysis, i.e. 93.4% versus 73.6% (HRT of 24 h). Product formation efficiencies of p-phenylenediamine and 5-aminosalicylic acid were above 95% and 50%, respectively. When HRT decreased to 8 h and 4 h, DE reduced to 69.9% and 44.9%, respectively. An additional electrode assembly improved DE to 96.4% (HRT of 8 h) and 80% (HRT of 4 h), while energy consumption (HRT of 4 h) was lower than that of HRT of 12 h with single electrode assembly under comparable DE. The PFB-BER with higher removal capacity, lower internal resistance and energy consumption provides a new solution to treat the high loading azo dye-containing wastewaters. Copyright © 2015. Published by Elsevier Ltd.

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Facilities design for TIBER II

International Nuclear Information System (INIS)

Thomson, S.L.; Blevins, J.D.

1987-01-01

This paper describes the conceptual design of the reactor building and reactor maintenance building for the TIBER II tokamak. These buildings are strongly influenced by the reactor configuration, and their characterization allows a better understanding of the economic and technical implications of the reactor design. Key features of TIBER II that affect the facilities design are the small size and compact arrangement, the use of an external vacuum vessel, and the complete reliance on remote maintenance. The building design incorporates requirements for equipment layout, maintenance operations and equipment, safety, and contamination control. 4 figs

Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

Science.gov (United States)

Henry, R; Tiselj, I; Snoj, L

2015-03-01

New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. Copyright © 2014 Elsevier Ltd. All rights reserved.

Briti teadlane: kübersõda ei tule / Thomas Rid ; intervjueerinud Raimo Poom

Index Scriptorium Estoniae

Rid, Thomas

2013-01-01

Sõjauuringute õppejõud Thomas Rid leiab, et internetis on palju ohte, kuid kübersõda ei ole üks neist. Ridi arvates on äärmiselt ebatõenäoline, et NATO käivitaks kunagi küberrünnaku tõttu 5. artikli. Interneti kõige suuremaks ohustajaks pole NSA, vaid vabadust piirata soovivad autoritaarsed riigid

Analysis of a bio-electrochemical reactor containing carbon fiber textiles for the anaerobic digestion of tomato plant residues.

Science.gov (United States)

Hirano, Shin-Ichi; Matsumoto, Norio

2018-02-01

A bio-electrochemical system packed with supporting material can promote anaerobic digestion for several types of organic waste. To expand the target organic matters of a BES, tomato plant residues (TPRs), generated year-round as agricultural and cellulosic waste, were treated using three methanogenic reactors: a continuous stirred tank reactor (CSTR), a carbon fiber textile (CFT) reactor, and a bio-electrochemical reactor (BER) including CFT with electrochemical regulation (BER + CFT). CFT had positive effects on methane fermentation and methanogen abundance. The microbial population stimulated by electrochemical regulation, including hydrogenotrophic methanogens, cellulose-degrading bacteria, and acetate-degrading bacteria, suppressed acetate accumulation, as evidenced by the low acetate concentration in the suspended fraction in the BER + CFT. These results indicated that the microbial community in the BER + CFT facilitated the efficient decomposition of TPR and its intermediates such as acetate to methane. Copyright © 2017 Elsevier Ltd. All rights reserved.

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

Science.gov (United States)

Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

2017-08-01

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

The combined removal of methyl mercaptan and hydrogen sulfide via an electro-reactor process using a low concentration of continuously regenerable Ag(II) active catalyst

International Nuclear Information System (INIS)

Muthuraman, Govindan; Chung, Sang Joon; Moon, Il Shik

2011-01-01

Highlights: → Simultaneous removal of H 2 S and CH 3 SH was achieved at electro-reactor. → Active catalyst Ag(II) perpetually regenerated in HNO 3 medium by electrochemical cell. → CH 3 SH destruction follows two reaction pathways. → H 2 S induced destruction of CH 3 SH has identified. → Low concentration of active Ag(II) (12.5 x 10 -4 mol L -1 ) is enough for complete destruction. - Abstract: In this study, an electrocatalytic wet scrubbing process was developed for the simultaneous removal of synthetic odorous gases namely, methyl mercaptan (CH 3 SH) and hydrogen sulfide (H 2 S). The initial process consists of the absorption of CH 3 SH and H 2 S gases by an absorbing solution, followed by their mediated electrochemical oxidation using a low concentration of active Ag(II) in 6 M HNO 3 . Experiments were conducted under different reaction conditions, such as CH 3 SH and H 2 S loadings, active Ag(II) concentrations and molar flow rates. The cyclic voltammetry for the oxidation of CH 3 SH corroborated the electro-reactor results, in that the silver in the 6 M HNO 3 reaction solution significantly influences the oxidation of CH 3 SH. At a low active Ag(II) concentration of 0.0012 M, the CH 3 SH removal experiments demonstrated that the CH 3 SH degradation was steady, with 100% removal at a CH 3 SH loading of 5 g m -3 h -1 . The electro-reactor and cyclic voltammetry results indicated that the removal of H 2 S (100%) follows a mediated electrocatalytic oxidation reaction. The simultaneous removal of 100% of the CH 3 SH and H 2 S was achieved, even with a very low active Ag(II) concentration (0.0012 M), as a result of the high efficiency of the Ag(II). The parallel cyclic voltammetry results demonstrated that a process of simultaneous destruction of both CH 3 SH and H 2 S follows an H 2 S influenced mediated electrocatalytic oxidation. The use of a very low concentration of the Ag(II) mediator during the electro-reactor process is promising for the complete

Reactor protection system. Revision 1

International Nuclear Information System (INIS)

Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

1975-04-01

The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

Data Quality Objective Summary Report for Phase II of the 105-F and DR Reactor Buildings

International Nuclear Information System (INIS)

Bauer, R.G.

1998-01-01

This data quality objective (DQO) process is to support planning and decision-making activities of Phase II decontamination and decommissioning (D and D) activities for the 105-F and 105-DR Reactor Buildings.The objective of this DQO is to determine the survey and characterization requirements for these rooms to provide the necessary information for worker safety, waste designation, recycle, reuse, and clean landfill disposal decisions during D and D

Average BER analysis of SCM-based free-space optical systems by considering the effect of IM3 with OSSB signals under turbulence channels.

Science.gov (United States)

Lim, Wansu; Cho, Tae-Sik; Yun, Changho; Kim, Kiseon

2009-11-09

In this paper, we derive the average bit error rate (BER) of subcarrier multiplexing (SCM)-based free space optics (FSO) systems using a dual-drive Mach-Zehnder modulator (DD-MZM) for optical single-sideband (OSSB) signals under atmospheric turbulence channels. In particular, we consider the third-order intermodulation (IM3), a significant performance degradation factor, in the case of high input signal power systems. The derived average BER, as a function of the input signal power and the scintillation index, is employed to determine the optimum number of SCM users upon the designing FSO systems. For instance, when the user number doubles, the input signal power decreases by almost 2 dBm under the log-normal and exponential turbulence channels at a given average BER.

Release procedure according to paragraph 29 StrlSchv on example of the nuclear research reactor TRIGA Heidelberg II; Durchfuehrung von Freigabeverfahren nach paragraph 29 am Beispiel des TRIGA Heidelberg II

Energy Technology Data Exchange (ETDEWEB)

Cremer, J. [Siempelkamp Nukleartechnik GmbH (SNT) (Germany); Sold, A. [Deutsches Krebsforschungszentrum Heidelberg (DKFZ) (Germany)

2005-07-01

The aim of this lecture is to show the schedule of a release procedure according to paragraph 29 StrlSchV on the example of the decommissioning of the nuclear research reactor TRIGA Heidelberg II. It is shown on the effort done by the radiation protection representative of this plant. Considering this example, starting with planning, application, survey and execution, the complex context of the release procedure is becomes apparent. Thereby the new applied measuring techniques that require a certain practice and the responsibility of the radiation protection representative in the radiation protection law play a relevant role. In such small facilities as the TRIGA Heidelberg II, the radiation protection staff are employed according to the plant's size and work is focussed on radiation protection research and laboratories. The decommissioning process with its wide range of radiation protection requirements represents new challenges which have to be coordinated with the present duties of the radiation protection representative. The supervision and the responsibility for the release procedure according to paragraph 29 are the largest and the most sensitive part of decommissioning of the nuclear research reactor TRIGA Heidelberg II. (orig.)

Natural biostimulants reduce the incidence of BER in sweet yellow pepper plants (Capsicum annuum L.

Directory of Open Access Journals (Sweden)

Nada Parađiković

2013-06-01

Full Text Available Modern greenhouse pepper production should evolve towards more sustainable systems. The growing technique which combines soilless culture and biostimulants may reduce nutrient and water use with beneficial impact on the environment. Therefore, this work aimed to investigate effects of biostimulants application on hydroponically produced pepper plants (Capsicum annuum L. in conditions of reduced fertilization. Positive influence of biostimulant treatment on yield parameters was observed along with significant decrease in incidence of blossom-end rot (BER in two pepper cultivars. Biostimulants application resulted in overall increase in macro- and microelement content in fruits of treated pepper cultivars. Generally, biostimulants improved the yield of pepper plants grown hydroponically by increasing the nutrient uptake and decreasing the occurrence of BER. Thus, the application of biostimulants could be considered as a good production strategy for obtaining high yields of nutritionally valuable vegetables with lower impact on the environment.

EBR-II: summary of operating experience

International Nuclear Information System (INIS)

Perry, W.H.; Leman, J.D.; Lentz, G.L.; Longua, K.J.; Olson, W.H.; Shields, J.A.; Wolz, G.C.

1978-01-01

Experimental Breeder Reactor II (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. The primary cooling system is a submerged-pool type. The early operation of the reactor successfully demonstrated the feasibility of a sodium-cooled fast breeder reactor operating as an integrated reactor, power plant, and fuel-processing facility. In 1967, the role of EBR-II was reoriented from a demonstration plant to an irradiation facility. Many changes have been made and are continuing to be made to increase the usefulness of EBR-II for irradiation and safety tests. A review of EBR-II's operating history reveals a plant that has demonstrated high availability, stable and safe operating characteristics, and excellent performance of sodium components. Levels of radiation exposure to the operating and maintenance workers have been low; and fission-gas releases to the atmosphere have been minimal. Driver-fuel performance has been excellent. The repairability of radioactive sodium components has been successfully demonstrated a number of times. Recent highlights include installation and successful operation of (1) the hydrogen-meter leak detectors for the steam generators, (2) the cover-gas-cleanup system and (3) the cesium trap in the primary sodium. Irradiations now being conducted in EBR-II include the run-beyond-cladding breach fuel tests for mixed-oxide and carbide elements. Studies are in progress to determine EBR-II's capability for conducting important ''operational safety'' tests. These tests would extend the need and usefulness of EBR-II into the 1980's

Using TRIGA Mark II research reactor for irradiation with thermal neutrons

Energy Technology Data Exchange (ETDEWEB)

Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

2015-03-15

Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

HYLIFE-II reactor chamber mechanical design

International Nuclear Information System (INIS)

House, P.A.

1992-01-01

Mechanical design features of the reactor chamber for the HYLIFE-11 inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams are used for shielding and blast protection. The system is designed for an 8 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (20 m/s) salt streams and also recover up to half of the dynamic head

A Qualidade Observada em Salas de Berçário

OpenAIRE

Tadeu, Bárbara; Aguiar, Cecília

2017-01-01

RESUMO Este estudo teve como objetivo investigar a qualidade de estruturas formais de educação para crianças com menos de 12 meses de idade. Participaram 30 salas de berçário de Setúbal e Palmela, no sul de Portugal. A qualidade global das salas foi avaliada com base na Infant/Toddler Environment Rating Scale - Revised Edition e as práticas de socialização, com base na Caregiver Interaction Scale. Os resultados revelam qualidade global média moderada das salas e elevada qualidade nas práticas...

RA Research nuclear reactor, Part II: radiation protection at the RA reactor in 1987

International Nuclear Information System (INIS)

Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

1987-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

The KNK II instrumentation for global and local supervision of the reactor core

International Nuclear Information System (INIS)

Steiger, W.O.

1990-01-01

After an introduction into the KNK plant itself, their historical development and their present situation, the instrumentation of the global and local supervision of the KNK II-core as well as the main safety-related i- and c-systems are described. Special emphasis is laid on the instrumentation of the reactor protection systems and the shutdown systems. After that some practices are reported about instrumentation behavior and lessons learned from the operation and maintenance of the above mentioned systems. At last follows a short description of the special instrumentation for the detection of failed fuel subassemblies and of the plant data processing system. (orig.)

The KNK II instrumentation for global and local supervision of the reactor core

International Nuclear Information System (INIS)

Steiger, W.O.

1991-01-01

After an introduction into the KNK plant itself, their historical development and their present situation, the instrumentation of the global and local supervision of the KNK II-core as well as the main safety-related instrumentation and control systems is described. Special emphasis is laid on the instrumentation of the reactor protection systems and the shut down systems. After that some practices are reported about instrumentation behavior and lessons learned from the operation and maintenance of the above mentioned systems. At last follows a short description of the special instrumentation for the detection of failed fuel subassemblies and of the plant data processing system. (author). 4 refs, 18 tabs

Evaluation of the computerized procedures Manual II (COPMA II)

International Nuclear Information System (INIS)

Converse, S.A.

1995-11-01

The purpose of this study was to evaluate the effects of a computerized procedure system, the Computerized Procedure Manual II (COPMA-II), on the performance and mental workload of licensed reactor operators. To evaluate COPMA-II, eight teams of two operators were trained to operate a scaled pressurized water reactor facility (SPWRF) with traditional paper procedures and with COPMA-II. Following training, each team operated the SPWRF under normal operating conditions with both paper procedures and COPMA-II. The teams then performed one of two accident scenarios with paper procedures, but performed the remaining accident scenario with COPMA-II. Performance measures and subjective estimates of mental workload were recorded for each performance trial. The most important finding of the study was that the operators committed only half as many errors during the accident scenarios with COPMA-II as they committed with paper procedures. However, time to initiate a procedure was fastest for paper procedures for accident scenario trials. For performance under normal operating conditions, there was no difference in time to initiate or to complete a procedure, or in the number of errors committed with paper procedures and with COPMA-II. There were no consistent differences in the mental workload ratings operators recorded for trials with paper procedures and COPMA-II

Social Bots und Fake News - Was EU- und UNESCO-Referenzrahmen zum Lernen im digitalen Zeitalter darüber (nicht sagen

Directory of Open Access Journals (Sweden)

Dietmar Janetzko

2017-07-01

Full Text Available In den letzten Jahren haben internationale Organisationen wie die EU und die UNESCO eine Reihe von Vorschlägen und Strategiepapieren zur Bildung und Ausbildung im Zusammenhang mit digitalen Medien entwickelt. Mit den dabei entstandenen Rahmenkonzepten der EU (Digital Competence, DigComp sowie der UNESCO (Media and Information Literacy, MIL werden im Kern zwei zusammenhängende Ziele verfolgt: (i digitale Bildung bzw. digitale Kompetenzen, Fähigkeiten und zugehörige Einstellungen umfassend zu kartographieren sowie (ii über die dabei konzipierten edukativ-politischen Rahmenkonzepte Projektförderungen, Bildungs- bzw. Ausbildungsinitiativen sowie Gesetzesvorlagen anzustossen. Tatsächlich sind DigComp und MIL bereits dabei, auf internationaler Ebene einen prägenden Einfluss zu nahezu allen Fragen der Bildung und Ausbildung im Bereich digitaler Medien auszuüben. Beide Initiativen haben innerhalb der genannten Organisationen Leuchtturmcharakter, werden bislang aber von der allgemeinen Öffentlichkeit und der medienpädagogischen Fachöffentlichkeit kaum wahrgenommen. Dessen ungeachtet verbindet sich mit DigComp und MIL jeweils ein impliziter Anspruch auf einen – im Bedarfsfall zu aktualisierenden – Gesamtentwurf zur Analyse und Gestaltung medienpädagogischer Bildung und Ausbildung. Dies gilt für Gesetzesvorlagen, Regulierungen, Forschungsaktivitäten. Sind diese Rahmenkonzepte anschlussfähig an medienpädagogische Debatten über disruptive Versuche, in via soziale Medien geführte öffentliche Debatten einzugreifen, die sich über social bots, fake news oder andere Formen der Einflussnahme manifestieren? Erschliessen sie dazu Reflexionsräume und Handlungsoptionen? Geleitet von diesen Fragen betrachtet der vorliegende Aufsatz, die Rahmenkonzepte der EU und UNESCO, DigComp and MIL. Dabei zeigt sich, dass beide Rahmenkonzepte von Schieflagen gekennzeichnet sind. DigComp und MIL überbetonen die instrumentelle, auf Verwertung am

Natura2000 : Information und Kommunikation über die Ausweisung und die Pflege von Natura2000 Schutsgebieten : Zusammenfassung Hauptbericht 1: die Ausweisung in 27 EU Mitgliedsstaten

NARCIS (Netherlands)

Apeldoorn, van R.C.; Kruk, R.W.; Bouwma, I.M.; Ferranti, F.; Blust, De G.; Sier, A.R.J.

2010-01-01

Die Europäischen Kommission (EK) möchte die Mitgliedsstaaten unterstützen und den Austausch von Wissen, Information und Praxiserfahrung fördern: sowohl über den Prozess der Ausweisung von Vogelschutz- und FFH- Gebieten, als über mögliche Schutzmassnahmen und Instrumente für diese Gebiete in den

Improving 3D-Turbo Code's BER Performance with a BICM System over Rayleigh Fading Channel

Directory of Open Access Journals (Sweden)

R. Yao

2016-12-01

Full Text Available Classical Turbo code suffers from high error floor due to its small Minimum Hamming Distance (MHD. Newly-proposed 3D-Turbo code can effectively increase the MHD and achieve a lower error floor by adding a rate-1 post encoder. In 3D-Turbo codes, part of the parity bits from the classical Turbo encoder are further encoded through the post encoder. In this paper, a novel Bit-Interleaved Coded Modulation (BICM system is proposed by combining rotated mapping Quadrature Amplitude Modulation (QAM and 3D-Turbo code to improve the Bit Error Rate (BER performance of 3D-Turbo code over Raleigh fading channel. A key-bit protection scheme and a Two-Dimension (2D iterative soft demodulating-decoding algorithm are developed for the proposed BICM system. Simulation results show that the proposed system can obtain about 0.8-1.0 dB gain at BER of 10^{-6}, compared with the existing BICM system with Gray mapping QAM.

Gas dynamics in the central cavity of HYLIFE-II reactor

International Nuclear Information System (INIS)

Chen, X.M.; Schrock, V.E.; Peterson, P.F.; Colella, P.

1992-01-01

In a HYLIFE-II ICF reactor, the microfusion of the D-T capsule in the center of the chamber produces X-rays that can ablate a thin layer off the liquid blanket which protects the first structural wall Thisablated material will implode toward the center line of the central cavity due to the initial vacuum and cylindrical geometry, and then rebound back to the liquid blanket vent through it and exert a pressure ''impulse'' onto the structural wall. The initial ablation occurs in a very short period with very small characteristic length and the implosion and rebounding processes feature very high pressures and temperatures. The proper design of the chamber relies on the reasonably accurate analysis of the gas dynamics in the central cavity and the gas-liquid interaction. In this paper, a second order Godunov numerical method is used to solve the compressible flow equations in the central cavity. The rarefaction and shock phenomena are very well captured by the numerical calculation. The equation of state for Flibe vapor is used in the calculation along with the parameters for the HYLIFE-II design. Since the radiation transport has not yet been included in the current calculations, the vapor possesses higher energy and therefore temperature. The total mass vaporized will also be underestimated in the later time of the calculation. The incorporation of a radiation calculation into this code is our next goal

Human action contribution to ET-RR-II reactor systems unstems unavailability

International Nuclear Information System (INIS)

Sabek, M.G.

2001-01-01

This paper gives a study on the human action contribution to the systems unavailability of ET-RR-II reactor as a result of the test and maintenance procedures. The human contribution is expressed in terms of Fussel-Vesely importance which is defined by the probability that event is contributing to system failure (unavailability). The human error basic events contribution was analyzed for all initiating events and systems fault trees. The calculations result shows a high contribution value (61%) for the human error to systems unavailability. This means that the operator and the maintenance people should be highly qualified trained. Moreover, there should be programs for continuous training. Also, the procedures of tests and maintenance should be in a simple way and clear in order to minimize the contribution of the human errors. The calculations were done using the IRRAS cods

Lessons learned from Gen II NPP staffing approaches applicable to new reactors - 15003

International Nuclear Information System (INIS)

Goodnight, C.

2015-01-01

This paper discusses lessons learned from the operation of the Gen II fleet of existing nuclear power plants (NPPs), in terms of staffing, that can be applied to the final design, deployment, and operation of new reactor designs. The most significant of these lessons is the need to appropriately staff the facility, having the right number of people with the required skills and experience. This begs the question of how to identify those personnel requirements. For NPPs, there are five key factors that ultimately will determine the effectiveness and costs of operating nuclear power plants (NPPs): 1) The Nuclear Steam Supply System (NSSS) and the layout of the plant site; 2) The processes which the operating organization applies; 3) The organizational structure of the operating organization; 4) The organizational culture of the operating organization, and 5) The regulatory framework under which the licensee must operate. In summary, this paper identifies opportunities to minimize staffing and costs learned from Gen II NPPs that may be applicable for new nuclear plants. (author)

Maintenance procedures for the TITAN-I and TITAN-II reversed field pinch reactors

International Nuclear Information System (INIS)

Grotz, S.P.; Duggan, W.; Krakowski, R.; Najmabadi, F.; Wong, C.P.C.

1989-01-01

The TITAN reactor is a compact, high-power-density (neutron wall loading 18 MW/m 2 ) machine, based on the reversed-field-pinch (RFP) confinement concept. Two designs for the fusion power core have been examined: TITAN-I is based on a self-cooled lithium loop with a vanadium-alloy structure for the first wall, blanket and shield; and TITAN-II is based on an aqueous loop-in-pool design with a LiNO 3 solution as the coolant and breeder. The compact design of the TITAN fusion power core, (FPC) reduces the system to a few small and relatively low mass components, making toroidal segmentation of the FPC unnecessary. A single-piece maintenance procedure is possible. The potential advantages of single-piece maintenance procedures are: (1) Short period of down time; (2) improved reliability; (3) no adverse effects resulting from unequal levels of irradiation; and (4) ability to continually modify the FPC design. Increased availability can be expected from a fully pre-tested, single-piece FPC. Pre-testing of the FPC throughout the assembly process and prior to installation into the reactor vault is discussed. (orig.)

Physics analysis of the TIBER-II engineering test reactor

International Nuclear Information System (INIS)

Uckan, N.A.; Houlberg, W.A.; Attenberger, S.E.; Dory, R.A.; Spong, D.A.; Tolliver, J.S.; Sheffield, J.

1987-11-01

Confinement capability, burn characteristics, heating and fueling requirements, and fast alpha particle effects are assessed for the TIBER-II engineering test reactor (ETR/ITER). Confinement predictions for a wide variety of empirical scaling laws show that ignition in TIBER-II (or similar ETR-like devices) is marginal at 10 MA, whereas the design goal to achieve noninductively driven, steady-state burn with Q > 5 can easily be attained. Operation at the higher plasma currents being discussed for ITER or the attainment of higher density limits and/or favorable H-mode scalings improves the ignition capability. Pellet penetration calculations indicate that density profile control with pellets may not be feasible even for pellet velocities up to about 50 km/s, however, density peaking could result from inward pinch effects, as frequently inferred from experiments. The fast alpha contribution to pressure is substantial (10 to 30%) at TIBER (or any ETR/ITER) burn temperatures (8 to 20 keV). A relatively low level of fast alpha radial diffusion or a modest level of thermal alpha buildup significantly influences the ignition and steady-state burn capability. The fast alpha population can also modify the background plasma ballooning mode stability boundaries, lowering the beta limit β/sub crit/ - in particular, operation at the high electron temperatures needed for efficient current drive can exacerbate this effect. The use of high-energy neutral beams offers the promise of two important improvements in projected performance: an effective method for noninductive current drive and a means for controlling the current density profile deep within the plasma, as required for stable operation at high beta levels. 14 refs., 10 figs., 1 tab

Physics analysis of the TIBER-II engineering test reactor

International Nuclear Information System (INIS)

Uckan, N.A.; Houlberg, W.A.; Attenberger, S.E.; Dory, R.A.; Spong, D.A.; Tolliver, J.S.; Sheffield, J.

1987-01-01

Confinement capability, burn characteristics, heating and fueling requirements, and fast-alpha particle effects are assessed for the TIBER-II engineering test reactor (ETR/ITER). Confinement predictions for a wide variety of empirical scaling laws show that ignition on TIBER-II (or similar ETR-like devices) is marginal at 10 MA, whereas the design goal to achieve noninductively driven, steady-state burn with Q > 5 can easily be attained. Operation at the higher plasma currents being discussed for ITER or the attainment of higher density limits and/or favorable H-mode scalings improves the ignition capability. Pellet penetration calculations indicate that density profile control with pellets may not be feasible even for pellet velocities up to bout 50 km/s; however, density peaking could result from inward pinch effects, as frequently inferred from experiments. The fast alpha contribution to pressure is substantial (10-30%) at TIBER (or any ETR/ITER) burn temperatures (8-20 keV). A relatively low level of fast alpha radial diffusion or a modest level of thermal alpha buildup significantly influences the ignition and steady-state burn capability. The fast alpha population can also modify the background plasma ballooning mode stability boundaries, lowering the beta limit β/sub crit/ - in particular, operation at the high electron temperatures needed for efficient current drive can exacerbate this effect. The use of high-energy neutral beams offers the promise of two important improvements in projected performance: an effective method for noninductive current drive and a means for controlling the current density profile deep within the plasma, as required for stable operation at high beta levels

Operation and maintenance of the RA reactor in 1964, I-II, Part II; Pogon i odrzavanje reaktora RA u 1964. godini, I-II, II Deo

Energy Technology Data Exchange (ETDEWEB)

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

1964-12-15

This volume of the report contains the following 15 Annexes: Improvement of the fuel cycle economy (record No. 37009803 in INIS DB); Analysis of neutron flux increase in horizontal experimental channels of the RA reactor record No. 37005698 in INIS DB); Application of the critical system for determining the thermal neutron flux in a research reactor with central horizontal reflector ( record No. 37055005 in INIS DB); Determining the capacity of the RA reactor heat exchanger dependent on the coolant water temperature and flow; Operation of the RA reactor in forced regime; Analysis of the CEN-132 heavy water pumps failures at the RA reactor from decontamination till present; Modifications in the vacuum loop of the distillation system; Report on decontamination of the evaporator and cleaning of the condenser of the distillation system; Operation of reactor at nominal power with reduced D{sub 2}O circulation; Cooling of the RA reactor with reduced flow rate in the heavy water loop; Measurement of the heavy water level in the fuel channels of the RA reactor; Conclusions of the experts group of the RA reactor at the meeting held on November 2 and 3 1964; Conclusions of the experts group at the meeting held on November 23 1964; After heat and the cooling problem after RA reactor shut-down; Measurement of noise and vibrations on the Ra reactor heavy water system; Calculation and measurement of the uranium temperature during irradiation in the experimental channel in the reflector of the RA reactor; Temperature measurement of the reactor materials samples irradiated in the fuel channels of the RA reactor; Study of the modifications in the synchronous generators, heavy water pumps and condenser batteries of the RA reactor.

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

Energy Technology Data Exchange (ETDEWEB)

Wang, C.Y.

1993-06-01

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

Energy Technology Data Exchange (ETDEWEB)

Wang, C.Y.

1993-01-01

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Research reactors in Austria - Present situation

International Nuclear Information System (INIS)

Boeck, H.; Musilek, A.; Villa, M.

2005-01-01

In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atominstitut and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down of the ASTRA on July 31th, 1999 and its immediate decommissioning reactor and the shut down of the Argonaut reactor in Graz on August 31st, 2004 only one reactor remains operational for keeping nuclear competence in Austria which is the 250 kW TRIGA Mark II reactor. (author)

Neural Network based Minimization of BER in Multi-User Detection in SDMA

OpenAIRE

VENKATA REDDY METTU; KRISHAN KUMAR,; SRIKANTH PULLABHATLA

2011-01-01

In this paper we investigate the use of neural network based minimization of BER in MUD. Neural networks can be used for linear design, Adaptive prediction, Amplitude detection, Character Recognition and many other applications. Adaptive prediction is used in detecting the errors caused in AWGN channel. These errors are rectified by using Widrow-Hoff algorithm by updating their weights andAdaptive prediction methods. Both Widrow-Hoff and Adaptive prediction have been used for rectifying the e...

Core map generation for the ITU TRIGA Mark II research reactor using Genetic Algorithm coupled with Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Türkmen, Mehmet, E-mail: tm@hacettepe.edu.tr [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey); Çolak, Üner [Energy Institute, Istanbul Technical University, Ayazağa Campus, Maslak, Istanbul (Turkey); Ergün, Şule [Nuclear Engineering Department, Hacettepe University, Beytepe Campus, Ankara (Turkey)

2015-12-15

Highlights: • Optimum core maps were generated for the ITU TRIGA Mark II Research Reactor. • Calculations were performed using a Monte Carlo based reactor physics code, MCNP. • Single-Objective and Multi-Objective Genetic Algorithms were used for the optimization. • k{sub eff} and ppf{sub max} were considered as the optimization objectives. • The generated core maps were compared with the fresh core map. - Abstract: The main purpose of this study is to present the results of Core Map (CM) generation calculations for the İstanbul Technical University TRIGA Mark II Research Reactor by using Genetic Algorithms (GA) coupled with a Monte Carlo (MC) based-particle transport code. Optimization problems under consideration are: (i) maximization of the core excess reactivity (ρ{sub ex}) using Single-Objective GA when the burned fuel elements with no fresh fuel elements are used, (ii) maximization of the ρ{sub ex} and minimization of maximum power peaking factor (ppf{sub max}) using Multi-Objective GA when the burned fuels with fresh fuels are used. The results were obtained when all the control rods are fully withdrawn. ρ{sub ex} and ppf{sub max} values of the produced best CMs were provided. Core-averaged neutron spectrum, and variation of neutron fluxes with respect to radial distance were presented for the best CMs. The results show that it is possible to find an optimum CM with an excess reactivity of 1.17 when the burned fuels are used. In the case of a mix of burned fuels and fresh fuels, the best pattern has an excess reactivity of 1.19 with a maximum peaking factor of 1.4843. In addition, when compared with the fresh CM, the thermal fluxes of the generated CMs decrease by about 2% while change in the fast fluxes is about 1%.Classification: J. Core physics.

Utilization of a typical 250 kW TRIGA Mark II reactor at a University

International Nuclear Information System (INIS)

Villa, M.; Boeck, H.; Musilek, A.

2007-01-01

The 250 kW TRIGA Mark-II reactor operates since March 1962 at the Atominstitut Vienna/Austria. Its main tasks are nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection and dosimetry, and low temperature physics and fusion research. Academic research is carried out by students in the above mentioned fields coordinated and supervised by about 70 staff members with the aim of a masters- or PhD degree in one of the above mentioned areas. During the past 15 years about 580 students graduated through the Atominstitut. In addition, the Atominstitut co-operates closely with the nearby located IAEA in research projects, coordinated research programs (CRP) and supplying expert services. Regular training courses are carried out for the IAEA for Safeguard Trainees, fellowship places are offered for scientists from developing countries and staff members carry out expert missions to research centres in Africa, Asia and South America. Special Nuclear Material (SNM) is stored for calibration purposes at the Atominstitut belonging to the IAEA. (author)

Reactor protection system

International Nuclear Information System (INIS)

Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

1977-10-01

The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

Operation experience and maintenance at the TRIGA Mark II L.E.N.A. reactor

International Nuclear Information System (INIS)

Gngoli, F.; Berzero, A.; Lana, F.; Rosti, G.; Meloni, S.

2008-01-01

The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis, mostly for neutron activation analysis purposes. Moreover the reactor was completely shutdown in the first six months of this year to allow the dismantling of the NADIR experimental setup. The paper presents: - Reactor operation from July 1990 to June 1992; - Reactor users in the time period January 1990 - December 1991; - Specific activities of some radionuclides in the filling materials; - Specific activity of some radionuclides in thermal column materials. Operations related to dismantling of NADIR experimental facility are described. Finally the new thermal column configuration is presented. Starting from the end inside the reactor tank, a graphite layer (35 cm thick) was positioned, followed by a bismuth layer (10 cm thick) to reduce gamma-ray intensity. The old graphite rods were then positioned leaving in the central part, on the equatorial plane of the thermal column, a cavity whose vertical section has 40 cm width and 20 cm height. The bottom of the cavity, towards to the reactor tank, has been lined with additional layers of graphite (10 cm), bismuth (10 cm) and again graphite (1 cm). The new configuration allowed new experiments to be performed. The cavity in the central part has been created to allow the irradiation of large biological samples such as experimental animal and human livers. This is a peculiar step in a neutron capture boron therapy project to be carried out at the University of Pavia. In order to avoid an implemented 41 Ar production in the void space between shutters and the thermal column outer end, the external surface of the thermal column has been coated with boral sheets. The neutron flux profile, both thermal and epithermal, and cadmium ratio for gold are shown. The flux distribution appears to be adequate to proceed with the neutron capture boron therapy experiment. The LENA Health Physics Service has checked all phases of

Proposed fuel cycle for the Integral Fast Reactor

International Nuclear Information System (INIS)

Burris, L.; Walters, L.C.

1985-01-01

One of the key features of ANL's Integral Fast Reactor (IFR) concept is a close-coupled fuel cycle. The proposed fuel cycle is similar to that demonstrated over the first five to six years of operation of EBR-II, when a fuel cycle facility adjacent to EBR-II was operated to reprocess and refabricate rapidly fuel discharged from the EBR-II. Locating the IFR and its fuel cycle facility on the same site makes the IFR a self-contained system. Because the reactor fuel and the uranium blanket are metals, pyrometallurgical processes (shortned to ''pyroprocesses'') have been chosen. The objectives of the IFR processes for the reactor fuel and blanket materials are to (1) recover fissionable materials in high yield; (2) remove fission products adequately from the reactor fuel, e.g., a decontamination factor of 10 to 100; and (3) upgrade the concentration of plutonium in uranium sufficiently to replenish the fissile-material content of the reactor fuel. After the fuel has been reconstituted, new fuel elements will be fabricated for recycle to the reactor

EBR-II: twenty years of operating experience

International Nuclear Information System (INIS)

Lentz, G.L.; Buschman, H.W.; Smith, R.N.

1985-01-01

Experimental Breeder Reactor No. 2 (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. For the last 20 years EBR-II has operated safely, has demonstrated stable operating characteristics, has shown excellent performance of its sodium components, and has had an excellent plant factor. These years of operating experience provide a valuable resource to the nuclear community for the development and design of future liquid metal fast reactors. This report provides a brief description of the EBR-II plant and its early operating experience, describes some recent problems of interest to the nuclear community, and also mentions some of the significant operating achievements of EBR-II. Finally, a few words and speculations on EBR-II's future are offered. 4 figs., 1 tab

Vorläufige Mitteilung über die Beschaffenheit der Ovarialtasche von Chrysochloris, Galeopithecus und Tupaja

NARCIS (Netherlands)

Lange, de Daniel

1922-01-01

Im Jahre 1917 ist in den Anatomischen Heften die bekannte und schöne Arbeit Sobotta’s erschienen über das Vorkommen und die Bedeutung der Ovarialtasche bei den Säugetieren ¹). Das Hauptergebniss dieser Arbeit ist, dass bei alien Säugetieren (vielleicht mit Ausnahne der Primaten und des Menschen) die

The evaluation of research reactor TRIGA MARK II safety

International Nuclear Information System (INIS)

Jordan, R.; Kozuh, M.; Mavko, B.

1994-01-01

In the paper the Probabilistic Safety Analysis (PSA) of a research reactor is described. Five different initiating events were selected and analyzed with the use of event trees. Seven reactor systems were modeled with fault trees. Three groups of radiation releases were introduced - Success, Reactor-Hall, Environment - and their frequencies were estimated. The importance factors of initiating events, human errors and basic events were calculated regarding the consequence groups. (author)

TRIGA reactor as an experimental tool

Energy Technology Data Exchange (ETDEWEB)

Nahrul Khair bin Alang Mohammad Rashid (PUSPATI, Selangor (Malaysia))

1981-01-01

Article reviewed on the general features, operation and capabilities, and utilization of a research reactor, PUSPATI TRIGA MARK II. The paper also described the arrangements for the use of the PUSPATI reactor.

Triga reactor as an experimental tool

International Nuclear Information System (INIS)

Nahrul Khair bin Alang Mohammad Rashid

1981-01-01

Article reviewed on the general features, operation and capabilities, and utilization of a research reactor, PUSPATI TRIGA MARK II. The paper also described the arrangements for the use of the PUSPATI reactor

Reactor protection system design using micro-computers

International Nuclear Information System (INIS)

Fairbrother, D.B.

1977-01-01

Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

Reactor protection system design using micro-computers

International Nuclear Information System (INIS)

Fairbrother, D.B.

1976-01-01

Reactor protection systems for nuclear power plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. The paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

Visualization of neutron flux and power distributions in TRIGA Mark II reactor as an educational tool

International Nuclear Information System (INIS)

Snoj, Luka; Ravnik, Matjaz; Lengar, Igor

2008-01-01

Modern Monte Carlo computer codes (e.g. MCNP) for neutron transport allow calculation of detailed neutron flux and power distribution in complex geometries with resolution of ∼1 mm. Moreover they enable the calculation of individual particle tracks, scattering and absorption events. With the use of advanced software for 3D visualization (e.g. Amira, Voxler, etc.) one can create and present neutron flux and power distribution in a 'user friendly' way convenient for educational purposes. One can view axial, radial or any other spatial distribution of the neutron flux and power distribution in a nuclear reactor from various perspectives and in various modalities of presentation. By visualizing the distribution of scattering and absorption events and individual particle tracks one can visualize neutron transport parameters (mean free path, diffusion length, macroscopic cross section, up-scattering, thermalization, etc.) from elementary point of view. Most of the people remember better, if they visualize the processes. Therefore the representation of the reactor and neutron transport parameters is a convenient modern educational tool for the (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. The visualization of neutron flux and power distributions in Jozef Stefan Institute TRIGA Mark II research reactor is treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. (authors)

Neutronics comparative analysis between MNSR and slowpoke-II reactors

International Nuclear Information System (INIS)

Khamis, I.; Khattab, K.

1999-01-01

Neutronics analysis of both MNSR and Slowpoke reactors were made. Calculations including flux distribution, power estimation, excess and shutdown reactivity margins, flooding effects of irradiation sites, and initial investigation of fuel conversion from high to low enriched uranium were discussed. A neutronic 3-D model, dedicated mainly for the MNSR, has been developed to perform such neutronic calculations for both reactors. Well-known cell and core calculation codes such as WIMSD4 and CITATIONS have been used. It was found out that it is possible to lower the fuel enrichment of the Miniature Neutron Source Reactor (MNSR) to 20% using U O 2 as fuel instead of U Al 4 . The number of fuel elements required for the new core is 199. The use of double thickness of the bottom reflector in Slowpoke reactor made it possible to load the reactor with lower enriched fuel compared to MNSR. Values of reactivity flooding effects for single or combination of inner irradiation sites were obtained accurately. Results show good agreement with reported data for MNSR. (author)

Stalin Coctail: Schostakowitsch Kammersinfonie Nr. 2 (aus Streichquartett Nr. 3 op. 73). Pärt, Arvo: Collage über das Thema B-A-C-H / Susanne Benda

Index Scriptorium Estoniae

Benda, Susanne

1995-01-01

Uuest heliplaadist "Stalin Coctail: Schostakowitsch Kammersinfonie Nr. 2 (aus Streichquartett Nr. 3 op. 73). Pärt, Arvo: Collage über das Thema B-A-C-H. Cantus in memoriam B. Britten. Denisov Variationen über Haydns Kanon Tod ist ein langer Schlaf. Schtschedrin Stalin Coctail. Moskauer Virtuosen, Vladimir Spivakov. RCA/BMG-Ariola CD 09026 68061 2 (WD: 67'33")

Tightly coupled transient analysis of EBR-II

International Nuclear Information System (INIS)

Makowitz, H.; Lehto, W.K.; Sackett, J.I.

1988-01-01

A Tightly Coupled transient analysis system for the Experimental Breeder Reactor-II (EBR-II) is currently being tested. The system consists of a faster than real time high fidelity reactor simulation, advanced graphics displays, expert system coupling, and real time data coupling via the EBR-II data acquisition system to and from the plant and the control system. The base, first generation software has been developed and is presently being tested. Various subsystem couplings and the total system integration are being checked out. This system should enhance the diagnostic and prognostic capability of EBR-II in the near term and provide automatic control during startup and power maneuvering in the future, as well as serve as a testbed for new control system development for advanced reactors

Redetermination of di-μ-hydrido-hexahydridotetrakis(tetrahydrofurandialuminium(IIImagnesium(II

Directory of Open Access Journals (Sweden)

Hima Kumar Lingam

2010-05-01

Full Text Available The structure of the title compound, [Mg(AlH42(C4H8O4], has been redetermined at 150 K. The MgII ion is hexacoordinated to four tetrahydrofuran (THF ligands, and two AlH4− anions through bridging H atoms. The Al—H distances are more precise compared to those previously determined [Nöth et al. (1995. Chem. Ber. 128, 999–1006; Fichtner & Fuhr (2002. J. Alloys Compd, 345, 386–396]. The molecule has twofold rotation symmetry.

Über dieses Heft

Directory of Open Access Journals (Sweden)

Editors / Herausgeber

2013-06-01

Full Text Available Bilder, audiovisuelle Medien, elektronische Medien, World Wide Web: Das Erzählen vollzieht sich mittlerweile in vielfältigen Formen. Der intensive Gebrauch von neuen Medien in unterschiedlichen Zusammenhängen begegnet, überlagert und vermischt sich mit der Rezeption traditioneller literarischer Medien, die nach wie vor weit verbreitet sind. Für die Erzählforschung ergibt sich aus der veränderten kulturellen Praxis eine Vielzahl von Folgen. Die vorliegende Ausgabe von DIEGESIS soll einen Eindruck davon vermitteln. So rückt etwa ein Korpus in den Vordergrund, das lange Zeit eine Randexistenz führte: Erzählungen, die nicht oder nicht ausschließlich durch Sprache vermittelt werden, und solche, die nicht mehr nur die Kulturtechnik des Lesens und Schreibens voraussetzen. Andere Techniken, andere Gesten begleiten den Umgang mit diesen Erzählungen, und diese Differenz wird an denjenigen Stellen besonders deutlich, wo unterschiedliche Medien ineinander verschachtelt sind. Die Erzeugung von Komplexität durch intermediale Relationen charakterisiert dieses neue Korpus und öffnet für die Erzählforschung neue Arbeitsfelder. Gerade die markanten Medienwechsel, die mit der Vermarktung erfolgreicher Stoffe – als Roman, als Film, als Computerspiel, als Comic, als YouTube-Parodie – einhergehen, werfen die doppelte Frage auf, was in diesen Varianten das gemeinsame narrative Substrat, und was hingegen medienspezifische Erzählverfahren sind. Der ersten Perspektive folgt in der vorliegenden Ausgabe Matthias Brütsch mit einer kritischen Prüfung transmedialer Narrativitätsvergleiche; die zweite entfalten Sebastian Armbrusts Beitrag über Komplexität in bekannten Fernsehserien wie House und The Wire und Markus Kuhns Analyse der narrativen Funktion der Handkamera als spezifischem Verfahren in Spielfilmen und fiktionalen Internetclips. Selbstverständlich haben sich auch bestehende Kategorien der Erzählforschung unter dem Eindruck gewandelt, dass

Power level control of the TRIGA Mark-II research reactor using the multifeedback layer neural network and the particle swarm optimization

International Nuclear Information System (INIS)

Coban, Ramazan

2014-01-01

Highlights: • A multifeedback-layer neural network controller is presented for a research reactor. • Off-line learning of the MFLNN is accomplished by the PSO algorithm. • The results revealed that the MFLNN–PSO controller has a remarkable performance. - Abstract: In this paper, an artificial neural network controller is presented using the Multifeedback-Layer Neural Network (MFLNN), which is a recently proposed recurrent neural network, for neutronic power level control of a nuclear research reactor. Off-line learning of the MFLNN is accomplished by the Particle Swarm Optimization (PSO) algorithm. The MFLNN-PSO controller design is based on a nonlinear model of the TRIGA Mark-II research reactor. The learning and the test processes are implemented by means of a computer program at different power levels. The simulation results obtained reveal that the MFLNN-PSO controller has a remarkable performance on the neutronic power level control of the reactor for tracking the step reference power trajectories

Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

Energy Technology Data Exchange (ETDEWEB)

Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-07-28

Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

The nuclear reactor systems

International Nuclear Information System (INIS)

Bacher, P.

2008-01-01

This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

LMFBR operational safety: the EBR-II experience

International Nuclear Information System (INIS)

Sackett, J.I.; Allen, N.L.; Dean, E.M.; Fryer, R.M.; Larson, H.A.; Lehto, W.K.

1978-01-01

The mission of the Experimental Breeder Reactor II (EBR-II) has evolved from that of a small LMFBR demonstration plant to a major irradiation-test facility. Because of that evolution, many operational-safety issues have been encountered. The paper describes the EBR-II operational-safety experience in four areas: protection-system design, safety-document preparation, tests of off-normal reactor conditions, and tests of elements with breached cladding

BER analysis of DS-UWB system employing a laplace distribution model

KAUST Repository

Mehbodniya, Abolfazl

2011-01-01

This letter takes a new approach to extract a closed-form expression for the bit error rate (BER) of direct-sequence ultra wideband (DS-UWB) system. In the analysis, the main signal is impaired by multi-user interference (MUI) and an external source of interference originated by simultanously transmitting multiband orthogonal frequency division multiplexing (MB-OFDM) systems which are located in the vicinity of the DS-UWB receiver. All the transmission channels are affected by Nakagami-m fading. A Laplacian distribution is considered for MUI to comply more with real statistical behaviors of this kind of interference. © IEICE 2011.

Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

International Nuclear Information System (INIS)

Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

1996-12-01

Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

International Nuclear Information System (INIS)

Boulaich, Y.; Nacir, B.; El Bardouni, T.; Zoubair, M.; El Bakkari, B.; Merroun, O.; El Younoussi, C.; Htet, A.; Boukhal, H.; Chakir, E.

2011-01-01

Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.

Proceedings on the Second Autumn School on Reactor Physics EROEFI II

Energy Technology Data Exchange (ETDEWEB)

Racz, A [ed.

1996-12-31

The main topics of the Reactor Physics School were neutron and reactor physical calculations, reactor safety, systems theory, simulation of accidents, reactor monitoring system, computer codes and procedures for solving specific problems in the field of nuclear reactors (especially safety). A special attention was paid to the AGNES project. Papers falling in the INIS scope have been abstracted and indexed individually for the INIS database. (K.A.).

Proceedings on the Second Autumn School on Reactor Physics EROEFI II

International Nuclear Information System (INIS)

Racz, A.

1995-01-01

The main topics of the Reactor Physics School were neutron and reactor physical calculations, reactor safety, systems theory, simulation of accidents, reactor monitoring system, computer codes and procedures for solving specific problems in the field of nuclear reactors (especially safety). A special attention was paid to the AGNES project. Papers falling in the INIS scope have been abstracted and indexed individually for the INIS database. (K.A.)

Kollektivismus, Individualismus und Persönlichkeit als Prädiktoren von Einstellungen gegenüber psychologischer Beratung - Ein Kulturvergleich zwischen deutschen und chinesischen Studierenden

OpenAIRE

Schmahl, Anne

2007-01-01

Die vorliegende Diplomarbeit beschäftigt sich mit einem Kulturvergleich zwischen deutschen (n=103) und chinesischen Studierenden (n=98) bezüglich ihrer Einstellungen gegenüber psychologischer Beratung. Als Prädiktoren der Einstellungen wurden hierbei zum einen die kulturelle Dimension Individualismus-Kollektivismus nach Hofstede (2005) und zum anderen das Fünf-Faktoren-Modell der Persönlichkeit (FFM) herangezogen. Der Einfluss der Persönlichkeit auf die Einstellungen gegenüber psychologischer...

US DOE Idaho national laboratory reactor decommissioning

International Nuclear Information System (INIS)

Szilagyi, Andrew

2012-01-01

The United States Department of Energy (DOE) primary contractor, CH2M-WG Idaho was awarded the cleanup and deactivation and decommissioning contract in May 2005 for the Idaho National Lab (INL). The scope of this work included dispositioning over 200 Facilities and 3 Reactors Complexes (Engineering Test Reactor (ETR), Materials Test Reactor (MTR) and Power Burst Facility (PBF) Reactor). Two additional reactors were added to the scope of the contract during the period of performance. The Zero Power Physics Reactor (ZPPR) disposition was added under a separate subcontractor with the INL lab contractor and the Experimental Breeder Reactor II (EBR-II) disposition was added through American Recovery and Reinvestment Act (ARRA) Funding. All of the reactors have been removed and disposed of with the exception of EBR-II which is scheduled for disposition approximately March of 2012. A brief synopsis of the 5 reactors is provided. For the purpose of this paper the ZPPR reactor due to its unique design as compared to the other four reactors, and the fact that is was relatively lightly contaminated and irradiated will not be discussed with the other four reactors. The ZPPR reactor was readily accessible and was a relatively non-complex removal as compared to the other reactors. Additionally the EBR-II reactor is currently undergoing D and D and will have limited mention in this paper. Prior to decommissioning the reactors, a risk based closure model was applied. This model exercised through the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA), Non-Time Critical Removal Action (NTCRA) Process which evaluated several options. The options included; No further action - maintain as is, long term stewardship and monitoring (mothball), entombment in place and reactor removal. Prior to commencing full scale D and D, hazardous constituents were removed including cadmium, beryllium, sodium (passivated and elemental), PCB oils and electrical components, lead

Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Khang, Ngo Phu [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

The Dalat nuclear research reactor was reconstructed from TRIGA MARK II reactor, built in 1963 with nominal power of 250 KW, and reached its planned nominal power of 500 kW for the first time in Feb. 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc. (author). 10 refs., 10 figs., 2 tabs.

Instrumentation and control improvements at Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Christensen, L.J.; Planchon, H.P.

1993-01-01

The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

Safety evaluation for instrumentation and control system upgrading project of Malaysian TRIGA MARK II PUSPATI Research reactor

International Nuclear Information System (INIS)

Ridha Roslan; Nik Mohd Faiz Khairuddin

2013-01-01

Full-text: Malaysian TRIGA MARK II research reactor has been in safe operation since its first criticality in 1982. The reactor is licensed to be operated by Malaysian Nuclear Agency to perform training and research development related activities. Due to its extensive operation since last three decades, the option of modifications for safety and safety-related item and component become a necessary to replace the outdated equipment to a stat-of-art, reliable technologies. This paper will present the current regulatory activities performed by Atomic Energy Licensing Board (AELB) to ensure the upgrading of analogue to digital instrumentation and control system is implemented in safe manner. The review activity includes documentation review, manufacturer quality audit and on-site inspection for commissioning. The review performed by AELB is based on The International Atomic Energy Agency (IAEA) Safety Requirements NS-R-4, entitled Safety of Research Reactors. During this endeavour, AELB seeks technical cooperation from Korea Institute of Nuclear Safety (KINS), the nuclear experts organization of the country of origin of the instrumentation and control technology. The regulatory activity is still on-going and is expected to be completed by issuance of Authorization for Restart on December 2013. (author)

Evolution of thermal-hydraulics testing in EBR-II

International Nuclear Information System (INIS)

Golden, G.H.; Planchon, H.P.; Sackett, J.I.; Singer, R.M.

1987-01-01

A thermal-hydraulics testing and modeling program has been underway at the Experimental Breeder Reactor-II (EBR-II) for 12 years. This work culminated in two tests of historical importance to commercial nuclear power, a loss of flow without scram and a loss of heat sink wihout scram, both from 100% initial power. These tests showed that natural processes will shut EBR-II down and maintain cooling without automatic control rod action or operator intervention. Supporting analyses indicate that these results are characteristic of a range of sizes of liquid metal cooled reactors (LMRs), if these reactors use metal driver fuel. This type of fuel is being developed as part of the Integral Fast Reactor Program at Argonne National Laboratory. Work is now underway at EBR-II to exploit the inherent safety of metal-fueled LMRs with regard to development of improved plant control strategies. (orig.)

Machoideal och/eller meningslös underhållning? : Om fem mäns berättelser kring krigsfilmen Saving Private Ryan med fokus på maskulinitet och publiken som meningsskapare.

OpenAIRE

Olsson, Eva

2005-01-01

Uppsatsen undersöker fem mäns berättelser kring krigsfilmen Saving Private Ryan, med fokus på maskulinitet. Föreställningarna och bilderna kring maskulinitet som kan urskiljas i intervjupersonernas berättelser, knyts till R.W. Connells teoretiska ramverk kring hegemonisk maskulinitet och de hierarkiska relationerna mellan olika maskuliniteter. Vidare resoneras det kring hur berättelserna exemplifierar en syn på mediepublik som aktiva meningsskapare, då intervjupersonerna tar fasta på olika as...

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

International Nuclear Information System (INIS)

Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

2002-01-01

An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

2014-01-01

The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

Fast reactor physics - an overview

International Nuclear Information System (INIS)

Lee, S.M.

2004-01-01

An introduction to the basic features of fast neutron reactors is made, highlighting the differences from the more conventional thermal neutron reactors. A discussion of important feedback reactivity mechanisms is given. Then an overview is presented of the methods of fast reactor physics, which play an important role in the successful design and operation of fast reactors. The methods are based on three main elements, namely (i) nuclear data bases, (ii) numerical methods and computer codes, and (iii) critical experiments. These elements are reviewed and the present status and future trends are summarized. (author)

Use of EBR-II as a principal fast breeder reactor irradiation test facility in the U.S

International Nuclear Information System (INIS)

Staker, R.G.; Seim, O.S.; Beck, W.N.; Golden, G.H.; Walters, L.C.

1975-01-01

The EBR-II as originally designed and operated by the Argonne National Laboratory was successful in demonstrating the operation of a sodium-cooled fast breeder power plant with a closed fuel reprocessing cycle. Subsequent operation has been as an experimental facility where thousands of irradiation tests have been performed. Conversion to this application entailed the design and fabrication of special irradiation subassemblies for in-core irradiations, additions to existing facilities for out-of-core irradiations, and additions to existing facilities for out-of-core experiments. Experimental subassemblies now constitute about one third of the core, and changes in the core configuration occur about monthly, requiring neutronic and thermal-hydraulics analyses and monitoring of the reactor dynamic behavior. The surveillance programs provided a wealth of information on irradiation induced swelling and creep, in-reactor fracture behavior, and the compatibility of materials with liquid sodium. (U.S.)

EBR-II: search for the lost subassembly

International Nuclear Information System (INIS)

King, R.W.; Buschman, H.W.; Poloncsik, J.; Remsburg, J.S.; Sine, H.W.

1983-01-01

Experimental Breeder Reactor II (EBR-II) has been operating for nearly 20 years as part of the foundation of the US Department of Energy's LMFBR development program. During that time, the EBR-II fuel-handling system has performed extremely well, especially considering the conditions under which much of the system operates and the reliability required to maintain the high plant factor routinely demonstrated by EBR-II. Since EBR-II is a pool-type reactor, much of the fuel handling is done remotely within the sodium-filled primary tank at 371 0 C. Activities involved in locating a misplaced fuel subassembly in the primary tank are described

Theoretical analysis of nuclear reactors (Phase II), I-V, Part III, Reactor poisoning; Razrada metoda teorijske analize nuklearnih reaktora (II faza) I-V, III Deo, Zatrovanje reaktora, II faza

Energy Technology Data Exchange (ETDEWEB)

Pop-Jordanov, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1962-10-15

This phase is dealing with influence of all the fission products except Xe{sup 135} on the reactivity of a reactor, usually named as reactor poisoning. The first part of the report is a review of methods for calculation of reactor poisoning. The second part shows the most frequently used method for calculation of cross sections and yields of pseudo products (for thermal neutrons). The system of equations was adopted dependent on the conditions of the available computer system. It is described in part three. Detailed method for their application is described in part four and results obtained are presented in part five.

Systematical investigations of the emission of carbon 14 from a TRIGA-Mark-II reactor - methods and results

International Nuclear Information System (INIS)

Pfeiffer, K.J.

1981-01-01

Almost no information is available about the extent of the carbon-14 releases from a research reactor. For this reason this report is dealing with the emission of C-14 from the Vienna TRIGA-Mark-II reactor. In addition the resulting radiation exposure is estimated. Due to the low activity concentrations of C-14 in research reactor effluents special requirements are necessary for sampling and measuring. A technique providing both sufficient lower limit of detection and little effort of sample preparation was developed. Carbon dioxide was trapped by bubbling air taken from the stack through washing bottles containing an aqueous solution of sodium hydroxide. After sampling a precipitate of CaCO 3 was formed and about 8 g of calcium carbonate were counted as a gel suspension by liquid scintillation counting. The formation of the gel was provided by mixing water with a scintillation cocktail originally developed for uptake of high quantities of aqueous solutions. The resulting lower limit of detection was about 50 Bq/kg carbon being equivalent to 9mBq/m 3 air. Concluding the measurements, which were carried out by weekly counting and a period of some 14 months, a normalized release rate of about 280 Bq (7, 1μCi) was found. This release rate is somewhat higher than the reported values for power reactors, because the main activity is produced by activation of air in experimental equipments. (author)

Safety evaluation of the Dalat research reactor operation

International Nuclear Information System (INIS)

Long, V.H.; Lam, P.V.; An, T.K.

1989-01-01

After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

Overview of the fast reactors fuels program

International Nuclear Information System (INIS)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides

A program for the a priori evaluation of detection limits in instrumental neutron activation analysis using a SLOWPOKE II reactor

International Nuclear Information System (INIS)

Galinier, J.L.; Zikovsky, L.

1982-01-01

A program that permits the a priori calculation of detection limits in monoelemental matrices, adapted to instrumental neutron activation analysis using a SLOWPOKE II reactor, is described. A simplified model of the gamma spectra is proposed. Products of (n,p) and (n,α) reactions induced by the fast components of the neutron flux that accompanies the thermal flux at the level of internal irradiation sites in the reactor have been included in the list of interfering radionuclides. The program calculates in a systematic way the detection limits of 66 elements in an equal number of matrices using 153 intermediary radionuclides. Experimental checks carried out with silicon (for short lifetimes) and aluminum and magnesium (for intermediate lifetimes) show satisfactory agreement with the calculations. These results show in particular the importance of the contribution of the (n,p) and (n,α) reactions in the a priori evaluation of detection limits with a SLOWPOKE type reactor [fr

A transient overpower experiment in EBR-II

International Nuclear Information System (INIS)

Herzog, J.P.; Tsai, H.; Dean, E.M.; Aoyama, T.; Yamamoto, K.

1994-01-01

The TOPI-IE test was a transient overpower test on irradiate mixed-oxide fuel pins in the Experimental Breeder Reactor-II (EBR-II). The test, the fifth in a series, was part of a cooperative program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan to conduct operational transient testing on mixed-oxide fuel pins in the metal-fueled EBR-II. The principle objective of the TOPI-1E test was to assess breaching margins for irradiated mixed-oxide fuel pins over the Plant Protection System (PPS) thresholds during a slow, extended overpower transient. This paper describes the effect of the TOPI-1E experiment on reactor components and the impact of the experiment on the long-term operability of the reactor. The paper discusses the role that SASSYS played in the pre-test safety analysis of the experiment. The ability of SASSYS to model transient overpower events is detailed by comparisons of data from the experiment with computed reactor variables from a SASSYS post-test simulation of the experiment

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

Photocatalytic reactors for treating water pollution with solar illumination. II: a simplified analysis for flow reactors

Energy Technology Data Exchange (ETDEWEB)

Sagawe, G.; Bahnemann, D. [Inst. fuer Technische Chemie, Univ. Hannover, Hannover (Germany); Brandi, R.J.; Cassano, A.E. [INTEC (Univ. Nacional del Litoral and CONICET), Santa Fe (Argentina)

2003-07-01

Very frequently outgoing streams of real wastewaters do not have a definite and constant composition. Additionally, when the degradation process makes use of solar irradiation, the photon flux is hardly constant. These two factors strongly militate against the use of very elaborate, exact models for analyzing the performance of the employed reactors. In these cases, approximate methods may be the most practical approach. One possible way is presented in this paper. The observed photonic efficiency concept developed in a previous contribution (sagawe et al., 2002a) is applied to continuous reactors for both steady state and transient operations of photocatalytic reactions applied to wastewaters decontamination processes. For this reactor the local observed photonic efficiency, defined at each reactor longitudinal position, is the convenient property to express the concentration spatial evolution. It is also shown that the description of the reactor performance employing a mass balance can be done in a rather simple way introducing a mass-moving coordinate transformation that remodel the mass inventory and permits working with simpler ordinary differential equations. (orig.)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Demonstration of passive safety features in EBR-II

International Nuclear Information System (INIS)

Planchon, H.P. Jr.; Golden, G.H.; Sackett, J.I.

1987-01-01

Two tests of great importance to the design of future commercial nuclear power plants were carried out in the Experimental Breeder Reactor-II on April 3, 1986. These tests, (viewed by about 60 visitors, including 13 foreign LMR specialists) were a loss of flow without scram and a loss of heat sink without scram, both from 100% initial power. In these tests, inherent feedback shut the reactor down without damage to the fuel or other reactor components. This resulted primarily from advantageous characteristics of the metal driver fuel used in EBR-II. Work is currently underway at EBR-II to develop a control strategy that promotes inherent safety characteristics, including survivability of transient overpower accidents. In parallel, work is underway at EBR-II on the development of state-of-the-art plant diagnostic techniques

Performance Evaluation of Monolith Based Immobilized Acetylcholinesterase Flow-Through Reactor for Copper(II Determination with Spectrophotometric Detection

Directory of Open Access Journals (Sweden)

Parawee Rattanakit

2014-01-01

Full Text Available A monolith based immobilized acetylcholinesterase (AChE flow-through reactor has been developed for the determination of copper(II using flow injection spectrophotometric system. The bioreactor was prepared inside a microcapillary column by in situ polymerization of butyl methacrylate, ethylene dimethacrylate, and 2,2-dimethoxy-1,2-diphynyletane-1-one in the presence of 1-decanol, followed by vinyl azlactone functionalization and AChE immobilization. The behavior of AChE before and after being immobilized on the monolith was evaluated by kinetic parameters from Lineweaver and Burk equation. The detection was based on measuring inhibition effect on the enzymatic activity of AChE by copper(II using Ellman’s reaction with spectrophotometric detection at 410 nm. The linear range of the calibration graph was obtained over the range of 0.02–3.00 mg L−1. The detection limit, defined as 10% inhibition (I10, was found to be 0.04 mg L−1. The repeatability was 3.35 % (n=5 for 1.00 mg L−1 of copper(II. The proposed method was applied to the determination of copper(II in natural water samples with sampling rate of 4 h−1.

Alternate form and placement of short lived reactor waste and associated fuel hardware for decommissioning of EBR-II

Energy Technology Data Exchange (ETDEWEB)

Planchon, H.P.; Singleterry, R.C. Jr.

1995-12-01

Upon the termination of EBR-II operation in 1994, the mission has progressed to decommissioning and waste cleanup of the facility. The simplest method to achieve this goal is to bury the raw fuel and activated steel in an approved burial ground or deep geologic repository. While this might be simple, it could be very expensive, consume much needed burial space for other materials, and leave large amounts of fissile easily available to future generations. Also, as with any operation, an associated risk to personnel and the public from the buried waste exists. To try and reduce these costs and risks, alternatives to burial are sought. One alternative explored here for EBR-II is to condition the fuel and store the fission products and steel either permanently or temporarily in the sealed primary boundary of the decommissioned reactor. The first problem is to identify which subassemblies are going to be conditioned and their current composition and decay time. The next problem is to identify the conditioning process and determine the composition and form of the waste streams. The volume, mass, heat, and curie load of the waste streams needs to be determined so a waste-assembly can be designed. The reactor vessel and internals need to be analyzed to determine if they can handle these loads. If permanent storage is the goal, then mechanisms for placing the waste-assembly in the reactor vessel and sealing the vessel are needed. If temporary storage is the goal, then mechanisms for waste-assembly placement and retrieval are needed. This paper answers the technical questions of volume, mass, heat, and curie loads while just addressing the other questions found in a safety analysis. The final conclusion will compare estimated risks from the burial option and this option.

Outage and BER analysis for ultrawideband-based WPAN in Nakagami-m fading channels

KAUST Repository

Mehbodniya, Abolfazl

2011-09-01

This paper presents a performance analysis of multiband orthogonal frequency-division multiplexing (MB-OFDM) in ultra wideband (UWB)-based personal area networks (UPANs). A UPAN consists of devices with different UWB technologies at the physical layer. Approximate expressions for the outage probability and average bit error rate (BER) are derived in closed form for the MB-OFDM target receiver, taking into account multi-user interference (MUI), as well as external interference in the form of time-hopping (TH) and direct-sequence (DS) UWB signals. © 2010 IEEE.

Preparation of chitin–silica composites by in vitro silicification of two-dimensional Ianthella basta demosponge chitinous scaffolds under modified Stöber conditions

Energy Technology Data Exchange (ETDEWEB)

Wysokowski, Marcin [Institute of Chemical Technology and Engineering, Faculty of Chemical Technology, Poznan University of Technology, M. Skłodowskiej-Curie 2, PL-60965 Poznan (Poland); Behm, Thomas [Institute of Experimental Physics, TU Bergakademie Freiberg, Liepziger 23, 09599 Freiberg (Germany); Born, René [Institute of Materials Science, Dresden University of Technology, Helmholtzstraße 10, 01069 Dresden (Germany); Bazhenov, Vasilii V. [Institute of Experimental Physics, TU Bergakademie Freiberg, Liepziger 23, 09599 Freiberg (Germany); Meißner, Heike; Richter, Gert [Faculty of Medicine Carl Gustav Carus, Dresden University of Technology, Fetscherstraße 74, 01307 Dresden (Germany); Szwarc-Rzepka, Karolina [Institute of Chemical Technology and Engineering, Faculty of Chemical Technology, Poznan University of Technology, M. Skłodowskiej-Curie 2, PL-60965 Poznan (Poland); Makarova, Anna; Vyalikh, Denis [Institute of Solid State Physics, Dresden University of Technology, Helmholtzstraße 10, 01069 Dresden (Germany); Schupp, Peter [Institute for Chemistry and Biology of the Marine Environment, University of Oldenburg, Emsstr. 20, 26382 Wilhelmshaven (Germany); Jesionowski, Teofil, E-mail: teofil.jesionowski@put.poznan.pl [Institute of Chemical Technology and Engineering, Faculty of Chemical Technology, Poznan University of Technology, M. Skłodowskiej-Curie 2, PL-60965 Poznan (Poland); Ehrlich, Hermann, E-mail: hermann.ehrlich@physik.tu-freiberg.de [Institute of Experimental Physics, TU Bergakademie Freiberg, Liepziger 23, 09599 Freiberg (Germany)

2013-10-15

Chitin is a biopolymer found in cell walls of various fungi and skeletal structures of numerous invertebrates. The occurrence of chitin within calcium- and silica-containing biominerals has inspired development of chitin-based hybrids and composites in vitro with specific physico-chemical and material properties. We show here for the first time that the two-dimensional α-chitin scaffolds isolated from the skeletons of marine demosponge Ianthella basta can be effectively silicified by the two-step method with the use of Stöber silica micro- and nanodispersions under Extreme Biomimetic conditions. The chitin–silica composites obtained at 120 °C were characterized by the presence of spherical SiO{sub 2} particles homogeneously distributed over the chitin fibers, which probably follows from the compatibility of Si–OH groups to the hydroxyl groups of chitin. The biocomposites obtained were characterized by various analytical techniques such as energy dispersive spectrometry, scanning electron microscopy, thermogravimetric/differential thermal analyses as well as X-ray photoelectron spectroscopy, Fourier transform infrared and Raman spectroscopy to determine possible interactions between silica and chitin molecule. The results presented proved that the character and course of the in vitro chitin silicification in Stöber dispersions depended considerably on the degree of hydrolysis of the SiO{sub 2} precursor. - Highlights: • 2D α-chitin scaffolds isolated from marine demosponge can be effectively silicified using Stöber silica. • The chitin–silica composites were obtained under Extreme Biomimetic conditions. • Character and course of the in vitro chitin silicification in Stöber dispersions is discussed.

Deactivation of the EBR-II complex

International Nuclear Information System (INIS)

Michelbacher, J.A.; Earle, O.K.; Henslee, S.P.; Wells, P.B.; Zahn, T.P.

1996-01-01

In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D and D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D and D plan has necessitated this current action

Deactivation of the EBR-II complex

Energy Technology Data Exchange (ETDEWEB)

Michelbacher, J A; Earle, O K; Henslee, S P; Wells, P B; Zahn, T P

1996-01-01

In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D and D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D and D plan has necessitated this current action.

Neutron beam utilization at the TRIGA Mark II reactor Vienna

International Nuclear Information System (INIS)

Villa, M.; Boeck, H.; Ismail, S.; Koerner, S.; Baron, M.; Hainbuchner, M.; Badurek, G.; Buchelt, R.J.

1999-01-01

A review is given about the research activities around the 250 kw TRIGA reactor Vienna, which are adequate to other neutron sources of comparable or bigger size. The topics selected for presentation range from neutron radiography, materials irradiation, neutron small-angle scattering, neutron activation analysis, neutron polarization to neutron interferometry. It is the aim of this presentation to stimulate programs for more efficient use around TRIGA research reactors with neutron flux densities of 1013 cm-2a-1 at the center of the reactor core. We briefly describe the experimental facilities installed at the 250 kw TRIGA reactor of the Austrian Universities in Vienna and present a great part of the current research activities performed with them. We believe that most of the techniques and experiments presented here are adequate for implementation to other reactors of similar or even higher power. Those technologies which require extremely specialized know-how not generally available at every research Inst.e will not be treated here or are just mentioned without any further details.(author)

Comparative analysis of Carnaval II Library

International Nuclear Information System (INIS)

Santos Bastos, W. dos

1981-01-01

The Carnaval II cross sections library from the french fast reactor calculation system is evaluated in two ways: 1 0 ) a comparative analysis of the calculations system for fast reactors at IEN (Instituto de Engenharia Nuclear) using a 'benchmark' model is done; 2 0 ) a comparative analysis in relation to the french system itself is also done, using calculations realized with two versions of the french library: the SETR-II and the CARNAVAL IV, the first one being anterior and the second one posterior to the Carnaval II version, the one used by IEN. (Author) [pt

Physical Characteristics of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy

1994-10-01

The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part I: Monte Carlo Simulation

International Nuclear Information System (INIS)

Huda, M.Q.; Chakrobortty, T.K.; Rahman, M.; Sarker, M.M.; Mahmood, M.S.

2003-05-01

This study deals with the neutronic analysis of the current core configuration of a 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and S(α, β) scattering functions from the ENDF/B-V library were used. The validation of the model against benchmark experimental results is presented. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is correctly simulating the TRIGA reactor. (author)

Importance of reduced sulfur for the equilibrium chemistry and kinetics of Fe(II), Co(II) and Ni(II) supplemented to semi-continuous stirred tank biogas reactors fed with stillage.

Science.gov (United States)

Shakeri Yekta, Sepehr; Lindmark, Amanda; Skyllberg, Ulf; Danielsson, Asa; Svensson, Bo H

2014-03-30

The objective of the present study was to assess major chemical reactions and chemical forms contributing to solubility and speciation of Fe(II), Co(II), and Ni(II) during anaerobic digestion of sulfur (S)-rich stillage in semi-continuous stirred tank biogas reactors (SCSTR). These metals are essential supplements for efficient and stable performance of stillage-fed SCSTR. In particular, the influence of reduced inorganic and organic S species on kinetics and thermodynamics of the metals and their partitioning between aqueous and solid phases were investigated. Solid phase S speciation was determined by use of S K-edge X-ray absorption near-edge spectroscopy. Results demonstrated that the solubility and speciation of supplemented Fe were controlled by precipitation of FeS(s) and formation of the aqueous complexes of Fe-sulfide and Fe-thiol. The relatively high solubility of Co (∼ 20% of total Co content) was attributed to the formation of compounds other than Co-sulfide and Co-thiol, presumably of microbial origin. Nickel had lower solubility than Co and its speciation was regulated by interactions with FeS(s) (e.g. co-precipitation, adsorption, and ion substitution) in addition to precipitation/dissolution of discrete NiS(s) phase and formation of aqueous Ni-sulfide complexes. Copyright © 2014 Elsevier B.V. All rights reserved.

Hilfe oder Hindernis? Meinungen finnischer Sprachstudierender über Mehrsprachigkeit als Ressource beim Deutschlernen

Directory of Open Access Journals (Sweden)

Sabine Grasz

2017-10-01

Full Text Available Der Fremdsprachenunterricht in Finnland befindet sich gerade in einer Umbruchphase, denn mit der Einführung neuer Lehrpläne im Jahr 2016 werden Verfahren der Mehrsprachigkeitsdidaktik in Zukunft eine zentrale Rolle spielen. In diesem Beitrag wird untersucht, ob und wie finnische Universitätsstudierende, die selbst von einer eher monolingual ausgerichteten Lerntradition geprägt sind, ihr mehrsprachiges Repertoire als Ressource beim Lernen weiterer Fremdsprachen sehen. Zu diesem Zweck wurde eine Befragung unter Lehramtsstudierenden der Fächer Anglistik und Skandinavistik, die Germanistik als Nebenfach studieren möchten, durchgeführt. Die Analyse zeigt, dass die Studierenden zwar mehrheitlich positive Einstellungen gegenüber Mehrsprachigkeit aufweisen, dass es aber Entwicklungsbedarf im Hinblick auf Sprachbewusstheit, multilinguale Fähigkeiten und bei den Einstellungen gegenüber Sprachmischungen gibt. Foreign language education in Finland is currently undergoing a period of change, as the introduction of new curricula in 2016 will lead to a central role of multilingualism in teaching. This article examines whether and how Finnish university students, who themselves were influenced by a more monolingual learning tradition, consider their multilingual repertoire as a resource for learning other foreign languages. For this purpose, a survey was conducted among teacher students of English and Swedish, who would like to study German as a minor. The analysis shows that while the majority of students have positive attitudes towards multilingualism, there is a need for development in terms of language awareness, multilingual skills, and attitudes towards language mixes.

A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

International Nuclear Information System (INIS)

Dolan, T.J.; Longhurst, G.R.; Garcia-Otero, E.

1992-01-01

We have designed a vacuum disengager system to remove tritium from the Flibe (Li 2 BeF 4 ) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction Φ ∼10 -5 of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation

Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

International Nuclear Information System (INIS)

Bhattacharyya, S.K.

1978-08-01

A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

Power reactors in Member States. 1979 edition

International Nuclear Information System (INIS)

1979-01-01

This is the fifth issue of a periodic computer-based listing of nuclear power reactors, presenting the situation as of 1 May 1979. The basic design data for all reactors in operation, under construction, planned and shut down have been included. The following two tables are included to give a general picture of the overall situation: Table I: Reactor types and net electrical power. Table II: Reactor units and net electrical powered by country cummulated by year

The study of time-dependent neutronics parameters of the 2MW TRIGA Mark II Moroccan research reactor using BUCAL1 computer code

International Nuclear Information System (INIS)

Bakkari, B. El; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Riyach, I.; Otmani, S.; Marcih, I.; Elbadri, H.; El Bardouni, T; Merroun, O.; Boukhal, H.; Zoubair, M.; Htet, A.; Chakir, M.

2010-01-01

The 2-MW TRIGA MARK II research reactor at Centre National de l'Energie, des Sciences et des Techniques Nucleaires (CNESTEN) achieved initial criticality on May 2, 2007 with 71 fuel elements. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower and training and production of radioisotopes for their use in agriculture, industry and medicine. This work aims to study the time-dependent neutronics parameters of the TRIGA reactor for elaborating and planning of an in-core fuel management strategy to maximize the utilization of the TRIGA fluxes, using a new elaborated burnup computer code called 'BUCAL1'. The code can be used to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. It was developed to incorporate the neutron absorption tally/reaction information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The use of Monte Carlo method and punctual cross section data characterizing the MCNP code allows an accurate simulation of neutron life cycle in the reactor, and the integration of data on the entire energy spectrum, thus a more accurate estimation of results than deterministic code can do. Also, for the purpose of this study, a full-model of the TRIGA reactor was developed using the MCNP5 code. The validation of the MCNP model of the TRIGA reactor was made by benchmarking the reactivity experiments. (author)

Overview of the fast reactors fuels program. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

Experience with lifetime limits for EBR-II core components

International Nuclear Information System (INIS)

Lambert, J.D.B.; Smith, R.N.; Golden, G.H.

1987-01-01

The Experimental Breeder Reactor No. 2 (EBR-II) is operated for the US Department of Energy by Argonne National Laboratory and is located on the Idaho National Engineering Laboratory where most types of American reactor were originally tested. EBR-II is a complete electricity-producing power plant now in its twenty-fourth year of successful operation. During this long history the reactor has had several concurrent missions, such as demonstration of a closed Liquid-Metal Reactor (LMR) fuel cycle (1964-69); as a steady-state irradiation facility for fuels and materials (1970 onwards); for investigating effects of operational transients on fuel elements (from 1981); for research into the inherent safety aspects of metal-fueled LMR's (from 1983); and, most recently, for demonstration of the Integral Fast Reactor (IFR) concept using U-Pu-Zr fuels. This paper describes experience gained at EBR-II in defining lifetime limits for LMR core components, particularly fuel elements

copper(II)

Indian Academy of Sciences (India)

Unknown

bis(2,2,6,6-tetramethyl-3,5-heptadionato)copper(II) ... Abstract. Equilibrium concentrations of various condensed and gaseous phases have been thermodyna- ... phere, over a wide range of substrate temperatures and total reactor pressures.

Maintenance of reactor recirculation pumps [Paper No.: II-1

International Nuclear Information System (INIS)

Ansari, M.A.; Bhat, K.P.

1981-01-01

At Tarapur Atomic Power Station (TAPS), two reactor recirculation pumps are provided, one each for the two reactor units. The performance of pumps has been uniformly good; however, leakage through the cartridge type, two stage, mechanical seals which are installed on these pumps was encountered on few occasions. The paper describes the leakage problems, identification of certain design deficiencies and rectification carried out at TAPS for overcoming these problems. (author)

Status of fast breeder reactor development in Germany

International Nuclear Information System (INIS)

Hueber, R.; Kathol, W.; Kempken, M.

1991-01-01

The KNK, the sodium cooled compact reactor is an experimental nuclear power plant of 20 MW electric power. Since 1977, it has been operated with fast reactor cores as KNK II. The KNK II/3 core was designed. The core fabrication has been largely completed. In 1990, the KNK II plant achieved a time availability of 56%. On January 8, 1991 KNK II was shut down for inspection. Since pre-nuclear commissioning was completed the Kalkar Nuclear Power Station SNR 300 has been operated in a mode similar to that of a power station. In March 1991 the financing partners decided not to prolong the standby phase because they do not think that the last construction permit and the operation permit will be issued within a definite period of time. The partners were convinced that the lack of progress in the licensing procedure was not caused by basic safety deficiencies of the project but by the way the licensing procedure was executed. The German fast breeder programme is now concentrated on contributions to the European Fast Reactor. (author)

Blinde Mimesis. Über Ordo und Kontingenz in der literaturgeschichtlichen Traditionsbildung (Horaz und Petron

Directory of Open Access Journals (Sweden)

Jürgen Paul Schwindt

2010-11-01

Full Text Available me truncus illapsus cerebro / sustulerat, nisi Faunus ictum / dextra levasset, Mercurialium / custos virorum.(Hor. carm. 2, 17, 27–30... et ipse Trimalchio capillatus caduceum tenebat Minervaque ducente Romam intrabat.(Petron. 29, 3Es ist zuweilen das Schicksal besonders gut erforschter Texte, daß sie im System der Bezüge, der Quellen und Verweise, das die gelehrte Erklärung über, neben und unter ihnen errichtet hat (Hyper-, Para-, Prae- oder Subtext, zu verschwimmen, konturlos zu werden, ...

Testing the applicability of the k 0-NAA method at the MINT's TRIGA MARK II reactor

International Nuclear Information System (INIS)

Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

2006-01-01

The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k 0 method has become the preferred standardization method of NAA (k 0 -NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k 0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k 0 -NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters (α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k 0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k 0 -NAA method at the MINT

Testing the applicability of the k0-NAA method at the MINT's TRIGA MARK II reactor

Science.gov (United States)

Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

2006-08-01

The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k0 method has become the preferred standardization method of NAA ( k0-NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k0-NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters ( α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k0-NAA method at the MINT.

International Working Group on Fast Reactors Eight Annual Meeting, Vienna, Austria, 15-18 April 1975. Summary Report. Part II

International Nuclear Information System (INIS)

1975-07-01

The Eighth Annual Meeting of the IAEA International Working Group on Past Reactors was held at the IAEA Headquarters in Vienna, Austria, from 15 to 18 April 1975. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programmes in the field of LMPBR’s and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

Deactivation of the EBR-II complex

Energy Technology Data Exchange (ETDEWEB)

Michelbacher, J.A.; Earle, O.K.; Henslee, S.P. [and others

1997-12-31

In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D&D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D&D plan has necessitated this current action. The EBR-II is a pool-type reactor. The primary system contains approximately 87,000 gallons of sodium, while the secondary system has 13,000 gallons. In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility has been built to react the sodium to a dry carbonate powder in a two stage process. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in the primary and secondary systems must be either reacted or inerted to preclude future concerns with sodium-air reactions that generate explosive mixtures of hydrogen and leave corrosive compounds. Residual amounts of sodium on components will effectively {open_quotes}solder{close_quotes} components in place, making future operation or removal unfeasible.

Deactivation of the EBR-II complex

International Nuclear Information System (INIS)

Michelbacher, J.A.; Earle, O.K.; Henslee, S.P.

1997-01-01

In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D ampersand D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D ampersand D plan has necessitated this current action. The EBR-II is a pool-type reactor. The primary system contains approximately 87,000 gallons of sodium, while the secondary system has 13,000 gallons. In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility has been built to react the sodium to a dry carbonate powder in a two stage process. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in the primary and secondary systems must be either reacted or inerted to preclude future concerns with sodium-air reactions that generate explosive mixtures of hydrogen and leave corrosive compounds. Residual amounts of sodium on components will effectively open-quotes solderclose quotes components in place, making future operation or removal unfeasible

Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-08-15

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Training experience at Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

1978-01-01

The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report

BER-3.2 report: Methodology for justification and optimization of protective measures including a case study

International Nuclear Information System (INIS)

Hedemann Jensen, P.; Sinkko, K.; Walmod-Larsen, O.; Gjoerup, H.L.; Salo, A.

1992-07-01

This report is a part of the Nordic BER-3 project's work to propose and harmonize Nordic intervention levels for countermeasures in case of nuclear accidents. This report focuses on the methodology for justification and optimization of protective measures in case of a reactor accident situation with a large release of fission products to the environment. The down-wind situation is very complicated. The dose to the exposed society is almost unpredictable. The task of the radiation protection experts: To give advice to the decision makers on averted doses by the different actions at hand in the situation - is complicated. That of the decision makers is certainly more: On half of the society they represent, they must decide if they wish to follow the advices from their radiation protection experts or if they wish to add further arguments - economical or political (or personal) - into their considerations before their decisions are taken. Two analysis methods available for handling such situations: cost-benefit analysis and multi-attribute utility analysis are described in principle and are utilized in a case study: The impacts of a Chernobyl-like accident on the Swedish island of Gotland in the Baltic Sea are analyzed with regard to the acute consequences. The use of the intervention principles found in international guidance (IAEA 91, ICRP 91), which can be summarized as the principles of justification, optimization and avoidance of unacceptable doses, are described. How to handle more intangible factors of a psychological or political character is indicated. (au) (6 tabs., 3 ills., 17 refs.)

Biological Tests for Boron Neutron Capture Therapy Research at the TRIGA Mark II Reactor in Pavia

Energy Technology Data Exchange (ETDEWEB)

Protti, N.; Ballarini, F.; Bortolussi, S.; De Bari, A.; Stella, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy); Nuclear Physics National Institute (INFN), Pavia (Italy); Bruschi, P. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy); Bakeine, J.G.; Cansolino, L.; Clerici, A.M. [Laboratory of Experimental Surgery, Department of Surgery, University of Pavia, Pavia (Italy)

2011-07-01

The thermal column of the TRIGA Mark II reactor of the Pavia University is used as an irradiation facility to perform biological tests and irradiations of living systems for Boron Neutron Capture Therapy (BNCT) research. The suitability of the facility has been ensured by studying the neutron flux and the photon background in the irradiation chamber inside the thermal column. This characterization has been realized both by flux and dose measurements as well as by Monte Carlo simulations. The routine irradiations concern in vitro cells cultures and different tumor animal models to test the efficacy of the BNCT treatment. Some results about these experiments will be described. (author)

Description of the Triton reactor

International Nuclear Information System (INIS)

1967-09-01

The Triton reactor is an enriched uranium pool type reactor. It began operation in 1959, after a divergence made on the June 30 the same year. Devoted to studies of radiation protection, its core can be displaced in the longitudinal direction. The pool can be separated in two unequal compartments by a wall. The Triton core is placed in a small compartment, the Nereide core in the big compartment. A third compartment without water is called Naiade II, is separated by a concrete wall in which is made a window closed by an aluminium plate (2.50 m x 2.70 m). The Naiade II hole is useful for protection experiments using the Nereide core. After a complete refitting, the power of the triton reactor that reached progressively from 1.2 MW to 2 MW, then 3 MW has reached in August 1965 6.5 MW. The reactor has been specialized in irradiations in fix position, the core become fix, the nereide core has been hung mobile. Since it has been used for structure materials irradiation, for radioelements fabrication and fundamental research. The following descriptions are valid for the period after August 1965 [fr

The integral fast reactor

International Nuclear Information System (INIS)

Till, C.E.

1987-01-01

On April 3rd, 1986, two dramatic demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the Integral Fast Reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics and also makes possible a simplified closed fuel cycle and waste process improvements

Revision of the stiletto fly genera Acupalpa Kröber and Pipinnipons Winterton (Diptera, Therevidae, Agapophytinae using cybertaxonomic methods, with a key to Australasian genera

Directory of Open Access Journals (Sweden)

Shaun Winterton

2011-05-01

Full Text Available Australian stiletto flies of the sister-genera Acupalpa Kröber, 1912 and Pipinnipons Winterton, 2001 (Diptera: Therevidae: Agapophytinae are revised. Twelve new species of Acupalpa are described, while Acupalpa imitans (White, 1915, comb. n. is transferred from Pipinnipons and Acupalpa albimanis (Kröber, 1914, comb. n. is transferred from Ectinorhynchus Macquart as a senior synonym of Acupalpa pollinosa Mann. The total number of species of Acupalpa is therefore increased to 19: A. albimanis (Kröber, comb. n., A. albitarsa Mann, A. boharti sp. n., A. divisa (Walker, A. dolichorhyncha sp. n., A. glossa sp. n., A. imitans (White, comb. n., A. irwini Winterton, A. melanophaeos sp. n., A. miaboolya sp. n., A. minuta sp. n., A. minutoides sp. n., A. notomelas sp. n., A. novayamarna sp. n., A. rostrata Kröber, A. semirufa Mann, A. westralica sp. n., A. yalgoo sp. n. and A. yanchep sp. n. Three new species of Pipinnipons are described, increasing the total number of species to five: P. chauncyvallis sp. n., P. fascipennis (Kröber, P. kampmeierae sp. n., P. kroeberi Winterton, and P. sphecoda sp. n. Pipinnipons and Acupalpa are rediagnosed in light of the new species presented herein and revised keys to species are included. A dichotomous key to genera of Australasian Therevidae is included. As an empirical example of cybertaxonomy, taxonomic descriptions were composed using a character matrix developed in Lucid Builder (in Structured Descriptive Data (SDD format to generate natural language descriptions supplemented by online specimen and image databases. Web resources are provided throughout the document including: a links to high resolution colour images of all species on Morphbank, b registration of authors, publications, taxon names and other nomenclatural acts in Zoobank, with assignment of Life Science Identifiers (LSIDs for each, c links to Genbank accession records for DNA sequences, and d assignment of LSIDs to specimen records with links

A atualidade da visão de Ab’Sáber / The current view of Ab’Sáber

Directory of Open Access Journals (Sweden)

Claudio Antonio Di Mauro

2012-04-01

Full Text Available A produção científica⃰ pela lavra do Professor Aziz Nacib Ab´Sáber é revisitada em, pelo menos, três aspectos considerados fundamentais: 1 a geomorfologia e a evolução paleoclimática no terciário e quaternário, quando o autor reconhece que os estudos de fisiologia de paisagens, ainda que essenciais para os objetivos dos geomorfologistas, somente podem ser esclarecidos à custa de pesquisas marcadamente interdisciplinares; 2 as mudanças climáticas em processos, que são ignoradas nos conhecimentos sobre o presente, e as heranças obtidas ao longo da história do Planeta Terra, para planejar, com cuidados, as ações que as sociedades promoveriam em busca do “desenvolvimento”; 3 o sítio urbano de São Paulo, em meados do século passado, quando, em menos de 15 anos, a cidade que nasceu e cresceu sobre colinas, e que elegeu as colinas como seu tipo ideal de sítio topográfico, desceu para as várzeas, ocupando setores e glebas importantes das planícies recém enxugadas pelas obras da Companhia Light. As contribuições, inicialmente elaboradas sob a perspectiva geomorfológica, tinham, em suas essências, a visão geográfica do cidadão do mundo.

Comparison of standardised decommissioning costing tools on pilot Vienna TRIGA MARK-II research reactor

International Nuclear Information System (INIS)

Hornacek, M.; Kristofova, K.; Slugen, V.; Zachar, M.; Stummer, T.

2017-01-01

The main purpose of the paper is to compare decommissioning costing code CERREX (Cost Estimation for Research Reactors in Excel) with advanced calculation methodology applied in eOMEGA-RR code. CERREX code was developed in line with the IAEA recommendations for decommissioning costing of research facilities and fully implements the ISDC (International Structure for Decommissioning Costing of Nuclear Installations) structure and costing methodology. In comparison with CERREX, usually applied in preliminary costing, the code eOMEGA-RR incorporates the realistic activity and material flow during decommissioning process (e.g. decontamination, dismantling and waste management). This advanced approach enables to carry out the decommissioning planning and costing more effectively. Moreover, the user-friendly interface helps to perform wide range of sensitivity analyses. In order to meet the above mentioned objectives, the model calculation costing case for TRIGA MARK-II research reactor in Vienna was developed in both calculation codes. The whole process covered four step-by-step procedures to be implemented. At first, inventory database taking into account physical as well as radiological parameters (e.g.: contamination, dose rates, nuclide vectors, limits and conditions) was developed. At second, advanced decommissioning costing case using CERREX and eOMEGA-RR code was created. At third, sensitivity analyses to estimate the impact of changing input parameters on calculated results were performed. Finally, costing results obtained from both cost calculation codes are compared and discussed. (authors)

Status report of Indonesian research reactors

International Nuclear Information System (INIS)

Arbie, B.; Supadi, S.

1995-01-01

A general description of the three Indonesia research reactors, their irradiation facilities and future prospect are given. The 250 kW Triga Mark II in Bandung has been in operation since 1965 and in 1972 its designed power was increased to 1000 kW. The core grid from the previous 250 kW Triga Mark II was then used by Batan for designing and constructing the Kartini reactor in Yogyakarta. This reactor commenced its operation in 1979. Both Triga reactors have served a wide spectrum of utilization such as for manpower training in nuclear engineering, radiochemistry, isotope production, and beam research in solid state physics. The Triga reactor management in Bandung has a strong cooperation with the Bandung Institute of Technology and the one in Yogyakarta with the Gadjah Mada University which has a Nuclear Engineering Department at its Faculty of Engineering. In 1976 there emerged an idea to have a high flux reactor appropriate for Indonesia's intention to prepare an infrastructure for both nuclear energy and non-energy industry era. Such an idea was then realized with the achievement of the first criticality of the RSG-GAS reactor at the Serpong area. It is now expected that by early 1992 the reactor will reach its full 30 MW power level and by the end of 1992 the irradiation facilities be utilizable fully for future scientific and engineering work. As a part of the national LEU fuel development program a study has been underway since early 1989 to convert the RSG-GAS reactor core from using oxide fuel to using higher loading silicide fuel. (author)

Activation and waste disposal of the TITAN RFP [reversed-field-pinch] reactors

International Nuclear Information System (INIS)

Cheng, E.T.; Conn, R.W.

1988-01-01

The TITAN-I lithium self-cooled and TITAN-II aqueous lithium nitrate solution-cooled fusion reactors are based on the reversed-field-pinch (RFP) toroidal confinement concept and operate at high power density with an 18.1 MW/m 2 neutron wall loading. These designs were analyzed to study the activation and waste disposal aspects of such high-power density reactors. It was found that because of the use of V-3Ti-1Si (TITAN-I) and reduced activation ferritic steel (TITAN-II) as structural alloys for the first wall, blanket, reflector, and shield components, all the TITAN components except the divertor collector plates can be classified as shallow-land burial (10CFR61 Class C or better) nuclear waste for disposal, provided that the impurity elements, niobium and molybdenum, can be controlled below about 1 and 0.3 appm levels, respectively. The average annual disposal masses were estimated to be 150 and 96 tonnes, respectively, for the 1,000 MW TITAN-I and TITAN-II reactors. This corresponds to about 11% of the total mass in the fusion power core of both reactors. The divertor collector plates are fabricated with tungsten because of its low particle sputtering properties. These divertor collector plates in the TITAN-I reactor will be qualified as Class C waste after 18.1 MW-y/m 2 operation. The waste disposal rating of the divertor collector plates in the TITAN-II reactor, however, is estimated to be a factor of 4 higher than allowed for Class C disposal, because of the soft neutron spectrum in the beryllium environment. The annual disposal mass of this non-Class C waste is 0.35 tons, about 0.04% of the average annual discharge mass for the TITAN-II reactor. An additional 74 m 3 annual discharge of Class C waste containing 14 C may be needed for the TITAN-II reactor because of the use of nitrate salt in the aqueous coolant as the tritium breeder. 13 refs., 6 tabs

Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Sook; Lee, Byung Doo

2007-11-15

Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report.

Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

International Nuclear Information System (INIS)

Kim, Hyun Sook; Lee, Byung Doo

2007-11-01

Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report

Safety assessments relating to the use of new fuels in research reactors: application to the case of FRM 2 reactor fuel

International Nuclear Information System (INIS)

Abou Yehia, H.; Bars, G.; Tran Dai

2001-01-01

After giving a brief reminder of the procedure applied in France for the licensing of the use of a new fuel type or design in a research reactor, we outline the main safety aspects associated with such a modification. Finally, by way of an example, we focus on the safety assessment relating to the IRIS irradiation device used in SILOE reactor, in particular for the qualification of the fuel dedicated to FRM II reactor of the Technical University of Munich. This qualification was carried out on a U 3 Si 2 fuel plate enriched to about 90 % in weight of 235 U and containing 1.5 g of uranium per cm 3 . The evaluation performed by the IPSN for GRS did not call into question the choice of U 3 Si 2 fuel plates for the FRM-II reactor. (authors)

Irradiation of microphones in the EBR-II core

International Nuclear Information System (INIS)

Gavin, A.P.; Anderson, T.T.; Bobis, J.P.

1976-06-01

Six ANL developed high temperature microphone (acoustic detectors) have been exposed in flowing sodium in the In-Core Instrument Test Facility (INCOT) in the Experimental Breeder Reactor-II (EBR-II) for seven months without any indications of serious degradation of signal output due to the exposure. The YY05 experiment (EBR-II INCOT experiment designation) was performed to obtain data which would be useful in evaluating the ability of the microphones whose active elements are lithium niobate to serve as sensors for acoustic surveillance of fast breeder reactors. The reactor was at full power for 136 days of the experiment exposure period. The microphone temperatures varied from 371 0 C (700 0 F) to 621 0 C (1150 0 F). Neutron exposure varied from 2.64 x 10 22 nvt for the microphone at the elevation of the bottom of the EBR-II core to 0.24 x 10 22 nvt for the microphone at the elevation of the top of an EBR-II fuel assembly. The maximum gamma dose was 5 x 10 12 rads

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

International Nuclear Information System (INIS)

Khan, Jahirul Haque

2013-01-01

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

Modernization of reactor instrumentation for research reactors at Trombay

International Nuclear Information System (INIS)

Darbhe, M.D.; Chaudhuri, H.

1989-01-01

The three research reactors at Trombay, viz., Apsara, Cirus and Zerlina were commissioned in 1956, 1960 and 1961 respectively. The nuclear instrumentation designs were based on the vacuum tube technology, which was prevalent during those days. The effect of component obsolescence of critical components like vacuum tubes, magnetic amplifiers and sensitrol meter relays was strongly felt since early 1970s. Also, the failure rates of the units were observed to show an increasing trend due to ageing and lack of good quality indigenous spares. Hence it was proposed to replace the nuclear instrumentation units for the three reactors, with those employing modern, state of the art solid state devices, keeping indigenous content as high as practicable. The work started in 1977 with the preparations of specifications and the project was scheduled to be completed in 1981. The project was divided into two phases. The Phase I comprising of nuclear channels common to all reactors and Phase II consisting exclusively of regulating system units of Cirus. The salient stages of project progress and completion were: (i) Fabrication and testing of final design prototypes was completed by end of 1982. (ii) Commissioning of new units at Apsara was completed in January 1984. (iii) Commissioning of new units at Cirus was completed in September 1984. An account of experience in all these stages and problems encountered is given. (author). 6 figs

Operation and maintenance of the 250 kW TRIGA Mark II reactor at the J. Stefan Institute

International Nuclear Information System (INIS)

Dimic, V.

1982-01-01

Over the last two years the TRIGA Mark II reactor in Ljubljana has been operated at an energy release of about 2250 MWh or about 4200 hours per year. In this period, about 2000 samples were irradiated. Since the last TRIGA Owners' Conference there has been an increase in all operational data because of a very extensive programme of irradiation of molybdenum for the everyday production of technetium-99 m by a solvent extraction method. Because of its age and absolencence replacement of the console electronics was considered some time ago. Therefore, partly new instrumentation was installed this year. A new console is under construction. Furthermore, a new core configuration was established after 7 fresh FLIP fuel elements were delivered by GA. At this time it was noticed that 2 dummy elements are stuck in the upper grid plate. They will be exchanged during the regular maintenance work in August this year. During the last two years the reactor has been operated without any longer shut-down due to technical difficulties. (author)

Measurements of neutron flux distributions in the core of the Ljubljana TRIGA Mark II Reactor

International Nuclear Information System (INIS)

Rant, J.; Ravnik, M.; Mele, I.; Dimic, V.

2008-01-01

Recently the Ljubljana TRIGA Mark II Reactor has been refurbished and upgraded to pulsed operation. To verify the core design calculations using TRIGAP and PULSTR1 codes and to obtain necessary data for future irradiation and neutron beam experiments, an extensive experimental program of neutron flux mapping and neutron field characterization was carried out. Using the existing neutron measuring thimbles complete axial and radial distributions in two radial directions were determined for two different core configurations. For one core configuration the measurements were also carried out in the pulsed mode. For flux distributions thin Cu (relative measurements) and diluted Au wires (absolute values) were used. For each radial position the cadmium ratio was determined in two axial levels. The core configuration was rather uniform, well defined (fresh fuel of a single type, including fuelled followers) and compact (no irradiation channels or gaps), offering unique opportunity to test the computer codes for TRIGA reactor calculations. The neutron flux measuring procedures and techniques are described and the experimental results are presented. The agreement between the predicted and measured power peaking factors are within the error limits of the measurements (<±5%) and calculations (±10%). Power peaking occurs in the B ring, and in the A ring (centre) there is a significant flux depression. (authors)

Current status for TRR-II Cold Neutron Source

International Nuclear Information System (INIS)

Lee, C.H.; Guung, T.C.; Lan, K.C.; Wang, C.H.; Chan, Y.K.; Shieh, D.J.

2001-01-01

The Taiwan Research Reactor (TRR) project (TRR-II) is carrying out at Institute of Nuclear Energy Research (INER) from October 1998 to December 2006. The purpose of Cold Neutron Source (CNS) project is to build entire CNS facility to generate cold neutrons within TRR-II reactor. The objective of CNS design is to install CNS facility with a competitive brightness of cold neutron beam to other facilities in the world. Based on the TRR-II CNS project schedule, the conceptual design for TRR-II CNS facility has been completed and the mock-up test facility for full-scale hydrogen loop has been designed. (author)

Microchannel Methanation Reactors Using Nanofabricated Catalysts, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — Makel Engineering, Inc. (MEI) and the Pennsylvania State University (Penn State) propose to develop and demonstrate a microchannel methanation reactor based on...

Actinide behavior in the integral fast reactor

International Nuclear Information System (INIS)

Courtney, J.C.

1993-05-01

Goal of this project is to determine the consumption of Np-237, Pu-240, Am-241, and Am-243 in the Integral Fast Reactor (IFR) fuel cycle. These four actinides set the long term waste management criteria for spent nuclear fuel; if it can be demonstrated that they can be efficiently consumed in the IFR, then requirements for nuclear waste repositories can be much less demanding. Irradiations in the Experimental Breeder Reactor II (EBR-II) at Argonne National Laboratory's site near Idaho Falls, Idaho, will be conducted to determine fission and transmutation rates for the four nuclides. The experimental effort involves target package design, fabrication, quality assurance, and irradiation. Post irradiation analyses are required to determine the fission rates and neutron spectra in the EBR-II core

Blockchain - Von Supply Chain Finance über Smart Payment bis Smart Contracting: BLOG-Beitrag auf https://www.de-hub.de

OpenAIRE

Henke, Michael

2018-01-01

Alle reden über die Blockchain. Das Thema wird engagiert bis overhyped diskutiert. Gleichzeitig jedoch werden der mächtige Einfluss und die künftigen Auswirkungen von Blockchains auf Supply Chain Management und Handel immer noch unterschätzt.

PODESY program for flux mapping of CNA II reactor:

International Nuclear Information System (INIS)

Ribeiro Guevara, Sergio

1988-01-01

The PODESY program, developed by KWU, calculates the spatial flux distribution of CNA II reactor through a three-dimensional expansion of 90 incore detector measurements. The calculation is made in three steps: a) short-term calculation which considers the control rod positions and it has to be done each time the flux mapping is calculated; b) medium-term calculation which includes local burn-up dependent calculation made by diffusion methods in macro-cell configurations (seven channels in hexagonal distribution), and c) long-term calculation, or macroscopic flux determination, that is a fitting and expansion of measured fluxes, previously corrected by local effects, using the eigen functions of the modified diffusion equation. The paper outlines development of step (c) of the calculation. The incore detectors have been located in the central zone of the core. In order to obtain low errors in the expansion procedure it is necessary to include additional points, whose flux values are assumed to be equivalent to detector measurements. These flux values are calculated with detector measurements and a spatial flux distribution calculated by a PUMA code. This PUMA calculation employs a smooth burn-up distribution (local burn-up variations are considered in step (b) of the whole calculation) representing the state of core evolution at the calculation time. The core evolution referred to ends when the equilibrium core condition is reached. Additionally, a calculation method to be employed in the plant in case of incore detector failures, is proposed. (Author) [es

FIX-II/3025, BWR FIX-II Pump Trip Experiment 3025, Immediate Split Size Break

International Nuclear Information System (INIS)

NILSSON, Lars; GUSTAFSSON, Per-Ake; GUSTAFSON, Lennart; JANCZAK, Rajmund; OESTERLUNDH, Ingrid

1992-01-01

1 - Description of test facility: The FIX-II facility is a volume scaled 1:777 representation of a Swedish BWR with external pumps. The pressure vessel contains a 36 rod full length bundle and a spray condenser at the top to allow steady state operation. The downcomer, bypass channels and guide tube volumes are represented by external piping. The intact loop represents three of the four external reactor loops. The broken loop is constructed such that both guillotine breaks and split breaks may be simulated. The facility is equipped with ADS-simulation, but no ECCS injection are included. The FIX-II loop is also suited to investigate response of pump trips and MSIV closures in internal pump reactors. 2 - Description of test: Test 3025 simulates an intermediate size split break in one of the four main recirculation lines. The break area was 31 per cent of the scaled down pipe area of the reactor. The initial power of the 36-rod bundle was 3.38 MW, corresponding to the hot channel power of the reactor

Seismic and cask drop excitation evaluation of the tower shielding reactor

International Nuclear Information System (INIS)

Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N.

1989-01-01

During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs

Seismic and cask drop excitation evaluation of the Tower Shielding Reactor

International Nuclear Information System (INIS)

Stover, R.L.; Harris, S.P.; Johnson, J.J.; Sumodobila, B.N.

1989-01-01

During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations

Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

Science.gov (United States)

Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

2012-10-01

A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. Copyright © 2012 Elsevier Ltd. All rights reserved.

Ein Lehrbuch über Psychiatrie und Psychotherapie – Geschlechtsunterschiede unter der Lupe? A Textbook for Psychiatry and Psychotherapy—Gender Differences under the Microscope?

Directory of Open Access Journals (Sweden)

Isabella Heuser

2007-07-01

Full Text Available Anke Rohde und Andreas Marneros haben sich ein hohes Ziel gesetzt: ein übersichtliches Lehrbuch für Psychiatrie und Psychotherapie, das erstmals eine systematische Zusammenstellung des Wissens über geschlechtsspezifische Aspekte von Symptomatologie, Epidemiologie, Diagnostik, Pharmakotherapie und Psychotherapie psychischer Störungen enthält. Bemerkenswert ist dabei zweierlei: Erstens werden bei der Betrachtung geschlechtsspezifischer Unterschiede im Sinne der Genderforschung Frauen und Männer berücksichtigt, zweitens wird ein vollständiger Überblick sowohl über die häufigen und bekannten psychischen Störungen als auch die äußerst seltenen psychiatrischen Erkrankungen der gesamten Altersspanne gegeben. Wenngleich die Qualität der einzelnen Beiträge deutlich zwischen sehr differenzierter und leider auch für ein Handbuch eindeutig zu oberflächlicher Betrachtung variiert, liegt in dem Band ein weitgehend informatives Nachschlagewerk vor, das einen ersten Überblick über die geschlechtsspezifische Psychiatrie gibt.Anke Rohde and Andres Marneros have given themselves a lofty goal: To put together a systematic collection for the first time that contains the entire body of knowledge on the gender-specific aspects of symptomatology, epidemiology, diagnostics, pharmacotherapy, and psychotherapy of psychological disturbances. There are two notable aspects: First, in accordance with contemporary gender analysis, the examination of the gender-specific differences takes both women and men into account; second, the book provides a complete overview of the common and well-known psychological disturbances as well as the extremely rare psychiatric illnesses along the entire age spectrum. Although the quality of the individual articles clearly ranges from the very sophisticated to a superficial treatment inappropriate for a textbook, the volume itself is, for the most part, an informative work of reference that offers an initial overview of

Mediennutzung und Informationsbeschaffung über Studium- bzw. Berufswahl - Ingenieurberufe in den Medien

Directory of Open Access Journals (Sweden)

Amina Ovcina Cajacob

2014-06-01

Full Text Available Diese Publikation basiert auf der Befragung von 4 125 Jugendlichen (SchülerInnen der Altersgruppe 12 bis 21 Jahre. Dafür wurden im Zeitraum vom 22. Oktober bis 12. November 2013 Jugendliche befragt, um Antworten auf folgenden Fragen zu bekommen: Welches Bild haben die Schüler vom Ingenieurberuf? Wie ist ihr allgemeines Mediennutzungsverhalten und wie beeinflussen Medien das Bild des Ingenieurs? Welche Berufe interessieren Jugendliche und wie informieren sie sich über Studien- bzw. Berufswahl? Welche Einflüsse spielen bei den Jugendlichen bezüglich der eigenen Berufs- und Studienwahl eine Rolle? Die Ergebnisse dieser umfassenden Analyse sind Grundlage für die Konzeption einer crossmedialen Sensibilisierungskampage, die zum Ziel hat, Vorurteile über den Ingenieurberuf abzubauen, Wissen und ein zeitgemässes Berufsbild zu vermitteln sowie Neugier auf ein Ingenieurstudium zu wecken. This publication is based on a survey of 4125 young people (pupils and students aged 12 to 21 years. In the period from October 22 to November 12, 2013 young people were interviewed to obtain answers to the following questions: . What an idea do young people have about an engineer and his profession? . What is their general media usage behavior? . How do media influence the image of the engineer? . Which professions are the most attractive for young people and how do they inform themselves about study or career choices? . Which influences are important for the study and career choice of young people? The results of this comprehensive analysis are the basis for the conception of a cross-media awareness campaign, which aims to reduce prejudices about the engineering profession, to impart knowledge as well as a contemporary professional image of engineer and to arouse interest for an engineering degree.

Culham conceptual Tokamak reactor MkII. Conceptual layout of buildings for a twin reactor power station

International Nuclear Information System (INIS)

Guthrie, J.A.S.; Harding, N.H.

1981-01-01

This paper discusses the conceptual design of the nuclear complex of a 2400 MWe twin fusion reactor power station utilising common services and a single containment building. The design is based upon environmental and maintenance logistical requirements, the provision of adequate storage, workshop and construction facilities and the constraints imposed by the geometry of the main and auxiliary reactor coolant systems. (author)

Overview of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Nguyen Thai Sinh; Luong Ba Vien

2016-01-01

The present reactor called Dalat Nuclear Research Reactor (DNRR) has been reconstructed from the former TRIGA Mark II reactor which was designed by General Atomic (GA, San Diego, California, USA), started building in early 1960s, put into operation in 1963 and operated until 1968 at nominal power of 250 kW. In 1975, all fuel elements of the reactor were unloaded and shipped back to the USA. The DNRR is a 500-kW pool-type research reactor using light water as both moderator and coolant. The reactor is used as a neutron source for the purposes of: (1) radioactive isotope production; (2) neutron activation analysis; and (3) research and training

The decommissioning of a small nuclear reactor

International Nuclear Information System (INIS)

Neset, K.; Christensen, G.C.; Lundby, J.E.; Roenneberg, G.A.

1990-02-01

The JEEP II reactor at Kjeller, Norway has been used as a model for a study of the decommissioning of a small research reactor. A radiological survey is given and a plan for volume reducing, packaging, certifying, classifying and shipping of the radioactive waste is described. 23 refs., 4 figs

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

International Nuclear Information System (INIS)

Uddin, M.N.; Sarker, M.M.; Khan, M.J.H.; Islam, S.M.A.

2010-01-01

The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal-hydraulics study of the TRIGA core.

Das Thema “Altern” in Arno Geigers Roman «Alles über Sally»

Directory of Open Access Journals (Sweden)

Meike Dackweiler

2013-11-01

Full Text Available Alles über Sally (All about Sally is the fifth novel of the successful Austrian author Arno Geiger. While it was both praised and criticized for being a contemporary adaption of the adultery novel, little importance was attached to the theme of ageing, which pervades the whole novel. Moreover, adulterous female characters over fifty are rare in contemporary German literature. Given these premises, this essay examines the composition of the ageing characters in the novel and provides a discussion of the social construction of images of ageing.

GFDM performance in terms of BER, PAPR and OOB and comparison to OFDM system

Energy Technology Data Exchange (ETDEWEB)

Antapurkar, Shwetal K., E-mail: h2014077@pilani.bits-pilani.ac.in; Pandey, Avinash, E-mail: h2014083@pilani.bits-pilani.ac.in [P G student, Dept. of EEE, Birla Institute of Technology and Science, Pilani, Rajasthan, India-333031 (India); Gupta, K. K., E-mail: kgupta@pilani.bits-pilani.ac.in [Asst Prof, Dept. of EEE, Birla Institute of Technology and Science, Pilani, Rajasthan, India-333031 (India)

2016-03-09

Generalized frequency division multiplexing is a multicarrier modulation technique which can be foreseen as a potential alternative for upcoming wireless networks. GFDM attractive features include reduced out-of-band radiation(OOB) and low peak-to-average ratio(PAPR), which are the crucial shortcomings of OFDM used in present day wireless communication networks. This paper gives detailed description of GFDM system model and further studies and validates through simulations, the performance of GFDM in terms of OOB, PAPR and Bit Error Rate (BER) and compares the results obtained with OFDM system.

GFDM performance in terms of BER, PAPR and OOB and comparison to OFDM system

International Nuclear Information System (INIS)

Antapurkar, Shwetal K.; Pandey, Avinash; Gupta, K. K.

2016-01-01

Generalized frequency division multiplexing is a multicarrier modulation technique which can be foreseen as a potential alternative for upcoming wireless networks. GFDM attractive features include reduced out-of-band radiation(OOB) and low peak-to-average ratio(PAPR), which are the crucial shortcomings of OFDM used in present day wireless communication networks. This paper gives detailed description of GFDM system model and further studies and validates through simulations, the performance of GFDM in terms of OOB, PAPR and Bit Error Rate (BER) and compares the results obtained with OFDM system.

Die kupferzeitliche Siedlung Pietrele an der Unteren Donau. Bericht über die Ausgrabungen und geomorphologischen Untersuchungen im Sommer 2010

DEFF Research Database (Denmark)

Nørgaard, Heide Wrobel

2012-01-01

Seit der Kampagne 2008 und dem letzten Zwischenbericht über den nichtmetallenen Schmuck wurden zahlreiche weitere Stücke dieser Kleinfundkategorie auf dem Tell und in der Außensiedlung gefunden. Auch das Formenspektrum hat sich erheblich erweitert. Der in der bisherigen Forschung als elitär anges...

Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

International Nuclear Information System (INIS)

Pahl, R.G.; Porter, D.L.; Lahm, C.E.; Hofman, G.L.

1990-01-01

Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to >15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel

CCF analysis of BWR reactor shutdown systems based on the operating experience at the TVO I/II in 1981-1993

International Nuclear Information System (INIS)

Mankamo, T.

1996-04-01

The work constitutes a part of the project conducted within the research program of the Swedish Nuclear Power Inspectorate SKI, aimed to develop the methods and data base for the Common Cause Failure (CCF) analysis of highly redundant reactor scram systems. The data analysis for the TVO I/II plant is focused on the hydraulic scram system, and control rods and drives. It covers operating experiences from 1981 through 1993. (9 refs., 9 figs., 7 tabs.)

Study and application of ANISN and DOT-II nuclear cores in reactor physics problems

International Nuclear Information System (INIS)

Dias, Artur Flavio

1980-01-01

To solve time-independent neutrons and/or gamma rays transport problems in nuclear reactors, two codes available at IPEN were studied and applied to solve benchmark problems. The ANISN code solves the one-dimensional Boltzmann transport equation for neutrons or gamma rays, in plane, spherical, or cylindrical geometries. The DOT-II code solves the same equation in two-dimensional space for plane, cylindrical and circular geometries. General anisotropic scattering allowed in both codes. Moreover, pointwise convergence criteria, and alternate step function difference equations are also used in order to remove the oscillating flux distributions, sometimes found in discrete ordinates solutions. Basic theories and numerical techniques used in these codes are studied and summarized. Benchmark problems have been solved using these codes. Comparisons of the results show that both codes can be used with confidence in the analysis of nuclear problems. (author)

WARRIOR II, a high performance modular electric robot system

International Nuclear Information System (INIS)

Downton, G.C.

1996-01-01

Initially designed for in-reactor welding by the Central Electricity Generating Board, WARRIOR has been developed using the concept of modular technology to become a light-weight, high performance robotic system. Research work on existing machines for in-reactor inspection and repair and heavy duty hydraulic manipulators was progressed in order to develop WARRIOR II, a versatile in-reactor welding system usable at any nuclear power station light enough to be deployed by existing remote handling equipment. WARRIOR II can be significantly reconfigured quickly to pursue different ends. (UK)

Interim waste storage for the Integral Fast Reactor

International Nuclear Information System (INIS)

Benedict, R.W.; Phipps, R.D.; Condiff, D.W.

1991-01-01

The Integral Fast Reactor (IFR), which Argonne National Laboratory is developing, is an innovative liquid metal breeder reactor that uses metallic fuel and has a close coupled fuel recovery process. A pyrochemical process is used to separate the fission products from the actinide elements. These actinides are used to make new fuel for the reactor. As part of the overall IFR development program, Argonne has refurbished an existing Fuel Cycle Facility at ANL-West and is installing new equipment to demonstrate the remote reprocessing and fabrication of fuel for the Experimental Breeder Reactor II (EBR-II). During this demonstration the wastes that are produced will be treated and packaged to produce waste forms that would be typical of future commercial operations. These future waste forms would, assuming Argonne development goals are fulfilled, be essentially free of long half-life transuranic isotopes. Promising early results indicate that actinide extraction processes can be developed to strip these isotopes from waste stream and return them to the IFR type reactors for fissioning. 1 fig

Thai research reactor

International Nuclear Information System (INIS)

Aramrattana, M.

1987-01-01

The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

Study of advanced fission power reactor development for the United States. Volume II

International Nuclear Information System (INIS)

1976-01-01

This report presents the results of a multi-phase research study which had as its objective the comparative study of various advanced fission reactors and evaluation of alternate strategies for their development in the USA through the year 2020. By direction from NSF, ''advanced'' reactors were defined as those which met the dual requirements of (1) offering a significant improvement in fissile fuel utilization as compared to light-water reactors and (2) currently receiving U.S. Government funding. (A detailed study of the LMFBR was specifically excluded, but cursory baseline data were obtained from ERDA sources.) Included initially were the High-Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Reactor (GCFR), Molten Salt Reactor (MSR), and Light-Water Breeder Reactor (LWBR). Subsequently, the CANDU Heavy Water Reactor (HWR) was included for comparison due to increased interest in its potential. This volume presents the reasoning process and analytical methods utilized to arrive at the conclusions for the overall study

Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

International Nuclear Information System (INIS)

Bozic, M.; Zagar, T.; Ravnik, M.

2001-01-01

The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

The integral fast reactor

International Nuclear Information System (INIS)

Till, C.E.

1987-01-01

On April 3rd, 1986, two demonstrations of the inherent capability of sodium-cooled fast reactors to survive unprotected loss of cooling accidents were carried out on the experimental sodium-cooled power reactor, EBR-II, on the Idaho site of Argonne National Laboratory. Transients potentially of the most serious kind, one an unprotected loss of flow, the other an unprotected loss of heat sink, both initiated from full power. In both cases the reactor quietly shut itself down, without damage of any kind. These tests were a part of the on-going development program at Argonne to develop an advanced reactor with significant new inherent safety characteristics. Called the integral fast reactor, or IFR, the basic thrust is to develop everything that is needed for a complete nuclear power system - reactor, closed fuel cycle, and waste processing - as a single optimized entity, and, for simplicity in concept, as an integral part of a single plant. The particular selection of reactor materials emphasizes inherent safety characteristics also makes possible a simplified close fuel cycle and waste process improvements. The paper describes the IFR concept, the inherent safety, tests, and status of IFR development today

RA reactor safety analysis, Part II - Accident analysis; Analiza sigurnosti rada Reaktora RA I-III, Deo II - Analiza akcidenta

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Radanovic, Lj; Milovanovic, M; Afgan, N; Kulundzic, P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

This part of the RA reactor safety analysis includes analysis of possible accidents caused by failures of the reactor devices and errors during reactor operation. Two types of accidents are analyzed: accidents resulting from uncontrolled reactivity increase, and accidents caused by interruption of cooling.

Design of a reactor inlet temperature controller for EBR-2 using state feedback

International Nuclear Information System (INIS)

Vilim, R.B.; Planchon, H.P.

1990-01-01

A new reactor inlet temperature controller for pool type liquid-metal reactors has been developed and will be tested in EBR-II. The controller makes use of modern control techniques to take into account stratification and mixing in the cold pool during normal operation. Secondary flowrate is varied so that the reactor inlet temperature tracks a setpoint while reactor outlet temperature, primary flowrate and secondary cold leg temperature are treated as exogenous disturbances and are free to vary. A disturbance rejection technique minimizes the effect of these disturbances on inlet temperature. A linear quadratic regulator improves inlet temperature response. Tests in EBR-II will provide experimental data for assessing the performance improvements that modern control can produce over the existing EBR-II analog inlet temperature controller. 10 refs., 8 figs

Physics of Fast and Intermediate Reactors. V. I. Proceedings of the Seminar on the Physics of Fast and Intermediate Reactors. V. I

Energy Technology Data Exchange (ETDEWEB)

NONE

1962-03-15

in all cases that of heir presentation during the Seminar. Changes have been made where it was considered that these would enhance the usefulness of these volumes as reference books. The subject grouping adopted is given below. Volume I - I. Neutron Physics: I.1. Data requirements, I.2. Cross-section measurements, I.3. Fission properties, I.4. Nuclear theory, I.5. Multi-group cross-sections; II. Integral Experiments: II.1. Critical experiments, II.2. Other integral experiments, II.3. Theoretical correlations; Volume II - III. Reactor Theory: III.1. Calculation methods, III.2. Effects of cross-section errors, III.3. Reactivity effects, III.4. Long-term effects, III.5. Reactor concept studies; Volume III - IV. Reactor Dynamics: IV.1. Kinetics, IV.2. Stability, IV.3. Doppler effect, IV.4. Safety problems; V. Physics of Specific Reactors.

The new German neutron source FRM-II

International Nuclear Information System (INIS)

Nuding, M.; Axmann, A.; Boening, K.

2002-01-01

The construction of a new high-flux research reactor, the FRM-II is finished. This new reactor shall replace the existing FRM, which has been operated very successfully for about 43 years. The report at first presents the main applications of the FRM-II and its core and plant design. After that a description of the tests performed during the licensing procedure is given. At the end some current topics are discussed and an outlook on the time schedule is presented [ru

Tritium management in fusion reactors

International Nuclear Information System (INIS)

Galloway, T.R.

1978-05-01

This is a review paper covering the key environmental and safety issues and how they have been handled in the various magnetic and inertial confinement concepts and reference designs. The issues treated include: tritium accident analyses, tritium process control, occupational safety, HTO formation rate from the gas-phase, disposal of tritium contaminated wastes, and environmental impact--each covering the Joint European Tokamak (J.E.T. experiment), Tokamak Fusion Test Reactor (TFTR), Russian T-20, The Next Step (TNS) designs by Westinghouse/ORNL and General Atomic/ANL, the ANL and ORNL EPR's, the G.A. Doublet Demonstration Reactor, the Italian Fintor-D and the ORNL Demo Studies. There are also the following full scale plant reference designs: UWMAK-III, LASL's Theta Pinch Reactor Design (RTPR), Mirror Fusion Reactor (MFR), Tandem Mirror Reactor (TMR), and the Mirror Hybrid Reactor (MHR). There are four laser device breakeven experiments, SHIVA-NOVA, LLL reference designs, ORNL Laser Fusion power plant, the German ''Saturn,'' and LLL's Laser Fusion EPR I and II

Simulation development for TRIGA reactor

International Nuclear Information System (INIS)

Handoyo, D.

1997-01-01

A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable

The research reactor TRIGA Heidelberg II

International Nuclear Information System (INIS)

Maier-Borst, W.; Krauss, O.

1988-01-01

The reactor is in operation since the beginning of 1978. On the base of the working experience gathered during that time employing the TRIGA in biomedical research, especially the irradiation units have been extended or newly developed. Several TRIGA users have reported difficulties in using the rotary irradiation system. It became obvious that the alternatives to the original Lazy Susan are not commonly known. In this report, the open rotary system fed by a hydraulic rabbit system, which has proved successful in this form during the past ten years is presented

Surveillance application using patten recognition software at the EBR-II Reactor Facility

International Nuclear Information System (INIS)

Olson, D.L.

1992-01-01

The System State Analyzer (SSA) is a software based pattern recognition system. For the past several year this system has been used at Argonne National Laboratory's Experimental Breeder Reactor 2 (EBR-2) reactor for detection of degradation and other abnormalities in plant systems. Currently there are two versions of the SSA being used at EBR-2. One version of SSA is used for daily surveillance and trending of the reactor delta-T and startups of the reactor. Another version of the SSA is the QSSA which is used to monitor individual systems of the reactor such as the Secondary Sodium System, Secondary Sodium Pumps, and Steam Generator. This system has been able to detect problems such as signals being affected by temperature variations due to a failing temperature controller

Germanismen im epilinguistischen Diskurs über die Zagreber Umgangssprache und in der Reflexion über den eigenen Sprachgebrauch

Directory of Open Access Journals (Sweden)

Goranka Rocco

2014-03-01

Full Text Available Abstract – This paper sheds light on the epilinguistic discourse on germanisms in themetropolitan dialect of Zagreb. The empirical study is based on two methods: 1 the qualitative analysis of the internet forum discussion on what the dialectof Zagreb is, what the germanisms "purger" and "purgerski" mean, to what extent germanisms and elements from other languages should beused and why they are used; 2 the written questionnaire assessing theself-perception of the use of germanisms. The results of both inquiries arecompared and contrasted in order to gain a more complete picture of languageattitudes and language use in a given situation and the way of expressing themin the epilinguistic discourse.Keywords:epilinguisticdiscourse, germanisms, self-perception, language attitudes, language useAbstract –  Die Studie wendet sich dem epilinguistischen Diskursüber die Germanismen in der Stadtsprache von Zagreb zu. Sie setzt sich zumZiel, anhand der qualitätiven Analyse der Internetforum-Beiträge und der schriftlichenBefragung herauszufinden, um welche Themen der epilinguistische Diskurs überdie Germanismen kreist, wie lebendig diese in der Sprache zweier verschiedenenGenerationen der gebürtiger Zagreber sind, inweiweit die Sprecher Veränderungenim eigenen Sprachgebruch wahrnehmen, ob und welche Ausdrücke von den ihnen alstypisch für die Zagreber Umgangssprache (purgerski betrachtet werden und,damit verbunden, wie die Bedeutung der auf die Stadtsprache und -mentalitätbezogenen Germanismen „purger" und "purgerski" definiert wird.Schlüsselwörter:epilinguistischerDiskurs, Germanismus, Selbstwahrnehmung, Spracheinstellungen, Sprachgebrauch.

Research reactor's role in Korea

International Nuclear Information System (INIS)

Choi, C-O.

1995-01-01

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960's in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs

PASC-1, Petten AMPX-II/SCALE-3 Code System for Reactor Neutronics Calculation

International Nuclear Information System (INIS)

Yaoqing, W.; Oppe, J.; Haas, J.B.M. de; Gruppelaar, H.; Slobben, J.

1995-01-01

1 - Description of program or function: The Petten AMPX-II/SCALE-3 Code System PASC-1 is a reactor neutronics calculation programme system consisting of well known IBM-oriented codes, that have been translated into FORTRAN-77, for calculations on a CDC-CYBER computer. Thus, the portability of these codes has been increased. In this system, some AMPX-II and SCALE-3 modules, the one-dimensional transport code ANISN and the 1 to 3-dimensional diffusion code CITATION are linked together on the CDC-CYBER/855 computer. The new cell code XSDRNPM-S and the old XSDRN code are included in the system. Starting from an AMPX fine group library up to CITATION, calculations can be performed for each individual module. Existing AMPX master interface format libraries, such as CSRL-IV, JEF-1, IRI and SCALE-45, and the old XSDRN-formatted libraries such as the COBB library can be used for the calculations. The code system contains the following modules and codes at present: AIM, AJAX, MALOCS, NITAWL-S, REVERT-I, ICE-2, CONVERT, JUAN, OCTAGN, XSDRNPM-S, XSDRN, ANISN and CITATION. The system will be extended with other SCALE modules and transport codes. 2 - Method of solution: The PASC-1 system is based on AMPX-II/SCALE-3 modules. Except for some SCALE-3 modules taken from the SCALIAS package, the original AMPX-II modules were IBM versions written in FORTRAN IV. These modules have been translated into CDC FORTRAN V. In order to test these modules and link them with some codes, some of the sample problem calculations have been performed for the whole PASC-1 system. During these calculations, some FORTRAN-77 errors were found in MALOCS, REVERT, CONVERT and some subroutines of SUBLIB (FORTRAN-77 subroutine library). These errors have been corrected. Because many corrections were made for the REVERT module, it is renamed as REVERT-I (improved version of REVERT). After these corrections, the whole system is running on a CDC-CYBER Computer (NOS-BE operating system). 3 - Restrictions on the

Component failure data base of TRIGA reactors

International Nuclear Information System (INIS)

Djuricic, M.

2004-10-01

This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

Status of advanced technology and design for water cooled reactors: Light water reactors

International Nuclear Information System (INIS)

1988-10-01

Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

Dynamic simulation of the air-cooled decay heat removal system of the German KNK-II experimental breeder reactor

International Nuclear Information System (INIS)

Schubert, B.K.

1984-07-01

A Dump Heat Exchanger and associated feedback control system models for decay heat removal in the German KNK-II experimental fast breeder reactor are presented. The purpose of the controller is to minimize temperature variations in the circuits and, hence, to prevent thermal shocks in the structures. The basic models for the DHX include the sodium-air thermodynamics and hydraulics, as well as a control system. Valve control models for the primary and intermediate sodium flow regulation during post shutdown conditions are also presented. These models have been interfaced with the SSC-L code. Typical results of sample transients are discussed

Non-destructive material investigation with thermal neutrons at the TRIGA Mark II reactor in Vienna

International Nuclear Information System (INIS)

Bastuerk, M.; Boeck, H.; Zamani, B.; Zawisky, M.; Rauch, H.

2004-01-01

Neutron tomography providing 3D information about interior of an object is a very efficient tool to visualize inner defects of the materials, non-destructively. In this study, some applications of neutron tomography in different fields such as geology, aerospace, civil engineering and archaeology were presented. Distribution of minerals in pumice and rock samples, visualization of inner defects within a new developed titan aluminum turbine blade, and distribution of silica gel as an important impregnating agent in construction and restoration of buildings were investigated. The measurements of tomography projections taken in the 0 to 180 o angle were performed with a thermal neutron flux of 10 5 at the TRIGA Mark II research reactor in Vienna, and the common filtered back projection method was used for the 3D image reconstruction. (author)

IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

International Nuclear Information System (INIS)

Sartori, Enrico; Blair Briggs, J.

2008-01-01

experimental series that were performed at 17 different reactor facilities. The Handbook is organized in a manner that allows easy inclusion of additional evaluations, as they become available. Additional evaluations are in progress and will be added to the handbook periodically. Content: FUND - Fundamental; GCR - Gas Cooled (Thermal) Reactor; HWR - Heavy Water Moderated Reactor; LMFR - Liquid Metal Fast Reactor; LWR - Light Water Moderated Reactor; PWR - Pressurized Water Reactor; VVER - VVER Reactor; Evaluations published as drafts 2 - Related Information: International Criticality Safety Benchmark Evaluation Project (ICSBEP); IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments; IRPHE-JAPAN, Reactor Physics Experiments carried out in Japan ; IRPHE/JOYO MK-II, JOYO MK-II core management and characteristics database ; IRPhE/RRR-SEG, Reactor Physics Experiments from Fast-Thermal Coupled Facility; IRPHE-SNEAK, KFK SNEAK Fast Reactor Experiments, Primary Documentation ; IRPhE/STEK, Reactor Physics Experiments from Fast-Thermal Coupled Facility ; IRPHE-ZEBRA, AEEW Fast Reactor Experiments, Primary Documentation ; IRPHE-DRAGON-DPR, OECD High Temperature Reactor Dragon Project, Primary Documents; IRPHE-ARCH-01, Archive of HTR Primary Documents ; IRPHE/AVR, AVR High Temperature Reactor Experience, Archival Documentation ; IRPHE-KNK-II-ARCHIVE, KNK-II fast reactor documents, power history and measured parameters; IRPhE/BERENICE, effective delayed neutron fraction measurements ; IRPhE-TAPIRO-ARCHIVE, fast neutron source reactor primary documents, reactor physics experiments. The International Handbook of Evaluated Reactor Physics Benchmark Experiments was prepared by a working party comprised of experienced reactor physics personnel from Belgium, Brazil, Canada, P.R. of China, Germany, Hungary, Japan, Republic of Korea, Russian Federation, Switzerland, United Kingdom, and the United States of America. The IRPhEP Handbook is available to authorised requesters from the

Measurements and calculations for determination of discharge of 41Ar from IFEs research reactor JEEP II at Kjeller, Norway

International Nuclear Information System (INIS)

Raaum, A.; Straelberg, E.

2003-01-01

41 Ar is formed by neutron irradiation of 40 Ar, which occurs naturally in air with a concentration of 9300 ppm. The discharge of 41 Ar from IFEs research reactor Jeep II is yearly reported to the Norwegian Radiation Protection Authority (NRPA). Until year 2000 the reported values were based on theoretical calculations of produced 41 Ar per operating hour of 6.8 GBq/h. During 2000 and 2001 the reactor was upgraded to increase the irradiation capacity and to meet the markets demand for irradiation of 5'Si-crystalls. After the upgrading, measurements and calculations were initiated to determine the new discharge rate for 41 Ar. During reactor operation an approximately constant discharge of 41 Ar is expected, mainly due to irradiation of air in open beam channels. In addition 41 Ar is released from irradiation pockets when they are opened to transfer samples in and out during reactor stop. The new value for discharge rate was determined from measurements of air samples from the discharge channel during operation and theoretical calculations of the release from the irradiation pockets. The new discharge rate was determined to 5.9 ± 0.5 GBq/h, which is a small reduction compared to the former value of 6.8 GBq/h. A small reduction in discharge rate was expected because the number of air-filled irradiation pockets was reduced after the upgrading. In a normal year the discharge of 41 Ar will be about 2 % of the Institutes discharge permission. (orig.)

Keim, Karl-Dieter, Das Fenster zum Raum : Traktat über die Erforschung sozialräumlicher Transformation / [rezensiert von] Helmut Klüter

OpenAIRE

Klüter, Helmut

2009-01-01

Rezensiertes Werk: Keim, Karl-Dieter: Das Fenster zum Raum : Traktat über die Erforschung sozialräumlicher Transformation / Karl-Dieter Keim. - Opladen : Leske + Budrich, 2003. - 171 S. : graph. Darst. - (Lehrtexte : Sozialwissenschaften) ISBN 3-8100-4050-9

SPLOSH II: A dynamics programme for nuclear - thermal - hydrodynamic behaviour of water-cooled reactors

International Nuclear Information System (INIS)

Moxon, D.

1966-01-01

A dynamics code is described that solves the two-group neutron diffusion equations simultaneously with the thermal and the hydraulic equations for an average channel of a water-cooled reactor. Other reactor channels can be represented as 'slaves', which have no feedback to the average channel. The fission power at any axial station in a slave channel is related to that in the average by prescribed time-dependent factors, and the hydraulic flow is determined from pressure-drop requirements dictated by the performance of the average channel. A finite difference model of the fuel element and can represents the behaviour of the fuel temperatures and surface heat flux. The representation of the hydraulic circuit has been made sufficiently general that the code is applicable to B.W.R., P.W.R. and pressure tube reactor designs. The code can be used to study transients resulting from imposed time variations in coolant flow, inlet enthalpy, system pressure, electrical torque supplied to the circulating pumps, (or alternatively, the angular velocity of the pump rotors,) moderator height, frictional resistances simulating blockages and control rod and fuel element insertions. The harmonic response can be obtained by injecting sinusoidal time variations until the starting transient has been damped out. Output includes axial distributions of the neutron fluxes, heat flux, coolant density and temperature, burn-but margin, and the fuel and can temperatures in both the average and the slave channels. The code was originally written in FORTRAN II for use on the IBM 7090. Computing times vary greatly with the problem and the desired accuracy but experience has shown that a computing time which is slower than real time by a factor thirty is adequate for a wide range of cases. The code has recently been converted to S2 and EGTRAN for use on the IBM 7030 and the English Electric Leo Marconi KDF 9 computers. (author)

Echt und modern? Diskurse über Männlichkeit

Directory of Open Access Journals (Sweden)

Florian Kahofer

2014-09-01

Full Text Available Der vorliegende Artikel befasst sich mit Repräsentationen von Männlichkeit im österreichischen Lifestyle-Magazin für Männer Wiener. Durch eine korpusbasierte Diskursanalyse wird ein umfassendes Korpus aller Ausgaben des Wieners von Anfang 2002 bis Ende 2012 untersucht. Auf theoretischer Ebene wird dabei eine Verbindung von Kritischer Männlichkeitsforschung (KMF und Feministisch Kritischer Diskursanalyse (FCDA unternommen. Es werden aktuelle Veröffentlichungen zu Kritischer Diskursanalyse und Männlichkeit vorgestellt und diskutiert. Durch den Einsatz einer Konkordanzsoftware werden Konkordanzen des Nomens MANN analysiert. Diese werden allerdings insofern eingeschränkt betrachtet, als nur Nominationen in der Form der häufigsten Adjektiv-Konstruktionen untersucht werden. Die Ergebnisse zeigen, dass neben den Diskursen über Krise und Neue Männlichkeit Themen wie Alter, Körper oder Beziehung auftauchen. Männlichkeit wird als ambivalent und vielfältig dargestellt. Deutungskämpfe um Männlichkeit lassen sich ausmachen.

In-vessel inspection before head removal: TMI II: Phase I. (Conceptual development)

International Nuclear Information System (INIS)

Calloway, N.E.; Greenlee, D.W.; Lawrence, G.R.; Paglia, A.L.; Piatt, T.D.; Tucker, B.A.

1981-08-01

The objective of the task is to provide for an internal inspection of the reactor vessel and the fuel assemblies prior to reactor vessel head removal. Because the degree of damage to equipment and fuel in the TMI-II reactor is not precisely known, it is important that as much information as possible be obtained on present conditions inside the reactor. This information will serve to benchmark the various analyses already completed or underway and will also guide the development of programs to obtain more information on the TMI-II core damage. In addition, the early look will provide data for planning the reactor disassembly program

Description of the advanced gas cooled type of reactor (AGR)

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

1996-11-01

The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

Description of the advanced gas cooled type of reactor (AGR)

International Nuclear Information System (INIS)

Nonboel, E.

1996-11-01

The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

EMI and BER/PER Analysis of Wi-Fi and Bluetooth Communication for CRIP Platform

DEFF Research Database (Denmark)

Ravelo, Blaise; Cabral, Jorge; Wagner, Stefan Rahr

This paper is devoted to the characterization of the electromagnetic interference (EMI) effects on wireless communication devices for telemedicine healthcare application. The proposed wireless system operates at the 2.4GHz ISM band with a hardware interface communicating by Wi-Fi and BLE (Bluetooth...... path loss model is investigated for the indoor RF link budget. The quality of the EM coverage and SNR and SINR on QoS with transmitted data BER and PER is analysed. It was examined that the QoS can be degraded according to the distance between the Wi-Fi and BLE device sensitivity and also...

Experimental and numerical determination of the dynamic properties of the reactor building of Atucha II NPP

International Nuclear Information System (INIS)

Ceballos, M.A.; Car, E.J.; Prato, T.A.; Prato, C.A.; Alvarez, L.M.; Godoy, A.R.

1995-01-01

Determination of the dynamic properties of the reactor building of Atucha II NPP is carried out in order to: i) Obtain valuable information for seismic qualification of the plant, and ii) Assess some procedures for testing and analysis that are used in the process of seismic evaluation of existing nuclear facilities founded on Quaternary soil deposits. Both steady state and impulsive dynamic tests were performed but attention is centered here in tile techniques used to determine natural frequencies and modal damping ratios with impulsive tests. Numerical analyses were performed by means of a 3-D model model of the superstructure together with foundation stiffness coefficients derived in a separate paper from steady state vibration tests, and also from analysis with a 2-D F.E. model of the soil layers capable of approximating the 3-D features of the problem. The computed foundation stiffness coefficients are compared both with those obtained from the tests and from an axisymmetric F.E. model; results indicate that foundation stiffness coefficients calculated with F.E. models with soil parameters given by laboratory tests performed on cored samples are significantly lower than those given by the steady state vibration tests. (author)

Fast Reactor Physics. Vol. II. Proceedings of a Symposium on Fast Reactor Physics and Related Safety Problems

International Nuclear Information System (INIS)

1968-01-01

Proceedings of a Symposium organized by the IAEA and held in Karlsruhe, 30 October - 3 November 1967. The meeting was attended by 183 scientists from 23 countries and three international organizations. Contents: (Vol.1) Review of national programmes (5 papers); Nuclear data for fast reactors (12 papers); Experimental methods (3 papers); Zoned systems (7 papers); Kinetics (7 papers). (Vol.11) Fast critical experiments (8 papers); Heterogeneity in fast critical experiments (5 papers); Fast power reactors (13 papers); Fast pulsed reactors (3 papers); Panel discussion. Each paper is in its original language (50 English, 11 French and 3 Russian) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

International Nuclear Information System (INIS)

1976-08-01

This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

Breazeale Reactor Modernization Program

International Nuclear Information System (INIS)

Davison, C. C.

2003-01-01

The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future

Recent operating experiences and programs at EBR-II

International Nuclear Information System (INIS)

Lentz, G.L.

1984-01-01

Experimental Breeder Reactor No. II (EBR-II) is a pool-type, unmoderated, sodium-cooled reactor with a design power of 62.5 MWt and an electrical generation capability of 20 MW. It has been operated by Argonne National Laboratory for the US government for almost 20 years. During that time, it has operated safely and has demonstrated stable operating characteristics, high availability, and excellent performance of its sodium components. The 20 years of operating experience of EBR-II is a valuable resource to the nuclear community for the development and design of future LMFBR's. Since past operating experience has been extensively reported, this report will focus on recent programs and events

RTNS-II utilization plan

Energy Technology Data Exchange (ETDEWEB)

Zwilsky, Klaus M.

1978-09-01

This plan describes a general program for the effective utilization of this resource by the fusion materials community. Because its flux is low relative to levels expected in commercial fusion reactors, the RINS-II is not expected to produce data of direct engineering significance (with some exceptions). Rather, it will be used chiefly to aid in the development of models of high energy neutron effects. Such models are needed in projecting engineering data obtained in high flux fission reactors to the fusion environment. Fission reactors, because of their relatively soft neutron spectra, cannot produce the high ratio of transmutations to displacements (except in an important special case) or the high energy recoil atoms appropriate to fusion reactors utilizing the D-T reaction.

Remote, under-sodium fuel handling experience at EBR-II

International Nuclear Information System (INIS)

King, R.W.; Planchon, H.P.

1995-01-01

The EBR-II is a pool-type design; the reactor fuel handling components and entire primary-sodium coolant system are submerged in the primary tank, which is 26 feet in diameter, 26 feet high, and contains 86,000 gallons of sodium. Since the reactor is submerged in sodium, fuel handling operations must be performed blind, making exact positioning and precision control of the fuel handling system components essential. EBR-II operated for 30 years, and the fuel handling system has performed approximately 25,000 fuel transfer operations in that time. Due to termination of the IFR program, EBR-II was shut down on September 30, 1994. In preparation for decommissioning, all fuel in the reactor will be transferred out of EBR-II to interim storage. This intensive fuel handling campaign will last approximately two years, and the number of transfers will be equivalent to the fuel handling done over about nine years of normal reactor operation. With this demand on the system, system reliability will be extremely important. Because of this increased demand, and considering that the system has been operating for about 32 years, system upgrades to increase reliability and efficiency are proceeding. Upgrades to the system to install new digital, solid state controls, and to take advantage of new visualization technology, are underway. Future reactor designs using liquid metal coolant will be able to incorporate imaging technology now being investigated, such as ultraviolet laser imaging and ultrasonic imaging

Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

Science.gov (United States)

Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

2014-02-01

The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. Copyright © 2013 Elsevier Ltd. All rights reserved.

Development of technologies for nuclear reactors of small and medium sized

International Nuclear Information System (INIS)

2011-08-01

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Experience with advanced driver fuels in EBR-II

International Nuclear Information System (INIS)

Lahm, C.E.; Koenig, J.F.; Pahl, R.G.; Porter, D.L.; Crawford, D.C.

1992-01-01

The Experimental Breeder Reactor II (EBR-II) is a complete nuclear power plant, incorporating a pool-type liquid-metal reactor (LMR) with a fuel-power thermal output of 62.5 MW and an electrical output of 20 MW. Initial criticality was in 1961, utilizing a metallic driver fuel design called the Mark-I. The fuel design has evolved over the last 30 yr, and significant progress has been made on improving performance. The first major innovations were incorporated into the Mark-II design, and burnup then increased dramatically. This design performed successfully, and fuel element lifetime was limited by subassembly hardware performance rather than the fuel element itself. Transient performance of the fuel was also acceptable and demonstrated the ability of EBR-II to survive severe upsets such as a loss of flow without scram. In the mid 1980s, with renewed interest in metallic fuels and Argonne's integral fast reactor (IFR) concept, the Mark-II design was used as the basis for new designs, the Mark-III and Mark-IV. In 1987, the Mark-III design began qualification testing to become a driver fuel for EBR-II. This was followed in 1989 by the Mark-IIIA and Mark-IV designs. The next fuel design, the Mark-V, is being planned to demonstrate the utilization of recycled fuel. The fuel cycle facility attached to EBR-II is being refurbished to produce pyroprocessed recycled fuel as part of the demonstration of the IFR

The new German neutron source FRM-II

International Nuclear Information System (INIS)

Nuding, M.

2003-01-01

The 'Technische Universitaet Muenchen' has built a new high-flux research reactor, the 'Forschungsreaktor Muenchen'-II. This new reactor will replace the 'Forschungsreaktor Muenchen' which has been operated very successfully for about 43 years. The 'Forschungsreaktor Muenchen'-II has been developed with first priority for beam-tube experiments, but it will also provide possibilities for irradiation experiments or isotope production. The reactor was designed to obtain a high and spectrally pure thermal neutron flux is available in a large volume outside of the core, where it is accessible for experimental use. In addition to beam-tubes which will end in the thermal neutron field there will be beam-tubes that will provide - with the help of 'spectrum shifters' -cold; hot and fast neutrons. Even through the thermal power of the 'Forschungsreaktor Muenchen'-II was limited to 20 MW an unperturbed maximum thermal neutron flux of about 8 x 10 14 cm -2 s -1 will be reached. Because of its 'compact-core-concept' the 'Forschungsreaktor Muenchen'-II will have the best flux-to-power-ratio worldwide: The fuel element and its highly enriched U 3 Si 2 -Al-fuel were tested during the licensing procedure of the 'Forschungsreaktor Muenchen'-II. Within the scope of this 'hydraulic test' the stability and the vibration behavior of the fuel plates as well as the long-tem behavior of the fuel element were investigated (Authors)

Experimental investigation into fast pyrolysis of biomass using an entrained flow reactor

Science.gov (United States)

Bohn, M.; Benham, C.

1981-02-01

Pyrolysis experiments were performed with steam as a carrier gas and two different feedstocks - wheat straw and powdered material derived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 7000 to 1400 C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II are due to the presence of plastics in the feedstock.

Zn(II Removal from Wastewater by Electrocoagulation/Flotation Method using New Configuration of a Split-Plate Airlift Electrochemical Reactor

Directory of Open Access Journals (Sweden)

Saad H. Ammar

2018-01-01

Full Text Available In this paper, split-plate airlift electrochemical reactor as an apparatus with new configuration for wastewater treatment was provided. Two aluminum plates were fixed inside the reactor and present two functions; first it works as split plates for internal loop generation of the airlift system (the zone between the two plates acts as riser while the other two zones act as downcomer and second it works as two electrodes for electrocoagulation process. Simulated wastewater contaminated with zinc ions was used to test the performance of this apparatus for zinc removal by studying the effect of different experimental variables such as initial concentration of zinc (50-800 ppm, electrical current density (2.67-21.4 mA/cm2, initial pH (3-11, air flowrate (12-50 LPH, and implicitly the electrocoagulation time. The results have shown the applicability of this split-plate airlift reactor as electrocoagulation cell in the treatment of wastewater such as wastewater containing Zink ions. The Zink removal percent was shown to increase upon increasing the current density and the electrolysis time. Also best removal percent was achieved in the initial pH range between 7 and 9. The minimum electrocoagulation time required for removal of ≥ 90% of Zn(II decreases from 90 to 22 min when operating current density increases from 2.67 to 21.4 mA/cm2.

1 kWe sodium borohydride hydrogen generation system Part II: Reactor modeling

OpenAIRE

Zhang, Jinsong; Zheng, Yuan; Gore, Jay P; Mudawar, Issam; Fisher, Timothy

2007-01-01

Sodium borohydride (NaBH4) hydrogen storage systems offer many advantages for hydrogen storage applications. The physical processes inside a NaBH4 packed bed reactor involve multi-component and multi-phase flow and multi-mode heat and mass transfer. These processes are also coupled with reaction kinetics. To guide reactor design and optimization, a reactor model involving all of these processes is desired. A onedimensional numerical model in conjunction with the assumption of homogeneous cata...

RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

International Nuclear Information System (INIS)

2015-01-01

The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

Reliability and extended-life potential of EBR-II

International Nuclear Information System (INIS)

King, R.W.

1985-01-01

Although the longlife potential of liquid-metal-cooled reactors (LMRs) has been only partially demonstrated, many factors point to the potential for exceptionally long life. EBR-II has the opportunity to become the first LMR to achieve an operational lifetime of 30 years or more. In 1984 a study of the extended-life potential of EBR-II identified the factors that contribute to the continued successful operation of EBR-II as a power reactor and experimental facility. Also identified were factors that could cause disruptions in the useful life of the facility. Although no factors were found that would inherently limit the life of EBR-II, measures were identified that could help ensure continued plant availability. These measures include the implementation of more effective surveillance, diagnostic, and control systems to complement the inherent safety and reliability features of EBR-II. An operating lifetime of well beyond 30 years is certainly feasible

Modeling the PUSPATI TRIGA Reactor using MCNP code

International Nuclear Information System (INIS)

Mohamad Hairie Rabir; Mark Dennis Usang; Naim Syauqi Hamzah; Julia Abdul Karim; Mohd Amin Sharifuldin Salleh

2012-01-01

The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)

Swelling and tensile properties of EBR-II-irradiated tantalum alloys for space reactor applications

International Nuclear Information System (INIS)

Grossbeck, M.L.; Wiffen, F.W.

1985-01-01

The tantalum alloys T-111, ASTAR-811C, Ta-10 W, and unalloyed tantalum were examined following EBR-II irradiation to a fluence of 1.7 x 10 26 neutrons/m 2 (E > 0.1 MeV) at temperatures from 650 to 950 K. Swelling was found to be negligible for all alloys; only tantalum was found to exhibit swelling, 0.36%. Tensile testing revealed that irradiated T-111 and Ta-10 W are susceptible to plastic instability, but ASTAR-811C and tantalum were not. The tensile properties of ASTAR-811C appeared adequate for current SP-100 space nuclear reactor designs. Irradiated, oxygen-doped T-111 exhibited no plastic deformation, and the abrupt failure was intergranular in nature. The absence of plastic instability in ASTAR-811C is encouraging for alloys containing carbide precipitates. These fine precipitates might prevent dislocation channeling, which leads to plastic instability in many bcc metals after irradiation. 10 refs., 13 figs., 8 tabs

Safety design features for current UK advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

1981-01-15

The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

Safety design features for current UK advanced gas-cooled reactors

International Nuclear Information System (INIS)

Yellowlees, J.M.; Cobb, E.C.

1981-01-01

The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

Self Assessment for the Safety of Research Reactor in Indonesia

International Nuclear Information System (INIS)

Melani, Ai; Chang, Soon Heung

2008-01-01

At the present Indonesia has no nuclear power plant in operation yet, although it is expected that the first nuclear power plant will be operated and commercially available in around the year of 2016 to 2017 in Muria Peninsula. National Nuclear Energy Agency (BATAN) has three research reactor; which are: Reactor Triga Mark II at Bandung, Reactor Kartini at Yogyakarta and Reactor Serbaguna (Multi Purpose Reactor) at Serpong. The Code of Conduct on the Safety of Research Reactors establishes 'best practice' guidelines for the licensing, construction and operation of research reactors. In this paper the author use the requirement in code of conduct to review the safety of research reactor in Indonesia

Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Mohd, Shukri; Kassim, Razali; Mahmood, Zal Uyun [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia); Radiman, Shahidan

1998-10-01

The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One of the project involved the Small Angle Neutron Scattering (SANS). (author)

SIMMER-II analysis of transition-phase experiments

International Nuclear Information System (INIS)

Wehner, T.R.; Bell, C.R.

1985-01-01

Analyses of Los Alamos transition-phase experiments with the SIMMER-II computer code are reported. These transient boilup experiments simulated the recriticality-induced transient motion of a boiling pool of molten fuel, molten steel and steel vapor, within a subassembly duct in a liquid-metal fast breeder reactor during the transition phase of a core-disruptive accident. The two purposes of these experiments were to explore and reach a better understanding of fast reactor safety issues, and to provide data for SIMMER-II verification. Experimental data, consisting of four pressure traces and a high-speed movie, were recorded for four sets of initial conditions. For three of the four cases, SIMMER-II-calculated pressures compared reasonably well with the experimental pressures. After a modification to SIMMER-II's liquid-vapor drag correlation, the comparison for the fourth case was reasonable also. 12 refs., 4 figs

Burn-up TRIGA Mark II benchmark experiment

International Nuclear Information System (INIS)

Persic, A.; Ravnik, M.; Zagar, T.

1998-01-01

Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

Cladding failure margins for metallic fuel in the integral fast reactor

International Nuclear Information System (INIS)

Bauer, T.H.; Fenske, G.R.; Kramer, J.M.

1987-01-01

The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. (orig./GL)

Contexto dos berçários e um programa de intervenção no desenvolvimento de bebés

Directory of Open Access Journals (Sweden)

Carla Skilhan Almeida

2013-12-01

Full Text Available O objetivo deste estudo foi investigar o contexto de berçários que atendesse a famílias de baixa renda e verificar o impacto de uma intervenção cognitivo-motora no desenvolvimento de bebés quanto à motricidade ampla, motricidade fina, linguagem e interação. Participaram do estudo 40 bebés, entre seis e oito meses, provenientes de dez berçários para famílias de baixa renda. As escolas foram agrupadas quanto à qualidade dos cuidados oferecidos; os bebés foram estratificados por gênero e aleatoriamente distribuídos em grupos interventivo (n= 20 e controle (n= 20. Foi utilizada uma ficha de observação para verificar a rotina dos berçários. A motricidade ampla e fina, a linguagem e a interação dos bebés foram avaliadas no pré e no pós intervenção. O programa de intervenção propiciou experiências de perseguição visual, manipulação de brinquedos e controle postural. Observaram-se restrições no espaço físico e nas oportunidades para brincar e interagir. Educadores centravam sua atenção no cuidado com higiene e alimentação dos bebés. Bebés provenientes de escolas com contextos mais apropriadas ao desenvolvimento apresentaram desempenhos superiores; e, a intervenção repercutiu positivamente no desenvolvimento global dos participantes. Em conclusão, ações educativas e estratégias interventivas devem ser implementadas nas creches, priorizando o processo de desenvol­vimento infantil.

Fast reactor operation in the United States

International Nuclear Information System (INIS)

Smith, R.R.; Cissel, D.W.