Reactor System Design

International Nuclear Information System (INIS)

Chi, S. K.; Kim, G. K.; Yeo, J. W.

2006-08-01

SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

Performance of the LAr scintillation veto of GERDA Phase II

Energy Technology Data Exchange (ETDEWEB)

Wiesinger, Christoph [Technische Universitaet Muenchen, Physik Dept. E15, James-Franck-Strasse, 85748 Garching (Germany); Collaboration: GERDA-Collaboration

2015-07-01

Gerda is an experiment to search for the neutrinoless double beta decay in {sup 76}Ge. Results of Phase I have been published in summer 2013 and Gerda is upgraded to Phase II. To reach the aspired background index of ≤ 10{sup -3} cts/(keV.kg.yr) for Phase II active background-suppression techniques are applied, including an active liquid argon (LAr) veto. It has been demonstrated with the LArGe test facility that the detection of argon scintillation light can be used to effectively suppress background events in the germanium, which simultaneously deposit energy in the LAr. The light instrumentation consisting of photomultiplier tubes (PMT) and wavelength-shifting fibers connected to silicon multipliers (SiPM) has been installed in Gerda. In this talk the low background design of the LAr veto and its performance during the commissioning runs are reported.

Three-dimensional Imaging for Large LArTPCs

Energy Technology Data Exchange (ETDEWEB)

Chao, C. [Brookhaven National Lab. (BNL), Upton, NY (United States); Qian, X. [Brookhaven National Lab. (BNL), Upton, NY (United States); Viren, B. [Brookhaven National Lab. (BNL), Upton, NY (United States); Diwan, M. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2017-12-14

High-performance event reconstruction is critical for current and future massive liquid argon time projection chambers (LArTPCs) to realize their full scientic potential. LArTPCs with readout using wire planes provides a limited number of 2D projections. In general, without a pixel-type readout it is challenging to achieve unambiguous 3D event reconstruction. As a remedy, we present a novel 3D imaging method, Wire-Cell, which incorporates the charge and sparsity information in addition to the time and geometry through simple and robust mathematics.

Design of the cryogenic systems for the Near and Far LAr-TPC detectors of the Short-Baseline Neutrino program (SBN) at Fermilab

Energy Technology Data Exchange (ETDEWEB)

Geynisman, M. [Fermilab; Bremer, J. [CERN; Chalifour, M. [CERN; Delaney, M. [Fermilab; Dinnon, M. [Fermilab; Doubnik, R. [Fermilab; Hentschel, S. [Fermilab; Kim, M. J. [Fermilab; Montanari, C. [INFN, Pavia; Monatanari, D. [Fermilab; Nichols, T. [Fermilab; Norris, B. [Fermilab; Sarychev, M. [Fermilab; Schwartz, F. [Fermilab; Tillman, J. [Fermilab; Zuckerbrot, M. [Fermilab

2017-08-31

The Short-Baseline Neutrino (SBN) physics program at Fermilab and Neutrino Platform (NP) at CERN are part of the international Neutrino Program leading to the development of Long-Baseline Neutrino Facility/Deep Underground Neutrino Experiment (LBNF/DUNE) science project. The SBN program consisting of three Liquid Argon Time Projection Chamber (LAr-TPC) detectors positioned along the Booster Neutrino Beam (BNB) at Fermilab includes an existing detector known as MicroBooNE (170-ton LAr-TPC) plus two new experiments known as SBN’s Near Detector (SBND, ~260 tons) and SBN’s Far Detector (SBN-FD, ~760 tons). All three detectors have distinctly different design of their cryostats thus defining specific requirements for the cryogenic systems. Fermilab has already built two new facilities to house SBND and SBN-FD detectors. The cryogenic systems for these detectors are in various stages of design and construction with CERN and Fermilab being responsible for delivery of specific sub-systems. This contribution presents specific design requirements and typical implementation solutions for each sub-system of the SBND and SBN-FD cryogenic systems.

Performance of the LAr scintillation veto of Gerda Phase II

Energy Technology Data Exchange (ETDEWEB)

Wiesinger, Christoph [Physik-Department and Excellence Cluster Universe, Technische Universitaet Muenchen, James-Franck-Strasse, 85748 Garching (Germany); Collaboration: GERDA-Collaboration

2016-07-01

Gerda is an experiment to search for the neutrinoless double beta decay in {sup 76}Ge. Results of Phase I have been published in summer 2013 and Gerda has been upgraded to Phase II. To reach the aspired background index of ∝10{sup -3} cts/(keV.kg.yr) for Phase II active background-suppression techniques are applied, including an active liquid argon (LAr) veto. It has been demonstrated with the LArGe test facility that the detection of argon scintillation light can be used to effectively suppress background events in the germanium detectors, which simultaneously deposit energy in the LAr. The light instrumentation consisting of photomultiplier tubes (PMT) and wavelength-shifting fibers connected to silicon photomultipliers (SiPM) has been installed in Gerda. In this talk the low background design of the LAr veto and its performance during Phase II start-up is reported.

Design of the cryogenic systems for the Near and Far LAr-TPC detectors of the Short-Baseline Neutrino program (SBN) at Fermilab

CERN Document Server

Geynisman, M; Chalifour, M; Delaney, M; Dinnon, M; Doubnik, R; Hentschel, S; Kim, M J; Montanari, C; Montanari, D; Nichols, T; Norris, B; Sarychev, M; Schwartz, F; Tillman, J; Zuckerbrot, M

2017-01-01

The Short-Baseline Neutrino (SBN) physics program at Fermilab and Neutrino Platform (NP) at CERN are part of the international Neutrino Program leading to the development of Long-Baseline Neutrino Facility/Deep Underground Neutrino Experiment (LBNF/DUNE) science project. The SBN program consisting of three Liquid Argon Time Projection Chamber (LAr-TPC) detectors positioned along the Booster Neutrino Beam (BNB) at Fermilab includes an existing detector known as MicroBooNE (170-ton LAr-TPC) plus two new experiments known as SBN’s Near Detector (SBND, ~260 tons) and SBN’s Far Detector (SBN-FD, ~760 tons). All three detectors have distinctly different design of their cryostats thus defining specific requirements for the cryogenic systems. Fermilab has already built two new facilities to house SBND and SBN-FD detectors. The cryogenic systems for these detectors are in various stages of design and construction with CERN and Fermilab being responsible for delivery of specific sub-systems. This contribution prese...

LArSoft: toolkit for simulation, reconstruction and analysis of liquid argon TPC neutrino detectors

Science.gov (United States)

Snider, E. L.; Petrillo, G.

2017-10-01

LArSoft is a set of detector-independent software tools for the simulation, reconstruction and analysis of data from liquid argon (LAr) neutrino experiments The common features of LAr time projection chambers (TPCs) enable sharing of algorithm code across detectors of very different size and configuration. LArSoft is currently used in production simulation and reconstruction by the ArgoNeuT, DUNE, LArlAT, MicroBooNE, and SBND experiments. The software suite offers a wide selection of algorithms and utilities, including those for associated photo-detectors and the handling of auxiliary detectors outside the TPCs. Available algorithms cover the full range of simulation and reconstruction, from raw waveforms to high-level reconstructed objects, event topologies and classification. The common code within LArSoft is contributed by adopting experiments, which also provide detector-specific geometry descriptions, and code for the treatment of electronic signals. LArSoft is also a collaboration of experiments, Fermilab and associated software projects which cooperate in setting requirements, priorities, and schedules. In this talk, we outline the general architecture of the software and the interaction with external libraries and detector-specific code. We also describe the dynamics of LArSoft software development between the contributing experiments, the projects supporting the software infrastructure LArSoft relies on, and the core LArSoft support project.

ARIADNE, a Photographic LAr TPC at the CERN Neutrino Platform

CERN Document Server

Mavrokoridis, K; Nessi, M; Roberts, A; Smith, N A; Touramanis, C; CERN. Geneva. SPS and PS Experiments Committee; SPSC

2016-01-01

This letter of intent describes a novel and innovative two-phase LAr TPC with photographic capabilities as an attractive alternative readout method to the currently accepted segmented THGEMs which will require many thousands of charge readout channels for kton-scale two-phase TPCs. These colossal LAr TPCs will be used for the future long-baseline-neutrino-oscillation experiments. Optical readout also presents many other clear advantages over current readout techniques such as ease of scalability, upgrade, installation and maintenance, and cost effectiveness. This technology has already been demonstrated at the Liverpool LAr facility with the photographic capturing of cosmic muon tracks and single gammas using a 40-litre prototype. We have now secured ERC funding to develop this further with the ARIADNE programme. ARIADNE will be a 1-ton two-phase LAr TPC utilizing THGEM and EMCCD camera readouts in order to photograph interactions, allowing for track reconstruction and particle identification. We are request...

Cold front-end electronics and Ethernet-based DAQ systems for large LAr TPC readout

CERN Document Server

D.Autiero,; B.Carlus,; Y.Declais,; S.Gardien,; C.Girerd,; J.Marteau; H.Mathez

2010-01-01

Large LAr TPCs are among the most powerful detectors to address open problems in particle and astro-particle physics, such as CP violation in leptonic sector, neutrino properties and their astrophysical implications, proton decay search etc. The scale of such detectors implies severe constraints on their readout and DAQ system. We are carrying on a R&D in electronics on a complete readout chain including an ASIC located close to the collecting planes in the argon gas phase and a DAQ system based on smart Ethernet sensors implemented in a µTCA standard. The choice of the latter standard is motivated by the similarity in the constraints with those existing in Network Telecommunication Industry. We also developed a synchronization scheme developed from the IEEE1588 standard integrated by the use of the recovered clock from the Gigabit link

Reactor system

International Nuclear Information System (INIS)

Miyano, Hiroshi; Narabayashi, Naoshi.

1990-01-01

The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

Influence of LAR and VAR on Para-Aminopyridine Antimalarials Targetting Haematin in Chloroquine-Resistance.

Science.gov (United States)

Warhurst, David C; Craig, John C; Raheem, K Saki

2016-01-01

Antimalarial chloroquine (CQ) prevents haematin detoxication when CQ-base concentrates in the acidic digestive vacuole through protonation of its p-aminopyridine (pAP) basic aromatic nitrogen and sidechain diethyl-N. CQ export through the variant vacuolar membrane export channel, PFCRT, causes CQ-resistance in Plasmodium falciparum but 3-methyl CQ (sontochin SC), des-ethyl amodiaquine (DAQ) and bis 4-aminoquinoline piperaquine (PQ) are still active. This is determined by changes in drug accumulation ratios in parasite lipid (LAR) and in vacuolar water (VAR). Higher LAR may facilitate drug binding to and blocking PFCRT and also aid haematin in lipid to bind drug. LAR for CQ is only 8.3; VAR is 143,482. More hydrophobic SC has LAR 143; VAR remains 68,523. Similarly DAQ with a phenol substituent has LAR of 40.8, with VAR 89,366. In PQ, basicity of each pAP is reduced by distal piperazine N, allowing very high LAR of 973,492, retaining VAR of 104,378. In another bis quinoline, dichlorquinazine (DCQ), also active but clinically unsatisfactory, each pAP retains basicity, being insulated by a 2-carbon chain from a proximal nitrogen of the single linking piperazine. While LAR of 15,488 is still high, the lowest estimate of VAR approaches 4.9 million. DCQ may be expected to be very highly lysosomotropic and therefore potentially hepatotoxic. In 11 pAP antimalarials a quadratic relationship between logLAR and logResistance Index (RI) was confirmed, while log (LAR/VAR) vs logRI for 12 was linear. Both might be used to predict the utility of structural modifications.

Land-Acquisition and Resettlement (LAR Conflicts: A Perspective of Spatial Injustice of Urban Public Resources Allocation

Directory of Open Access Journals (Sweden)

Jinxia Zhu

2018-03-01

Full Text Available Land acquisition and resettlement (LAR is an important step in urban development. As one of the ‘externalities of development’, LAR conflicts have affected social stability and development in rural areas of China. With social conflict research shifting from value identity to resource allocation, few studies have examined the relationship between the spatial injustice of urban public resources and LAR conflict. To mitigate this research gap and formulate effective policies, this study aims to reinterpret the obstacles of LAR conflicts from the perspective of the spatial injustice of urban public facilities allocation in Hangzhou City by examining 195 administrative litigation cases. Spatial accessibility was used for estimating the spatial justice of urban public resources allocation. A classification and regression tree (CART model was applied to identify the advantage and disadvantage factors behind LAR conflict, and explored the logical and structural relationships among these factors. Results showed that a spatial mismatch between the spatial behavior preferences of human activity and the spatial injustice of urban public resources allocation had significantly accelerated LAR conflicts. When the spatial behavior preferences of human activity and spatial distribution of urban public resources correspond to each other pre- and after LAR, basic rights to social space are safeguarded and various groups can equitably share spatial resources. There are no conflicts. Conversely, respondents expressed a high level of dissatisfaction in comparison to their pre-LAR conditions, and LAR conflict undeniably occurs. This approach also proposes some good LAR policies by regulating the spatial injustice of urban public resources allocation associated with LAR with the aim of long-term urban sustainable development for Hangzhou.

File list: His.Lar.50.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

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File list: His.Lar.20.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.20.AllAg.Larval_brain dm3 Histone Larvae Larval brain SRX1426943,SRX1426945... http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.20.AllAg.Larval_brain.bed ...

LArGe. A liquid argon scintillation veto for GERDA

International Nuclear Information System (INIS)

Heisel, Mark

2011-01-01

LArGe is a GERDA low-background test facility to study novel background suppression methods in a low-background environment, for possible applications in the GERDA experiment. GERDA searches for the neutrinoless double-beta decay in 76 Ge, by operating naked germanium detectors submersed into 65 m 3 of liquid argon. Similarly, LArGe runs Ge-detectors in 1 m 3 (1.4 tons) of liquid argon, which in addition is instrumented with photomultipliers to detect argon scintillation light. The light is used in anti-coincidence with the germanium detectors, to effectively suppress background events that deposit energy in the liquid argon. This work adresses the design, construction, and commissioning of LArGe. The background suppression efficiency has been studied in combination with a pulse shape discrimination (PSD) technique for various sources, which represent characteristic backgrounds to GERDA. Suppression factors of a few times 10 3 have been achieved. First background data of LArGe (without PSD) yield a background index of (0.12-4.6).10 -2 cts/(keV.kg.y) (90% c.l.), which is at the level of the Gerda phase I design goal. Furthermore, for the first time we measure the natural 42 Ar abundance (in parallel to Gerda), and have indication for the 2νββ-decay in natural germanium. (orig.)

LArGe. A liquid argon scintillation veto for GERDA

Energy Technology Data Exchange (ETDEWEB)

Heisel, Mark

2011-04-13

LArGe is a GERDA low-background test facility to study novel background suppression methods in a low-background environment, for possible applications in the GERDA experiment. GERDA searches for the neutrinoless double-beta decay in {sup 76}Ge, by operating naked germanium detectors submersed into 65 m{sup 3} of liquid argon. Similarly, LArGe runs Ge-detectors in 1 m{sup 3} (1.4 tons) of liquid argon, which in addition is instrumented with photomultipliers to detect argon scintillation light. The light is used in anti-coincidence with the germanium detectors, to effectively suppress background events that deposit energy in the liquid argon. This work adresses the design, construction, and commissioning of LArGe. The background suppression efficiency has been studied in combination with a pulse shape discrimination (PSD) technique for various sources, which represent characteristic backgrounds to GERDA. Suppression factors of a few times 10{sup 3} have been achieved. First background data of LArGe (without PSD) yield a background index of (0.12-4.6).10{sup -2} cts/(keV.kg.y) (90% c.l.), which is at the level of the Gerda phase I design goal. Furthermore, for the first time we measure the natural {sup 42}Ar abundance (in parallel to Gerda), and have indication for the 2{nu}{beta}{beta}-decay in natural germanium. (orig.)

Readout Electronics for the ATLAS LAr Calorimeter at HL-LHC

CERN Document Server

Chen, H; The ATLAS collaboration

2011-01-01

The ATLAS experiment is one of the two general-purpose detectors designed to study proton-proton collisions (14 TeV in the center of mass) produced at the Large Hadron Collider (LHC) and to explore the full physics potential of the LHC machine at CERN. The ATLAS Liquid Argon (LAr) calorimeters are high precision, high sensitivity and high granularity detectors designed to provide precision measurements of electrons, photons, jets and missing transverse energy. ATLAS (and its LAr Calorimeters) has been operating and collecting p-p collisions at LHC since 2009. The on-detector electronics (front-end) part of the current readout electronics of the calorimeters measures the ionization current signals by means of preamplifiers, shapers and digitizers and then transfers the data to the off-detector electronics (back-end) for further elaboration, via optical links. Only the data selected by the level-1 calorimeter trigger system are transferred, achieving a bandwidth reduction to 1.6 Gbps. The analog trigger sum sig...

The Liquid Argon Software Toolkit (LArSoft): Goals, Status and Plan

Energy Technology Data Exchange (ETDEWEB)

Pordes, Rush [Fermilab; Snider, Erica [Fermilab

2016-08-17

LArSoft is a toolkit that provides a software infrastructure and algorithms for the simulation, reconstruction and analysis of events in Liquid Argon Time Projection Chambers (LArTPCs). It is used by the ArgoNeuT, LArIAT, MicroBooNE, DUNE (including 35ton prototype and ProtoDUNE) and SBND experiments. The LArSoft collaboration provides an environment for the development, use, and sharing of code across experiments. The ultimate goal is to develop fully automatic processes for reconstruction and analysis of LArTPC events. The toolkit is based on the art framework and has a well-defined architecture to interface to other packages, including to GEANT4 and GENIE simulation software and the Pandora software development kit for pattern recognition. It is designed to facilitate and support the evolution of algorithms including their transition to new computing platforms. The development of the toolkit is driven by the scientific stakeholders involved. The core infrastructure includes standard definitions of types and constants, means to input experiment geometries as well as meta and event- data in several formats, and relevant general utilities. Examples of algorithms experiments have contributed to date are: photon-propagation; particle identification; hit finding, track finding and fitting; electromagnetic shower identification and reconstruction. We report on the status of the toolkit and plans for future work.

File list: ALL.Lar.50.AllAg.Larval_brain [Chip-atlas[Archive

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File list: ALL.Lar.10.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

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Reactor protection systems for the Replacement Research Reactor, ANSTO

International Nuclear Information System (INIS)

Morris, C.R.

2003-01-01

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

Koda = Koda / Ülar Mark, Kariina Kristiina Kaufmann, Kadri Tonto ; kommenteerinud Katrin Koov

Index Scriptorium Estoniae

Mark, Ülar, 1968-

2016-01-01

Minimalistliku vormiga eramu KODA Tallinnas, Järvekalda tee 33. Autorid Ülar Mark, Taavi Jakobson, Hannes Tamjärv, Kalev Ramjalg, Marek Strandberg. Arhitektid Ülar Mark, Kariina Kristiina Kaufmann, Kadri Tonto, Hannes Praks

File list: Oth.Lar.05.daf-12.AllCell [Chip-atlas[Archive

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File list: Oth.Lar.50.daf-12.AllCell [Chip-atlas[Archive

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File list: Oth.Lar.20.daf-12.AllCell [Chip-atlas[Archive

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Regionaalpoliitika ja arhitektuur / Ülar Mark

Index Scriptorium Estoniae

Mark, Ülar

1999-01-01

Kagu-Eesti Regionaalarengu Programmi raames Eesti Kunstiakadeemia arhitektuurikateedri IV kursuse eriala valikaines 1998/99 tehtud tööst "HyperMobiilne Reaalsus & Kagu Eesti" (juhendajad Ralf Tamm, Ülar Mark). Ühises töös osalesid Tartu Ülikooli geograafiatudengid (juhendaja Rein Ahas). 4 illustratsiooni

TRIGA reactor main systems

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2007-01-01

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

File list: Pol.Lar.50.RNA_polymerase_III.AllCell [Chip-atlas[Archive

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File list: Pol.Lar.10.RNA_polymerase_III.AllCell [Chip-atlas[Archive

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File list: Pol.Lar.05.RNA_polymerase_III.AllCell [Chip-atlas[Archive

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Development of new readout electronics for the ATLAS LAr Calorimeter at the sLHC

CERN Document Server

Strässner, A

2009-01-01

The readout of the ATLAS Liquid Argon (LAr) calorimeter is a complex multi-channel system to amplify, shape, digitize and process signals of the detector cells. The current on-detector electronics is not designed to sustain the ten times higher radiation levels expected at sLHC in the years beyond 2019/2020, and will be replaced by new electronics with a completely different readout scheme. The future on-detector electronics is planned to send out all data continuously at each bunch crossing, as opposed to the current system which only transfers data at a trigger-accept signal. Multiple high-speed and radiation-resistant optical links will transmit 100 Gb/s per front-end board. The off-detector processing units will not only process the data in real-time and provide digital data buffering, but will also implement trigger algorithms. An overview about the various components necessary to develop such a complex system is given. The current R&D activities and architectural studies of the LAr Calorimeter group...

Matlab implementation of LASSO, LARS, the elastic net and SPCA

DEFF Research Database (Denmark)

2005-01-01

There are a number of interesting variable selection methods available beside the regular forward selection and stepwise selection methods. Such approaches include LASSO (Least Absolute Shrinkage and Selection Operator), least angle regression (LARS) and elastic net (LARS-EN) regression. There al...... exists a method for calculating principal components with sparse loadings. This software package contains Matlab implementations of these functions. The standard implementations of these functions are available as add-on packages in S-Plus and R....

File list: Pol.Lar.05.RNA_Polymerase_II.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.05.RNA_Polymerase_II.AllCell ce10 RNA polymerase RNA Polymerase II Larvae h...ttp://dbarchive.biosciencedbc.jp/kyushu-u/ce10/assembled/Pol.Lar.05.RNA_Polymerase_II.AllCell.bed ...

Investigation and development of the suppression methods of the {sup 42}K background in LArGe

Energy Technology Data Exchange (ETDEWEB)

Lubashevskiy, Alexey [Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, D-69117 Heidelberg (Germany); Collaboration: GERDA-Collaboration

2013-07-01

GERDA is an ultra-low background experiment aimed for the neutrinoless double beta decay search. The search is performed using HPGe detectors operated in liquid argon (LAr). One of the most dangerous backgrounds in GERDA is the background from {sup 42}K which is a daughter isotope of cosmogenically produced {sup 42}Ar. {sup 42}K ions are collected towards to the detector by the electric field of the detector. Estimation of the background contribution and development of the suppression methods were performed in the low background test facility LArGe. For this purpose encapsulated HPGe and bare BEGe detectors were operated in 1m{sup 3} of LAr in the LArGe setup. It is equipped with scintillation veto, so particles which deposit part of their energy in LAr can be detected by 9 PMTs. In order to better understand background and to increase statistics the LAr of LArGe was spiked with specially produced {sup 42}Ar. All these investigations allowed us to estimate background contribution from {sup 42}K and demonstrate the possibility to suppress it in future measurements in GERDA Phase II.

La Medea de Lars von Trier

Directory of Open Access Journals (Sweden)

Iratxe Fresneda Delgado

2013-04-01

Full Text Available El presente artículo analiza el modo en el que Lars von Trier recrea para el cine el estereotipo de Medea. Mediante el análisis fílmico de la película y apoyándose en los estudios culturales, el texto se interroga acerca de la importancia y el poder potencial del cine a la hora recuperar el antiguo mito y demostrar su vigencia. El análisis amplía horizontes para la compresión de los mecanismos que articulan el entramado de significados de la película, donde Von Trier aporta una nueva visión del arquetipo de Medea uniéndola, a la tradición pictórica del Romanticismo. Una influencia que habita en las posteriores obras del director danés, donde el paisaje, la naturaleza, se erige en elemento catalizador de las pulsiones humanas, en su cómplice y testigo.This paper explores the way that Lars von Trier’s film recreates the stereotype of Medea. Using film analysis and based on cultural studies the article asks about the importance and potential power of cinema to recover the ancient myth and show their effects. The analysis expands horizons for the understanding of the mechanisms that link the network of meanings of the film, where the author offers a new vision of Medea's archetype attaching it to the pictorial tradition tied to the Romanticism. An influence that can be seen in the later works of Lars von Trier, where the landscape, the nature, stands as a catalyst of human drives, as his accomplice and witness.

File list: InP.Lar.10.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.10.AllAg.Larval_brain dm3 Input control Larvae Larval brain SRX1426944,SRX1...426946 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.10.AllAg.Larval_brain.bed ...

File list: InP.Lar.50.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.50.AllAg.Larval_brain dm3 Input control Larvae Larval brain SRX1426944,SRX1...426946 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.50.AllAg.Larval_brain.bed ...

File list: InP.Lar.05.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.05.AllAg.Larval_brain dm3 Input control Larvae Larval brain SRX1426944,SRX1...426946 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.05.AllAg.Larval_brain.bed ...

File list: InP.Lar.20.AllAg.Larval_brain [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.20.AllAg.Larval_brain dm3 Input control Larvae Larval brain SRX1426944,SRX1...426946 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.20.AllAg.Larval_brain.bed ...

File list: InP.Lar.05.Input_control.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.05.Input_control.AllCell ce10 Input control Input control Larvae SRX331089,...,SRX015099 http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/assembled/InP.Lar.05.Input_control.AllCell.bed ...

Reactor control system. PWR

International Nuclear Information System (INIS)

2009-01-01

At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

File list: Oth.Lar.50.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Oth.Lar.50.AllAg.3rd_instar dm3 TFs and others Larvae 3rd instar SRX318781,SRX31878...5306 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Oth.Lar.50.AllAg.3rd_instar.bed ...

File list: Oth.Lar.10.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Oth.Lar.10.AllAg.3rd_instar dm3 TFs and others Larvae 3rd instar SRX104963,SRX10497...4971,SRX331403 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Oth.Lar.10.AllAg.3rd_instar.bed ...

File list: Oth.Lar.20.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Oth.Lar.20.AllAg.3rd_instar dm3 TFs and others Larvae 3rd instar SRX318781,SRX31878...1403,SRX495243 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Oth.Lar.20.AllAg.3rd_instar.bed ...

File list: Oth.Lar.05.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Oth.Lar.05.AllAg.3rd_instar dm3 TFs and others Larvae 3rd instar SRX104964,SRX10496...1403,SRX495243 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Oth.Lar.05.AllAg.3rd_instar.bed ...

Application of SDSM and LARS-WG for simulating and downscaling of rainfall and temperature

Science.gov (United States)

Hassan, Zulkarnain; Shamsudin, Supiah; Harun, Sobri

2014-04-01

Climate change is believed to have significant impacts on the water basin and region, such as in a runoff and hydrological system. However, impact studies on the water basin and region are difficult, since general circulation models (GCMs), which are widely used to simulate future climate scenarios, do not provide reliable hours of daily series rainfall and temperature for hydrological modeling. There is a technique named as "downscaling techniques", which can derive reliable hour of daily series rainfall and temperature due to climate scenarios from the GCMs output. In this study, statistical downscaling models are used to generate the possible future values of local meteorological variables such as rainfall and temperature in the selected stations in Peninsular of Malaysia. The models are: (1) statistical downscaling model (SDSM) that utilized the regression models and stochastic weather generators and (2) Long Ashton research station weather generator (LARS-WG) that only utilized the stochastic weather generators. The LARS-WG and SDSM models obviously are feasible methods to be used as tools in quantifying effects of climate change condition in a local scale. SDSM yields a better performance compared to LARS-WG, except SDSM is slightly underestimated for the wet and dry spell lengths. Although both models do not provide identical results, the time series generated by both methods indicate a general increasing trend in the mean daily temperature values. Meanwhile, the trend of the daily rainfall is not similar to each other, with SDSM giving a relatively higher change of annual rainfall compared to LARS-WG.

File list: InP.Lar.20.Input_control.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.20.Input_control.AllCell dm3 Input control Input control Larvae SRX040614,S...457599,SRX1426944,SRX1426946,SRX1426948 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.20.Input_control.AllCell.bed ...

File list: His.Lar.05.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.05.AllAg.1st_instar dm3 Histone Larvae 1st instar SRX013088,SRX013009,SRX01...3064,SRX013044,SRX013097 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.05.AllAg.1st_instar.bed ...

File list: Pol.Lar.10.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.10.AllAg.3rd_instar dm3 RNA polymerase Larvae 3rd instar SRX287908,SRX28790...7,SRX013070,SRX013082 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Pol.Lar.10.AllAg.3rd_instar.bed ...

File list: His.Lar.20.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.20.AllAg.1st_instar dm3 Histone Larvae 1st instar SRX013009,SRX013088,SRX01...3044,SRX013064,SRX013097 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.20.AllAg.1st_instar.bed ...

File list: Pol.Lar.50.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.50.AllAg.3rd_instar dm3 RNA polymerase Larvae 3rd instar SRX287908,SRX28790...7,SRX013070,SRX013082 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Pol.Lar.50.AllAg.3rd_instar.bed ...

File list: ALL.Lar.50.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.50.AllAg.3rd_instar dm3 All antigens Larvae 3rd instar SRX1038029,SRX103803...SRX467060,SRX495306,SRX013058 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.50.AllAg.3rd_instar.bed ...

File list: Pol.Lar.05.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.05.AllAg.3rd_instar dm3 RNA polymerase Larvae 3rd instar SRX287907,SRX28790...8,SRX013070,SRX013082 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Pol.Lar.05.AllAg.3rd_instar.bed ...

File list: His.Lar.50.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.50.AllAg.1st_instar dm3 Histone Larvae 1st instar SRX013009,SRX013088,SRX01...3044,SRX013064,SRX013097 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.50.AllAg.1st_instar.bed ...

File list: His.Lar.10.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.10.AllAg.1st_instar dm3 Histone Larvae 1st instar SRX013088,SRX013064,SRX01...3044,SRX013009,SRX013097 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.10.AllAg.1st_instar.bed ...

File list: Pol.Lar.20.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.20.AllAg.3rd_instar dm3 RNA polymerase Larvae 3rd instar SRX287908,SRX28790...7,SRX013070,SRX013082 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Pol.Lar.20.AllAg.3rd_instar.bed ...

Morphological evolution of copper nanoparticles: Microemulsion reactor system versus batch reactor system

Science.gov (United States)

Xia, Ming; Tang, Zengmin; Kim, Woo-Sik; Yu, Taekyung; Park, Bum Jun

2017-07-01

In the synthesis of nanoparticles, the reaction rate is important to determine the morphology of nanoparticles. We investigated morphology evolution of Cu nanoparticles in this two different reactors, microemulsion reactor and batch reactor. In comparison with the batch reactor system, the enhanced mass and heat transfers in the emulsion system likely led to the relatively short nucleation time and the highly homogeneous environment in the reaction mixture, resulting in suppressing one or two dimensional growth of the nanoparticles. We believe that this work can offer a good model system to quantitatively understand the crystal growth mechanism that depends strongly on the local monomer concentration, the efficiency of heat transfer, and the relative contribution of the counter ions (Br- and Cl-) as capping agents.

File list: InP.Lar.05.Input_control.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.05.Input_control.AllCell dm3 Input control Input control Larvae SRX645428,S...287678,SRX1426944,SRX1426946,SRX1426948,SRX1426950 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.05.Input_control.AllCell.bed ...

File list: His.Lar.10.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.10.AllAg.2nd_instar dm3 Histone Larvae 2nd instar SRX013087,SRX013015,SRX01...3112,SRX013042,SRX013043,SRX013096 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.10.AllAg.2nd_instar.bed ...

File list: His.Lar.05.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.05.AllAg.2nd_instar dm3 Histone Larvae 2nd instar SRX013087,SRX013096,SRX01...3043,SRX013015,SRX013112,SRX013042 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.05.AllAg.2nd_instar.bed ...

File list: His.Lar.50.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.50.AllAg.2nd_instar dm3 Histone Larvae 2nd instar SRX013015,SRX013042,SRX01...3112,SRX013043,SRX013087,SRX013096 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.50.AllAg.2nd_instar.bed ...

File list: His.Lar.20.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available His.Lar.20.AllAg.2nd_instar dm3 Histone Larvae 2nd instar SRX013015,SRX013042,SRX01...3112,SRX013043,SRX013096,SRX013087 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/His.Lar.20.AllAg.2nd_instar.bed ...

Nuclear reactor safety systems

International Nuclear Information System (INIS)

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

1980-01-01

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

Histogramming in the LATOME-firmware for the Phase-1 upgrade of the ATLAS LAr calorimeter readout

Energy Technology Data Exchange (ETDEWEB)

Horn, Philipp; Hentges, Rainer; Straessner, Arno [Institut fuer Kern- und Teilchenphysik, Dresden (Germany)

2016-07-01

Due to the increased luminosity and the higher effective event rate after the phase 1 upgrade the ATLAS LAr detector needs new trigger electronics. The so-called LATOME-Board was designed as a LAr Digital Processing Blade (LPDB) to reconstruct the energy deposited by the particles and is an important part of the read out system. A prototype has already been build and the firmware for the on-board FPGA is under development. The insertion of a histogram-builder in this device gives the unique opportunity to look at untriggered data. This talk provides an insight in the LATOME-firmware and shows the different possibilities to implement the histogram-builder.

FFTF reactor assembly system technology

International Nuclear Information System (INIS)

Mangelsdorf, T.A.

1975-01-01

An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

The computerized reactor period measurement system for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1996-01-01

The article simply introduces the hardware, principle, and software of the computerized reactor period measurement system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between fission yield and pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computerized measurement system makes the reactor period measurement into automatical and intelligent and also improves the speed and precision of period data on-line process

Computer measurement system of reactor period for China fast burst reactor-II

International Nuclear Information System (INIS)

Zhao Wuwen; Jiang Zhiguo

1997-01-01

The author simply introduces the hardware, principle, and software of the reactor period computer measure system for China Fast Burst Reactor-II (CFBR-II). It also gives the relation between Fission yield and Pre-reactivity of CFBR-II reactor system of bared reactor with decoupled-component and system of bared reactor with multiple light-material. The computer measure system makes the reactor period measurement into automation and intellectualization and also improves the speed and precision of period data process on-line

Twentyfourth Podcast - Interview with Lars Holmgaard Christensen

DEFF Research Database (Denmark)

2008-01-01

Every wednesday the Doctoral School of Human Centred Informatics hosts a small research seminar, where PhD students and senior researchers can share and discuss their ongoing work. Today, we bring an interview from spring 2008. On February 27, Lars Holmgaard Christensen presented his paper "Homo...

File list: ALL.Lar.10.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.10.AllAg.1st_instar dm3 All antigens Larvae 1st instar SRX013088,SRX013064,...SRX013044,SRX013009,SRX013072,SRX013097,SRX013055,SRX013056 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.10.AllAg.1st_instar.bed ...

File list: ALL.Lar.50.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.50.AllAg.1st_instar dm3 All antigens Larvae 1st instar SRX013009,SRX013072,...SRX013088,SRX013044,SRX013064,SRX013097,SRX013056,SRX013055 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.50.AllAg.1st_instar.bed ...

File list: ALL.Lar.20.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.20.AllAg.1st_instar dm3 All antigens Larvae 1st instar SRX013009,SRX013072,...SRX013088,SRX013044,SRX013064,SRX013097,SRX013055,SRX013056 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.20.AllAg.1st_instar.bed ...

File list: ALL.Lar.05.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.05.AllAg.3rd_instar dm3 All antigens Larvae 3rd instar SRX104964,SRX104963,...SRX287718,SRX331401,SRX287658,SRX331366,SRX287906,SRX287678 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.05.AllAg.3rd_instar.bed ...

File list: ALL.Lar.10.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.10.AllAg.3rd_instar dm3 All antigens Larvae 3rd instar SRX104963,SRX104968,...SRX287718,SRX022334,SRX104976,SRX467107,SRX013086,SRX013058 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.10.AllAg.3rd_instar.bed ...

File list: ALL.Lar.20.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.20.AllAg.3rd_instar dm3 All antigens Larvae 3rd instar SRX1038029,SRX103803...SRX013082,SRX467107,SRX016173,SRX215499,SRX495243,SRX013058 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.20.AllAg.3rd_instar.bed ...

File list: ALL.Lar.05.AllAg.1st_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.05.AllAg.1st_instar dm3 All antigens Larvae 1st instar SRX013088,SRX013009,...SRX013064,SRX013044,SRX013072,SRX013055,SRX013097,SRX013056 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.05.AllAg.1st_instar.bed ...

File list: Unc.Lar.50.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Unc.Lar.50.AllAg.3rd_instar dm3 Unclassified Larvae 3rd instar SRX1038029,SRX103803...1,SRX1038032,SRX1038030,SRX022335,SRX032124,SRX032123,SRX013058 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Unc.Lar.50.AllAg.3rd_instar.bed ...

File list: Oth.Lar.20.Adenine_N6-methylation.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Oth.Lar.20.Adenine_N6-methylation.AllCell ce10 TFs and others Adenine N6-methylatio...n Larvae http://dbarchive.biosciencedbc.jp/kyushu-u/ce10/assembled/Oth.Lar.20.Adenine_N6-methylation.AllCell.bed ...

Digital control system of advanced reactor

International Nuclear Information System (INIS)

Peng Huaqing; Zhang Rui; Liu Lixin

2001-01-01

This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

File list: Pol.Lar.05.RNA_polymerase_II.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.05.RNA_polymerase_II.AllCell dm3 RNA polymerase RNA polymerase II Larvae SR...SRX151962,SRX182775,SRX661503,SRX013070,SRX013072,SRX013113,SRX013082,SRX151961 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Pol.Lar.05.RNA_polymerase_II.AllCell.bed ...

File list: Pol.Lar.20.RNA_polymerase_II.AllCell [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Pol.Lar.20.RNA_polymerase_II.AllCell dm3 RNA polymerase RNA polymerase II Larvae SR...SRX661503,SRX026742,SRX013070,SRX013072,SRX182775,SRX151961,SRX013082,SRX013113 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Pol.Lar.20.RNA_polymerase_II.AllCell.bed ...

File list: Unc.Lar.10.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Unc.Lar.10.AllAg.3rd_instar dm3 Unclassified Larvae 3rd instar SRX1038029,SRX103803...1,SRX1038032,SRX1038030,SRX032124,SRX032123,SRX022335,SRX022334,SRX013058 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Unc.Lar.10.AllAg.3rd_instar.bed ...

File list: ALL.Lar.05.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.05.AllAg.2nd_instar dm3 All antigens Larvae 2nd instar SRX013087,SRX013096,...SRX013043,SRX013015,SRX013112,SRX013042,SRX013113,SRX013016,SRX013114 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.05.AllAg.2nd_instar.bed ...

File list: ALL.Lar.50.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.50.AllAg.2nd_instar dm3 All antigens Larvae 2nd instar SRX013015,SRX013042,...SRX013112,SRX013016,SRX013114,SRX013043,SRX013087,SRX013096,SRX013113 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.50.AllAg.2nd_instar.bed ...

File list: ALL.Lar.20.AllAg.2nd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available ALL.Lar.20.AllAg.2nd_instar dm3 All antigens Larvae 2nd instar SRX013015,SRX013042,...SRX013112,SRX013043,SRX013016,SRX013114,SRX013096,SRX013087,SRX013113 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/ALL.Lar.20.AllAg.2nd_instar.bed ...

File list: Unc.Lar.20.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Unc.Lar.20.AllAg.3rd_instar dm3 Unclassified Larvae 3rd instar SRX1038029,SRX103803...1,SRX1038032,SRX1038030,SRX022335,SRX032124,SRX022334,SRX032123,SRX013058 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Unc.Lar.20.AllAg.3rd_instar.bed ...

File list: Unc.Lar.05.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available Unc.Lar.05.AllAg.3rd_instar dm3 Unclassified Larvae 3rd instar SRX1038031,SRX103802...9,SRX1038032,SRX1038030,SRX022335,SRX022334,SRX032124,SRX013058,SRX032123 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/Unc.Lar.05.AllAg.3rd_instar.bed ...

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

Pneumatic transport systems for TRIGA reactors

International Nuclear Information System (INIS)

Bolton, John A.

1970-01-01

Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

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

Reactor protection and shut-down system

International Nuclear Information System (INIS)

Klar

1980-01-01

The reactor protection system being a part of the reactor safety system. The requirements on the reactor protection system are: high safety with regard to signal processing, high availability, self-reporting of faults etc. The functional sections of the reactor protection system are the analog section, the logic section and the generating of output signals. Description of the operation characteristics and of the extension of function. (orig.)

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Samsun’da Ağır Ceza Mahkemesine Yansıyan Cinsel Suçlar

Directory of Open Access Journals (Sweden)

Berna Aydın

2004-04-01

Full Text Available Çalışmamızda cinsel suçların karakteristik özelliklerini belirlemek ve Türk Ceza Kanunu (TCK’nda yapılan değişiklikler çerçevesinde, cinsel suçlarla ilgili ortaya çıkan farklılıkları değerlendirmek amaçlanmıştır. 1999-2003 yılları arasında Samsun 1. Ağır Ceza Mahkemesi’nde karara bağlanmış dava dosyalarına ait karar kartonları incelenerek, cinsel suçlarla ilgili davalar çalışma kapsamına alınmıştır. Mağdurların %9,4’ü erkek iken, sanıkların tümünün erkek olduğu belirlenmiştir. Sanıkların %43,4’lük oranla en sık 21-30 yaş grubunda, mağdurların ise en sık (%31,6 16-18 yaş grubunda olduğu görülmüştür. Yargıya yansıyan cinsel suçlar arasında en sık ırza geçme suçları olduğu belirlenmiştir. Cinsel suçların adli ve tıbbi boyutu ile ilgilenen tıp ve hukuk mensupları, yeni yasal düzenlemeler yanı sıra cinsel suçların karakteristik özellikleri hakkında da bilgi sahibi olmalıdır. Her ilde ve büyük ilçe merkezlerinde cinsel suçlar değerlendirme merkezleri kurulması gerekmektedir. Anahtar kelimeler: Cinsel suçlar, cinsel saldırı, sanık, mağdur, mahkeme kararı

Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Reactor safety assessment system

International Nuclear Information System (INIS)

Sebo, D.E.; Bray, M.A.; King, M.A.

1987-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

Finding Neutrinos in LArTPCs using Convolutional Neural Networks

Science.gov (United States)

Wongjirad, Taritree

2017-09-01

Deep learning algorithms, which have emerged over the last decade, are opening up new ways to analyze data for many particle physics experiments. MicroBooNE, which is a neutrino experiment at Fermilab, has been exploring the use of such algorithms, in particular, convolutional neural networks (CNNS). CNNs are the state-of-the-art method for a large class of problems involving the analysis of images. This makes CNNs an attractive approach for MicroBooNE, whose detector, a liquid argon time projection chamber (LArTPC), produces high-resolution images of particle interactions. In this talk, I will discuss the ways CNNs can be applied to tasks like neutrino interaction detection and particle identification in MicroBooNE and LArTPCs.

Brinquedo e educação: na escola e no lar

Directory of Open Access Journals (Sweden)

Edda Bomtempo

Full Text Available Como o artigo mostra, brinquedo, brincar e jogar são considerados de grande importância para o desenvolvimento humano. Os tópicos considerados são: aprender brincando ou brincar para aprender; papéis educacionais dos brinquedos e jogos; os professores e os brinquedos; brinquedos e jogos no lar e na escola; o lar como um lugar divertido. Pais e professores precisam estar informados do valor dos brinquedos, brincadeiras e jogos para o desenvolvimento das crianças e dos adolescentes.

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

Breeding description for fast reactors and symbiotic reactor systems

International Nuclear Information System (INIS)

Hanan, N.A.

1979-01-01

A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

Kui lokaalne saab olla, et jääda globaalselt huvitavaks? / Lars Nittve ; interv. Anna Mustonen

Index Scriptorium Estoniae

Nittve, Lars

2005-01-01

Stockholmi moodsa kunsti muuseumi Moderna Museeti direktor Lars Nittve räägib kaasaegse kunstimuuseumi kogumispoliitika väljakutsetest ja rahvuslikkuse aspektist. Lars Nittve eluloolisi andmeid, lühidalt Moderna Museet'ist

Distonia laríngea respiratória

Directory of Open Access Journals (Sweden)

Lebl Mariana Dantas Aumond

2003-01-01

Full Text Available A distonia laríngea respiratória (DLR é uma desordem rara caracterizada por espasmos da musculatura adutora das pregas vocais durante a fase inalatória da respiração, com manifestação clínica de dispnéia e estridor. O diagnóstico etiológico do estridor laríngeo, entretanto, nem sempre é fácil de ser realizado, principalmente em situações emergenciais, de forma que a DLR pode não ser diagnosticada, o que nos leva a supor ser mais freqüente do que usualmente é descrita. O diagnóstico da DRL requer primeiramente a realização de uma história médica e exames laringológico e neurológico apropriados, com ênfase na verificação da presença de características distônicas e na exclusão de outras etiologias causadoras de movimentos paradoxais de pregas vocais. Muitos tratamentos foram propostos para a DLR, mas nenhum deles apresentou resultados satisfatórios. O uso da Toxina Botulínica do tipo A (Botox® no músculo tireoaritenoídeo tem oferecido melhoras admiráveis, apesar dos poucos casos descritos. Apresentamos dois casos clínicos de pacientes com DLR tratados com Botox® que apresentavam o fechamento glótico inspiratório causado tanto pelos espasmos anômalos dos músculos tireoaritenoídeos, como pela movimentação paradoxal da epiglote. Dentro da classificação proposta por Koufman e Blabock para as distonias laríngeas, inserimos um novo subtipo de DLR caracterizado pela presença de paroxismos de adução de estruturas glóticas e supraglóticas durante a respiração.

Increasing the efficiency of photon collection in LArTPCs: the ARAPUCA light trap

Science.gov (United States)

Cancelo, G.; Cavanna, F.; Escobar, C. O.; Kemp, E.; Machado, A. A.; Para, A.; Segreto, E.; Totani, D.; Warner, D.

2018-03-01

The Liquid Argon Time Projection Chambers (LArTPCs) are a choice for the next generation of large neutrino detectors due to their optimal performance in particle tracking and calorimetry. The detection of Argon scintillation light plays a crucial role in the event reconstruction as well as the time reference for non-beam physics such as supernovae neutrino detection and baryon number violation studies. In this contribution, we present the current R&D work on the ARAPUCA (Argon R&D Advanced Program at UNICAMP), a light trap device to enhance Ar scintillation light collection and thus the overall performance of LArTPCs. The ARAPUCA working principle is based on a suitable combination of dichroic filters and wavelength shifters to achieve a high efficiency in light collection. We discuss the operational principles, the last results of laboratory tests and the application of the ARAPUCA as the alternative photon detection system in the protoDUNE detector.

Nuclear reactor safety system

International Nuclear Information System (INIS)

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

1983-01-01

The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

Nuclear reactor shutdown system

International Nuclear Information System (INIS)

Mangus, J.D.; Cooper, M.H.

1982-01-01

An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

File list: InP.Lar.50.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.50.AllAg.3rd_instar dm3 Input control Larvae 3rd instar SRX287726,SRX331369...87917,SRX287921,SRX288023,SRX467107,SRX016172,SRX016173 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.50.AllAg.3rd_instar.bed ...

Sosyal Pekiştireçlerin ve Model Davranışlarının, Çocukların Ahlaki Yargılarının Şekillenmesindeki Etkisi (Bandura Örneği)

OpenAIRE

GÜREL, Ramazan

2014-01-01

Bu makale, Bandura'nın 5-11 yaş aralığındaki çocukların ahlaki yargılarının şekillenmesinde model alma ve sosyal pekiştireçlerin etkisini test etmek üzere gerçekleştirdiği araştırmasının ulaştığı sonuçları değerlendirmeyi amaçlamaktadır. Bu doğrultuda makalede, Bandura tarafından üç farklı deney ortamında 5-11 yaşlarındaki çocuklar üzerinde, ahlaki gelişim ve uyum sağlamada sosyal öğrenme, pekiştireçler ve model davranışlarının etkilerine dair gerçekleştirilen deneysel faaliyetlerin uygu...

Basic concept of common reactor physics code systems. Final report of working party on common reactor physics code systems (CCS)

International Nuclear Information System (INIS)

2004-03-01

A working party was organized for two years (2001-2002) on common reactor physics code systems under the Research Committee on Reactor Physics of JAERI. This final report is compilation of activity of the working party on common reactor physics code systems during two years. Objectives of the working party is to clarify basic concept of common reactor physics code systems to improve convenience of reactor physics code systems for reactor physics researchers in Japan on their various field of research and development activities. We have held four meetings during 2 years, investigated status of reactor physics code systems and innovative software technologies, and discussed basic concept of common reactor physics code systems. (author)

Memecik Zeytinyağlarının Biyokimyasal Karakterizasyonu

Directory of Open Access Journals (Sweden)

Huri İlyasoğlu

2015-02-01

Full Text Available Zeytinyağı kalitesi botanik ve coğrafi orijin ile ilişkili olduğundan zeytinyağının çeşit ve bölgesel karakte­rizasyonu önem kazanmıştır. Bu çalışmada, ülkemizin ekonomik açıdan en önemli zeytinyağlarından biri­si olan Memecik zeytinyağlarının kalite parametreleri, yağ asitleri, triaçilgliserol ve sterol kompozisyonu, a-tokoferol ve toplam fenolik madde içerikleri, fenolik madde ve aroma profili belirlenmiştir. Analiz edilen zeytinyağı örneklerinin kalite parametreleri, yağ asitleri ve sterol kompozisyonun (kampesterol hariç, Türk Gıda Kodeksi ile uyumlu olduğu saptanmıştır. Memecik zeytinyağlarında tespit edilen başlıca yağ asitleri oleik asit (% 73.37-75.64, palmitik asit (% 11.45-13.84 ve linoleik asittir (% 7.33-8.91. Triolein (% 63.50-68.32, palmitodiolein (% 18.25-25.82 ve dioleolinolein (% 6.01-9.18 zeytinyağı örneklerinde tespit edilen başlıca triaçilgliserollerdir. Sterol kompozisyonu oluşturan başlıca steroller, b-sitosterol (% 80.76-83.00, D5-avenasterol (% 11.02-12.78 ve kampesteroldür (% 4.01-4.97. Memecik çeşidi zeytinyağlarının fenolik madde profilinde, hidroksitirozol, tirozol, ferulik asit, p-kumarik asit, apigenin ve luteolin ve aroma profi­linde hekzanal, hekzanol, E-2-hekzenal, E-2-hekzenol ve Z-3-hekzenol tespit edilmiştir.

File list: InP.Lar.20.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.20.AllAg.3rd_instar dm3 Input control Larvae 3rd instar SRX467108,SRX287726...67103,SRX288024,SRX288023,SRX104976,SRX016172,SRX467107,SRX016173 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.20.AllAg.3rd_instar.bed ...

File list: InP.Lar.10.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.10.AllAg.3rd_instar dm3 Input control Larvae 3rd instar SRX016172,SRX016173...87701,SRX287922,SRX467104,SRX287918,SRX287718,SRX104976,SRX467107 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.10.AllAg.3rd_instar.bed ...

File list: InP.Lar.05.AllAg.3rd_instar [Chip-atlas[Archive

Lifescience Database Archive (English)

Full Text Available InP.Lar.05.AllAg.3rd_instar dm3 Input control Larvae 3rd instar SRX016173,SRX288024...87921,SRX287718,SRX331401,SRX287658,SRX331366,SRX287906,SRX287678 http://dbarchive.biosciencedbc.jp/kyushu-u/dm3/assembled/InP.Lar.05.AllAg.3rd_instar.bed ...

Molecular ecology of anaerobic reactor systems

DEFF Research Database (Denmark)

Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter

2003-01-01

Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible for these ......Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

The Optimization of power reactor control system

International Nuclear Information System (INIS)

Danupoyo, S.D.

1997-01-01

A power reactor is an important part in nuclear powered electrical plant systems. Success in controlling the power reactor will establish safety of the whole power plant systems. Until now, the power reactor has been controlled by a classical control system that was designed based on output feedback method. To meet the safety requirements that are now more restricted, the recently used power reactor control system should be modified. this paper describes a power reactor control system that is designed based on a state feedback method optimized with LQG (Linear-quadrature-gaussian) method and equipped with a state estimator. A pressurized-water type reactor has been used as the model. by using a point kinetics method with one group delayed neutrons. the result of simulation testing shows that the optimized control system can control the power reactor more effective and efficient than the classical control system

Halkın Genetiği Değiştirilmiş Ürünlere/Üretilme Süreçlerine Yönelik Algıları ve Etik İnançları

Directory of Open Access Journals (Sweden)

Özlen Özgen

2014-02-01

Full Text Available Bu araştırmanın amacı, halkın genetiği değiştirilmiş ürünlere ve üretilme süreçlerine yönelik fayda/risk algıları ile etik inançları arasındaki ilişkinin incelenmesidir. Araştırma materyalinin toplanmasında karşılıklı görüşme tekniği kullanılmıştır. Likert tipi cümlelere verilen yanıtlar puanlanmış, geçerlik ve güvenirlik analizi yapılmıştır. Yaş değişkenine bağlı farklılığın belirlenebilmesi için t-testi uygulanmıştır. Bireylerin genetiği değiştirilmiş ürünlere/üretilme süreçlerine bağlı fayda/risk algıları ile etik inançları arasındaki ilişkinin incelenmesi amacı ile Pearson korelasyon analizi yapılmıştır. Bulgular; tüketicilerin genetiği değiştirilmiş ürünlere ilişkin fayda algılarının, genetiği değiştirilmiş ürünlere ve üretilme süreçlerine ilişkin risk algılarının, genetiği değiştirilmiş ürünlere ve üretilme süreçlerine ilişkin etik inançlarının yaşa bağlı olarak değiştiğini göstermektedir. Pearson korelasyon analizi sonucunda p<0.001 düzeyinde anlamlı ilişkiler saptanmıştır.

DarkSide-20k: A 20 Tonne Two-Phase LAr TPC for Direct Dark Matter Detection at LNGS

Energy Technology Data Exchange (ETDEWEB)

Aalseth, C.E.; et al.

2017-07-25

Building on the successful experience in operating the DarkSide-50 detector, the DarkSide Collaboration is going to construct DarkSide-20k, a direct WIMP search detector using a two-phase Liquid Argon Time Projection Chamber (LArTPC) with an active (fiducial) mass of 23 t (20 t). The DarkSide-20k LArTPC will be deployed within a shield/veto with a spherical Liquid Scintillator Veto (LSV) inside a cylindrical Water Cherenkov Veto (WCV). Operation of DarkSide-50 demonstrated a major reduction in the dominant $^{39}$Ar background when using argon extracted from an underground source, before applying pulse shape analysis. Data from DarkSide-50, in combination with MC simulation and analytical modeling, shows that a rejection factor for discrimination between electron and nuclear recoils of $\\gt3\\times10^9$ is achievable. This, along with the use of the veto system, is the key to unlocking the path to large LArTPC detector masses, while maintaining an "instrumental background-free" experiment, an experiment in which less than 0.1 events (other than $\

Developing Light Collection Enhancements and Wire Tensioning Methods for LArTPC Neutrino Detectors

Energy Technology Data Exchange (ETDEWEB)

Spagliardi, Fabio [Univ. of Manchester (United Kingdom)

2017-01-01

Liquid argon Time Projection Chambers (LArTPCs) are becoming widely used as neutrino detectors because of their image-like event reconstruction which enables precision neutrino measurements. They primarily use ionisation charge to reconstruct neutrino events. It has been shown, however, that the scintillation light emitted by liquid argon could be exploited to improve their performance. As the neutrino measurements planned in the near future require large-scale experiments, their construction presents challenges in terms of both charge and light collection. In this dissertation we present solutions developed to improve the performance in both aspects of these detectors. We present a new wire tensioning measurement method that allows a remote measurement of the tension of the large number wires that constitute the TPC anode. We also discuss the development and installation of WLS-compound covered foils for the SBND neutrino detector at Fermilab, which is a technique proposed t o augment light collection in LArTPCs. This included preparing a SBND-like mesh cathode and testing it in the Run III of LArIAT, a test beam detector also located at Fermilab. Finally, we present a study aimed at understanding late scintillation light emitted by recombining positive argon ions using LArIAT data, which could affect large scale surface detectors.

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

REACTOR: an expert system for diagnosis and treatment of nuclear reactor accidents

International Nuclear Information System (INIS)

Nelson, W.R.

1982-01-01

REACTOR is an expert system under development at EG and G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system technology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future plans for validation and evaluation of REACTOR are presented. The concept of using both event-oriented and function-oriented strategies for accident diagnosis is discussed. The response tree concept for representing expert knowledge is also introduced

Experimental observation of an extremely high electron lifetime with the ICARUS-T600 LAr-TPC

CERN Document Server

Antonello, M; Benetti, P; Boffelli, F; Bubak, A; Calligarich, E; Centro, S; Cesana, A; Cieslik, K; Cline, D B; Cocco, A G; Dabrowska, A; Dermenev, A; Dolfini, R; Falcone, A; Farnese, C; Fava, A; Ferrari, A; Fiorillo, G; Gibin, D; Gninenko, S; Guglielmi, A; Haranczyk, M; Holeczek, J; Kirsanov, M; Kisiel, J; Kochanek, I; Lagoda, J; Mania, S; Menegolli, A; Meng, G; Montanari, C; Otwinowski, S; Picchi, P; Pietropaolo, F; Plonski, P; Rappoldi, A; Raselli, G L; Rossella, M; Rubbia, C; Sala, P; Scaramelli, A; Segreto, E; Sergiampietri, F; Stefan, D; Sulej, R; Szarska, M; Terrani, M; Torti, M; Varanini, F; Ventura, S; Vignoli, C; Wang, H; Yang, X; Zalewska, A; Zani, A; Zaremba, K

2014-01-01

The ICARUS T600 detector, the largest liquid Argon Time Projection Chamber (LAr-TPC) realized after many years of RD activities, was installed and successfully operated for 3 years at the INFN Gran Sasso underground Laboratory. One of the most important issues was the need of an extremely low residual electronegative impurity content in the liquid Argon, in order to transport the free electrons created by the ionizing particles with a very small attenuation along the drift path. The solutions adopted for the Argon re-circulation and purification systems have permitted to reach impressive results in terms of Argon purity and a free electron lifetime exceeding 15 ms, corresponding to about 20 parts per trillion of equivalent O2 contamination, a milestone for any future project involving LAr-TPC's and the development of higher detector mass scales.

Propose Reactor Control and Monitoring System for RTP

International Nuclear Information System (INIS)

Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

2011-01-01

Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

Avaliação laríngea em pacientes reumatológicos Laryngeal assessment in reumatic disease patients

Directory of Open Access Journals (Sweden)

Hugo Valter Lisboa Ramos

2005-08-01

Full Text Available As doenças reumáticas produzem alterações sistêmicas e podem, por isso, comprometer os vasos sangüíneos, as serosas e as mucosas de todo o trato aerodigestivo. Casos esporádicos de acometimento laríngeo por doenças reumáticas têm sido descritos. Esse estudo tem por objetivo avaliar e descrever as alterações laríngeas encontradas em pacientes reumatológicos. FORMA DE ESTUDO: coorte transversal. MATERIAL E MÉTODO: Estudo transversal com pacientes portadores de lúpus eritematoso sistêmico, esclerodermia e doença mista do tecido conjuntivo. Os pacientes submeteram-se a exame clínico otorrinolaringológico e à videolaringoestroboscopia. RESULTADOS: Foram incluídos no estudo 27 pacientes sendo que 26 conseguiram realizar a videolaringoestroboscopia. Alterações laríngeas foram observadas em 11 dos 12 pacientes portadores de lúpus, nos 11 pacientes portadores de esclerodermia e nos 3 pacientes portadores de doença mista do tecido conjuntivo. Lesões sugestivas de nódulo em bambu foram identificados em 5 pacientes e 92,3% dos pacientes apresentaram sinais laríngeos de síndrome faringolaríngea do refluxo. CONCLUSÃO: Neste estudo identificamos 5 lesões sugestivas de nódulos em bambu e sinais laríngeos de refluxo em quase todos os pacientes.Rheumatic diseases usually promote several systemic disorders, which can affect blood vessels, mucosa and serosa of the aerodigestive tract. Scarce laryngeal involvement has been described in these patients and this study aims at investigating laryngeal alterations found in patients with rheumatic diseases. STUDY DESIGN: transversal cohort. MATERIAL AND METHOD: A transversal study was developed with systemic lupus erythematous, systemic sclerosis and mixed connective tissue disease's patients. They were evaluated by means of clinical examinations and videolaryngoestroboscopy. RESULTS: Twenty-seven patients were included in the study, 26 succeeded in completing the

Dosimetry system of the RB reactor

International Nuclear Information System (INIS)

Lolic, B.; Vukadin, D.

1962-01-01

Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

JAERI thermal reactor standard code system for reactor design and analysis SRAC

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1985-01-01

SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

Dead Time in the LAr Calorimeter Front-End Readout

CERN Document Server

Gingrich, D M

2002-01-01

We present readout time, latency, buffering, and dead-time calculations for the switched capacitor array controllers of the LAr calorimeter. The dead time is compared with algorithms for the dead-time generation in the level-1 central trigger processor.

RSAS: a Reactor Safety Assessment System

International Nuclear Information System (INIS)

Sebo, D.E.; Dixon, B.W.; Bray, M.A.

1985-01-01

The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

Genetic association analysis of LARS2 with type 2 diabetes

DEFF Research Database (Denmark)

Reiling, E; Jafar-Mohammadi, B; van 't Riet, E

2010-01-01

was to establish the true contribution of this variant and common variants in LARS2 (MAF > 5%) to type 2 diabetes risk. METHODS: We combined genome-wide association data (n = 10,128) from the DIAGRAM consortium with independent data derived from a tagging single nucleotide polymorphism (SNP) approach in Dutch...... individuals (n = 999) and took forward two SNPs of interest to replication in up to 11,163 Dutch participants (rs17637703 and rs952621). In addition, because inspection of genome-wide association study data identified a cluster of low-frequency variants with evidence of type 2 diabetes association, we...... of sequence variants in LARS2 in type 2 diabetes susceptibility, we found no evidence to support previous data indicating a role in type 2 diabetes susceptibility....

Small space reactor power systems for unmanned solar system exploration missions

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the application of small nuclear reactor space power systems to the Mariner Mark II Cassini spacecraft/mission was conducted. The purpose of the study was to identify and assess the technology and performance issues associated with the reactor power system/spacecraft/mission integration. The Cassini mission was selected because study of the Saturn system was identified as a high priority outer planet exploration objective. Reactor power systems applied to this mission were evaluated for two different uses. First, a very small 1 kWe reactor power system was used as an RTG replacement for the nominal spacecraft mission science payload power requirements while still retaining the spacecraft's usual bipropellant chemical propulsion system. The second use of reactor power involved the additional replacement of the chemical propulsion system with a small reactor power system and an electric propulsion system. The study also provides an examination of potential applications for the additional power available for scientific data collection. The reactor power system characteristics utilized in the study were based on a parametric mass model that was developed specifically for these low power applications. The model was generated following a neutronic safety and operational feasibility assessment of six small reactor concepts solicited from U.S. industry. This assessment provided the validation of reactor safety for all mission phases and generatad the reactor mass and dimensional data needed for the system mass model

Nuclear reactor coolant and cover gas system

International Nuclear Information System (INIS)

George, J.A.; Redding, A.H.; Tower, S.N.

1976-01-01

A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

Proceedings of workshop on reactor shutdown system

International Nuclear Information System (INIS)

1997-03-01

India has gained considerable experience in design, development, construction and operation of research and power reactors during the last four decades. Reactor shutdown system (RSS) is the most important engineered safety system of any reactor. A lot of technological developments have taken place to improve the reactor shutdown systems, particularly with advancement in reliability analysis and instrumentation and control. If the reactor is not shutdown, the fuel may melt, releasing radioactivity and possibly reactivity addition as in the case of Fast Breeder Reactor (FBR). Apart from radiological safety consequences, large investment has to be written off. The function of the RSS is to stop fission chain reaction and prevent breach of fuel. The design of RSS is multidisciplinary. It requires reactor physics analysis, design of absorber rods, drive mechanisms, safety logic to order shutdown and instrumentation to detect unsafe conditions. High reliability is essential and this requires two independent shutdown systems. This book contains the proceedings of the workshop on reactor shutdown system and papers relevant to INIS are indexed separately

MAPLE-X10 reactor digital control system

International Nuclear Information System (INIS)

Deverno, M.T.; Hinds, H.W.

1991-10-01

The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MW t , pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability

Arrival of the last cryostat for the ATLAS LAr calorimeter at CERN

CERN Multimedia

Aleksa, M; Oberlack, H

On Wednesday, 4th June the last cryostat for the ATLAS LAr calorimeter (end-cap A) arrived at CERN and was immediately unloaded from the truck in building 180 (see Figures 1 and 2), where the integration of the LAr calorimeters into their cryostats takes place. The transport from the Italian company SIMIC, where both end-cap calorimeters have been produced took longer than expected due to delays because of the G8 summit. Thanks to the great effort by the CERN Host State office and the French-German steering group that supplies the end-cap cryostat as an in-kind contribution to the LAr collaboration, an exceptional convoy was finally available and the cryostat could make its way to CERN. Fig.1 (left): Truck with the end-cap cryostat. Fig.2 (right): Unloading the cryostat in bldg. 180. Each end-cap cryostat will contain an electromagnetic calorimeter wheel, two wheels of a hadronic calorimeter, and a forward calorimeter. The design of the cryostat as a double vessel structure made of Aluminum fulfills t...

Solvent refined coal reactor quench system

Science.gov (United States)

Thorogood, Robert M.

1983-01-01

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

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

ATLAS LAr Calorimeter Performance in LHC Run-2

CERN Document Server

Morgenstern, Stefanie; The ATLAS collaboration

2018-01-01

Liquid-argon (LAr) sampling calorimeters are employed by ATLAS for all electromagnetic calorimetry in the pseudo-rapidity region $\\eta<3.2$, and for hadronic and forward calorimetry in the region from $\\eta=1.5$ to $\\eta=4.9$. In the first LHC run a total luminosity of $27\\,\\mathrm{fb}^{-1}$ has been collected at centre-of-mass energies of $7-8\\,\\mathrm{TeV}$. After detector consolidation during a long shutdown, Run-2 started in 2015 and $86.4\\,\\mathrm{fb}^{-1}$ of data at a centre-of-mass energy of $13\\,\\mathrm{TeV}$ have been recorded. In order to realize the level-1 acceptance rate of $100\\,\\mathrm{kHz}$ in Run-2 data taking, the number of readout samples recorded and used for the energy and the time measurement has been modified from five to four while keeping the expected performance. The well calibrated and highly granular LAr calorimeter reached its design values both in energy measurement as well as in direction resolution. This contribution will give an overview of the detector operation, hardware...

Nuclear reactor auxiliary heat removal system

International Nuclear Information System (INIS)

Thompson, R.E.; Pierce, B.L.

1977-01-01

An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

Nuclear reactor trip system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

Each parameter of the processes of a nuclear reactor and components operatively associated with it is monitored by a set of four like sensors. A trip system normally operates on a ''two out four'' configuration; i.e., to trip the reactor it is necessary that at least two sensors of a set sense an off-normal parameter. This assumes that all sensors are in normal operating condition. However, when a sensor is in test or is subject to maintenance or is defective or disabled, the ''two out of four''configuration would be reduced to a ''one out of three'' configuration because the affected sensor is taken out of service. This would expose the system to the possibility that a single sensor failure, which may be spurious, will cause a trip of the reactor. To prevent this, it is necessary that the affected sensor be bypassed. If only one sensor is bypassed, the system operates on a ''two out of three'' configuration. With two sensors bypassed, the sensing of an off-normal parameter by a third sensor trips the reactor. The by-pass circuit also disables the circuit coupling the by-passed sensor to the trip circuit. (author)

Mobil reklamlar ve mobil reklam araçlarına yönelik tutumlar

OpenAIRE

Barutçu, Süleyman; Öztürk Göl, Meltem

2009-01-01

Bu çalışmanın amacı, pazarlama bölümü yöneticilerinin dikkatlerini mobil iletişim teknolojilerindeki gelişmelerin sonucu olarak ortaya çıkan mobil pazarlama ve mobil reklam uygulamalarına çekmektir. Çalışmanın kavramsal analiz bölümünde mobil pazarlama, mobil reklamların önemi açıklanmış ve mobil reklam araçları (SMS, MMS ve Bluetooth Reklamları) analiz edilmiştir. Araştırma bölümünde ise mobil telefon kullanıcılarının mobil reklam araçlarına yönelik tutumlarının karşılaştırmalı olarak belirl...

TREAT Reactor Control and Protection System

International Nuclear Information System (INIS)

Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

1985-01-01

The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

Computerized reactor pressure vessel materials information system

International Nuclear Information System (INIS)

Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

1980-10-01

A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

Coolant clean-up system in nuclear reactor

International Nuclear Information System (INIS)

Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

1987-01-01

Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

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.)

High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

International Nuclear Information System (INIS)

1983-01-01

The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

Budget impact of pasireotide LAR for the treatment of acromegaly, a rare endocrine disorder.

Science.gov (United States)

Zhang, J J; Nellesen, D; Ludlam, W H; Neary, M P

2016-01-01

Acromegaly is a rare disorder characterized by the over-production of growth hormone (GH). Patients often experience a range of chronic comorbidities including hypertension, cardiac dysfunction, diabetes, osteoarthropathy, and obstructive sleep apnea. Untreated or inadequately controlled patients incur substantial healthcare costs, while normalization of GH levels may reduce morbidity and mortality rates to be comparable to the general population. To assess the 3-year budget impact of pasireotide LAR on a US managed care health plan following pasireotide LAR availability. Two separate economic models were developed: one from the perspective of an entire health plan and another from the perspective of a pharmacy budget. The total budget impact model includes costs of drug therapies and other costs for treatment, monitoring, management of adverse events, and comorbidities. The pharmacy cost calculator only considers drug costs. The total estimated budget impact associated with the introduction of pasireotide LAR is 0.31 cents ($0.0031) per member per month (PMPM) in the first year, 0.78 cents ($0.0078) in the second year, and 1.42 cents ($0.0142) in the third year following FDA approval. Costs were similar or lower from a pharmacy budget impact perspective. For each patient achieving disease control, cost savings from reduced comorbidities amounted to $10,240 per year. Published data on comorbidities for acromegaly are limited. In the absence of data on acromegaly-related costs for some comorbidities, comorbidity costs for the general population were used (may be under-estimates). The budget impact of pasireotide LAR is expected to be modest, with an expected increase of 1.42 cents PMPM on the total health plan budget in the third year after FDA approval. The efficacy of pasireotide LAR in acromegaly, as demonstrated in head-to-head trials compared with currently available treatment options, is expected to be associated with a reduction of the prevalence of

Development of Vibration Diagnostic System in Research Reactors

International Nuclear Information System (INIS)

EL-Kafas, A. A.

1999-01-01

Early failure detection and diagnosis system are an important group with increasing interest with the operating support system. Already existing system to monitor integrity of primary system components are vibration and acoustic monitoring system (2,3). The development of vibration diagnostic system for MARIA reactor (30 MW)-the second research reactor in Poland -was made. The new system is applied for the Egypt research reactor (ETRR-1). This paper represents the result obtained during the operation of this activity that carried out at MARIA and ETRR-1 reactors

ATLAS LAr Calorimeter Performance in LHC Run-2

CERN Document Server

Morgenstern, Stefanie; The ATLAS collaboration

2018-01-01

Liquid argon (LAr) sampling calorimeters are employed by ATLAS for all electromagnetic calorimetry in the pseudo-rapidity region eta<3.2, and for hadronic and forward calorimetry in the region from eta=1.5 to eta=4.9. In the first LHC run a total luminosity of 27 fb-1 has been collected at c.o.m energies of 7-8 TeV. After detector consolidation during a long shutdown, Run-2 started in 2015 and 86.4fb-1 of data at a c.o.m energy of 13 TeV have been recorded. In order to realize the level-1 acceptance rate of 100 kHz in Run-2 data taking, the number of read-out samples recorded and used for the energy and the time measurement has been modified from five to four while keeping the expected performance. The well calibrated and highly granular LAr Calorimeter reached its design values both in energy measurement as well as in direction resolution. This contribution will give an overview of the detector operation, hardware improvements, changes in the monitoring and data quality procedures, to cope with increased ...

Nuclear reactor refueling system

International Nuclear Information System (INIS)

Wade, E.E.

1978-01-01

A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

Space reactor electric systems: system integration studies, Phase 1 report

International Nuclear Information System (INIS)

Anderson, R.V.; Bost, D.; Determan, W.R.; Harty, R.B.; Katz, B.; Keshishian, V.; Lillie, A.F.; Thomson, W.B.

1983-01-01

This report presents the results of preliminary space reactor electric system integration studies performed by Rockwell International's Energy Systems Group (ESG). The preliminary studies investigated a broad range of reactor electric system concepts for powers of 25 and 100 KWe. The purpose of the studies was to provide timely system information of suitable accuracy to support ongoing mission planning activities. The preliminary system studies were performed by assembling the five different subsystems that are used in a system: the reactor, the shielding, the primary heat transport, the power conversion-processing, and the heat rejection subsystems. The subsystem data in this report were largely based on Rockwell's recently prepared Subsystem Technology Assessment Report. Nine generic types of reactor subsystems were used in these system studies. Several levels of technology were used for each type of reactor subsystem. Seven generic types of power conversion-processing subsystems were used, and several levels of technology were again used for each type. In addition, various types and levels of technology were used for the shielding, primary heat transport, and heat rejection subsystems. A total of 60 systems were studied

Modular reactor head shielding system

International Nuclear Information System (INIS)

Jacobson, E. B.

1985-01-01

An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

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

Functional systems of a pressurized water reactor

International Nuclear Information System (INIS)

Heinzel, V.

1982-01-01

The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

Building reactor operator sustain expert system with C language integrated production system

International Nuclear Information System (INIS)

Ouyang Qin; Hu Shouyin; Wang Ruipian

2002-01-01

The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

The unique safety challenges of space reactor systems

International Nuclear Information System (INIS)

Lanes, S.J.; Marshall, A.C.

1991-01-01

Compact reactor systems can provide high levels of power for extended periods in space environments. Their relatively low mass and their ability to operate independently of their proximity to the sun make reactor power systems high desirable for many civilian and military space missions. The US Department of Energy is developing reactor system technologies to provide electrical power for space applications. In addition, reactors are now being considered to provide thermal power to a hydrogen propellant for nuclear thermal rocketry. Space reactor safety issues differ from commercial reactor issues, in some areas, because of very different operating requirements and environments. Accidents similar to those postulated for commercial reactors must be considered for space reactors during their operational phase. Safety strategies will need to be established that account for the consequences of the loss of essential power

Pressurized water reactor systems

International Nuclear Information System (INIS)

Meyer, P.J.

1975-01-01

Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

Gas-cooled reactor for space power systems

International Nuclear Information System (INIS)

Walter, C.E.; Pearson, J.S.

1987-05-01

Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

Reactor shutdown back-up system

International Nuclear Information System (INIS)

Hirao, Seizo; Sakashita, Motoaki.

1982-01-01

Purpose: To prevent back flow of poison upon injection to a moderator recycling pipeway. Constitution: In a nuclear reactor comprising a moderator recycling system for recycling and cooling moderator through a control rod guide pipe and a rapid poison injection system for rapidly injecting a poison solution at high density into the moderator by way of the same control rod guide pipe as a reactor shutdown back-up system, a mechanism is provided for preventing the back flow of a poison solution at high density into the moderator recycling system upon rapid injection of poison. An orifice provided in the joining pipeway to the control rod guide pipe on the side of the moderator recycling system is utilized as the back flow preventing device for the poison solution and the diameter for the orifice is determined so as to provide a constant ratio between the pressure loss in the control rod guide pipe and the pressure loss in the moderator recycling system pipe line upon usual reactor operation. (Kawakami, Y.)

LAr calorimeter for SCC with a common vacuum bulkhead---a concept to improve hermeticity

International Nuclear Information System (INIS)

Pope, W.L.; Watt, R.D.

1989-11-01

A new concept for a Barrel/Endcap LAr Calorimeter (LAC) is described in which the Barrel and Endcaps are in separate vacuum enclosures but share a common vacuum bulkhead (CVB). We explore 2 possible bulkhead construction types; welded plate sandwich panels, and brazed sandwich panels in which the core is an isotropic cellular solid--foamed aluminum. Gas lines and electric cables from he innermost Drift Chamber pass through radial holes in the core of the sandwich bulkhead. The CVB concept offers the potential to obtain a more hermetic calorimeter with significantly reduced dead material and/or space in the interface region common to conventional design LAr detectors for the SSC with Endcap features. To utilize a common additional steps to remove the Drift Chamber, a large increase in Endcap standby heat leak, and perhaps, new cryogenic safety issues. We find that significant amount of dead mass can be removed from critical regions of the vacuum shells when compared to a promising SSC LAC reference design. It is also shown that the increased standby heat leak of this concept can be easily removed by existing cooling capacity in another large LAr calorimeter. It is further shown that shut-downs need not be appreciably longer. Finally, it is argued that cryogen spill hazards can be avoided if the Endcap's LAr is removed during Drift chamber maintenance shutdowns, and that cryogenic safety is not compromised

ATLAS LAr Calorimeter Trigger Electronics Phase-1 Upgrade

CERN Document Server

Aad, Georges; The ATLAS collaboration

2017-01-01

The upgrade of the Large Hadron Collider (LHC) scheduled for a shut-down period of 2019-2020, referred to as the Phase-I upgrade, will increase the instantaneous luminosity to about three times the design value. Since the current ATLAS trigger system does not allow sufficient increase of the trigger rate, an improvement of the trigger system is required. The Liquid Argon (LAr) Calorimeter read-out will therefore be modified to use digital trigger signals with a higher spatial granularity in order to improve the identification efficiencies of electrons, photons, tau, jets and missing energy, at high background rejection rates at the Level-1 trigger. The new trigger signals will be arranged in 34000 so-called Super Cells which achieves 5-10 times better granularity than the trigger towers currently used and allows an improved background rejection. The readout of the trigger signals will process the signal of the Super Cells at every LHC bunch-crossing at 12-bit precision and a frequency of 40 MHz. The data will...

Safety analysis of reactor's cooling system

International Nuclear Information System (INIS)

1999-01-01

Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

Gaseous fuel reactors for power systems

Science.gov (United States)

Kendall, J. S.; Rodgers, R. J.

1977-01-01

Gaseous-fuel nuclear reactors have significant advantages as energy sources for closed-cycle power systems. The advantages arise from the removal of temperature limits associated with conventional reactor fuel elements, the wide variety of methods of extracting energy from fissioning gases, and inherent low fissile and fission product in-core inventory due to continuous fuel reprocessing. Example power cycles and their general performance characteristics are discussed. Efficiencies of gaseous fuel reactor systems are shown to be high with resulting minimal environmental effects. A technical overview of the NASA-funded research program in gaseous fuel reactors is described and results of recent tests of uranium hexafluoride (UF6)-fueled critical assemblies are presented.

Review of Operation and Maintenance Support Systems for Research Reactors

International Nuclear Information System (INIS)

Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

2014-01-01

Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

Nuclear reactor power supply system

International Nuclear Information System (INIS)

Cook, B.M.

1982-01-01

The redundant signals from the sensor assemblies measuring the process parameters of a nuclear reactor power supply are transmitted each in its turn to a protection system which operates to actuate the protection apparatus for signals indicating off-process conditions. Each sensor assembly includes a number of like sensors measuring the same parameters. The sets of process signals derived from the sensor assemblies are each in its turn transmitted from the protection system to the control system which impresses control signals on the reactor or its components to counteract the tendency for conditions to drift off-normal status requiring operation of the protection system. A parameter signal selector prevents a parameter signal which differs from the other parameter signals of the set by more than twice the allowable variation from passing to the control system. Test signals are periodically impressed by a test unit on a selected pair of a selection unit and control channels. This arrangement eliminates the possibility that a single component failure which may be spurious will cause an inadvertent trip of the reactor during test. (author)

Limit regulation system for pressurized water nuclear reactors

International Nuclear Information System (INIS)

Aleite, W.; Bock, H.W.

1976-01-01

Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

Reactor protection systems of 500 MWe PHWRs

Energy Technology Data Exchange (ETDEWEB)

Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

1997-03-01

The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

DarkSide-20k: A 20 tonne two-phase LAr TPC for direct dark matter detection at LNGS

Science.gov (United States)

Aalseth, C. E.; Acerbi, F.; Agnes, P.; Albuquerque, I. F. M.; Alexander, T.; Alici, A.; Alton, A. K.; Antonioli, P.; Arcelli, S.; Ardito, R.; Arnquist, I. J.; Asner, D. M.; Ave, M.; Back, H. O.; Barrado Olmedo, A. I.; Batignani, G.; Bertoldo, E.; Bettarini, S.; Bisogni, M. G.; Bocci, V.; Bondar, A.; Bonfini, G.; Bonivento, W.; Bossa, M.; Bottino, B.; Boulay, M.; Bunker, R.; Bussino, S.; Buzulutskov, A.; Cadeddu, M.; Cadoni, M.; Caminata, A.; Canci, N.; Candela, A.; Cantini, C.; Caravati, M.; Cariello, M.; Carlini, M.; Carpinelli, M.; Castellani, A.; Catalanotti, S.; Cataudella, V.; Cavalcante, P.; Cavuoti, S.; Cereseto, R.; Chepurnov, A.; Cicalò, C.; Cifarelli, L.; Citterio, M.; Cocco, A. G.; Colocci, M.; Corgiolu, S.; Covone, G.; Crivelli, P.; D'Antone, I.; D'Incecco, M.; D'Urso, D.; Da Rocha Rolo, M. D.; Daniel, M.; Davini, S.; de Candia, A.; De Cecco, S.; De Deo, M.; De Filippis, G.; De Guido, G.; De Rosa, G.; Dellacasa, G.; Della Valle, M.; Demontis, P.; Derbin, A.; Devoto, A.; Di Eusanio, F.; Di Pietro, G.; Dionisi, C.; Dolgov, A.; Dormia, I.; Dussoni, S.; Empl, A.; Fernandez Diaz, M.; Ferri, A.; Filip, C.; Fiorillo, G.; Fomenko, K.; Franco, D.; Froudakis, G. E.; Gabriele, F.; Gabrieli, A.; Galbiati, C.; Garcia Abia, P.; Gendotti, A.; Ghisi, A.; Giagu, S.; Giampa, P.; Gibertoni, G.; Giganti, C.; Giorgi, M. A.; Giovanetti, G. K.; Gligan, M. L.; Gola, A.; Gorchakov, O.; Goretti, A. M.; Granato, F.; Grassi, M.; Grate, J. W.; Grigoriev, G. Y.; Gromov, M.; Guan, M.; Guerra, M. B. B.; Guerzoni, M.; Gulino, M.; Haaland, R. K.; Hallin, A.; Harrop, B.; Hoppe, E. W.; Horikawa, S.; Hosseini, B.; Hughes, D.; Humble, P.; Hungerford, E. V.; Ianni, An.; Jillings, C.; Johnson, T. N.; Keeter, K.; Kendziora, C. L.; Kim, S.; Koh, G.; Korablev, D.; Korga, G.; Kubankin, A.; Kuss, M.; Kuźniak, M.; La Commara, M.; Lehnert, B.; Li, X.; Lissia, M.; Lodi, G. U.; Loer, B.; Longo, G.; Loverre, P.; Lussana, R.; Luzzi, L.; Ma, Y.; Machado, A. A.; Machulin, I. N.; Mandarano, A.; Mapelli, L.; Marcante, M.; Margotti, A.; Mari, S. M.; Mariani, M.; Maricic, J.; Martoff, C. J.; Mascia, M.; Mayer, M.; McDonald, A. B.; Messina, A.; Meyers, P. D.; Milincic, R.; Moggi, A.; Moioli, S.; Monroe, J.; Monte, A.; Morrocchi, M.; Mount, B. J.; Mu, W.; Muratova, V. N.; Murphy, S.; Musico, P.; Nania, R.; Navrer Agasson, A.; Nikulin, I.; Nosov, V.; Nozdrina, A. O.; Nurakhov, N. N.; Oleinik, A.; Oleynikov, V.; Orsini, M.; Ortica, F.; Pagani, L.; Pallavicini, M.; Palmas, S.; Pandola, L.; Pantic, E.; Paoloni, E.; Paternoster, G.; Pavletcov, V.; Pazzona, F.; Peeters, S.; Pelczar, K.; Pellegrini, L. A.; Pelliccia, N.; Perotti, F.; Perruzza, R.; Pesudo, V.; Piemonte, C.; Pilo, F.; Pocar, A.; Pollmann, T.; Portaluppi, D.; Pugachev, D. A.; Qian, H.; Radics, B.; Raffaelli, F.; Ragusa, F.; Razeti, M.; Razeto, A.; Regazzoni, V.; Regenfus, C.; Reinhold, B.; Renshaw, A. L.; Rescigno, M.; Retière, F.; Riffard, Q.; Rivetti, A.; Rizzardini, S.; Romani, A.; Romero, L.; Rossi, B.; Rossi, N.; Rubbia, A.; Sablone, D.; Salatino, P.; Samoylov, O.; Sánchez García, E.; Sands, W.; Sanfilippo, S.; Sant, M.; Santorelli, R.; Savarese, C.; Scapparone, E.; Schlitzer, B.; Scioli, G.; Segreto, E.; Seifert, A.; Semenov, D. A.; Shchagin, A.; Shekhtman, L.; Shemyakina, E.; Sheshukov, A.; Simeone, M.; Singh, P. N.; Skensved, P.; Skorokhvatov, M. D.; Smirnov, O.; Sobrero, G.; Sokolov, A.; Sotnikov, A.; Speziale, F.; Stainforth, R.; Stanford, C.; Suffritti, G. B.; Suvorov, Y.; Tartaglia, R.; Testera, G.; Tonazzo, A.; Tosi, A.; Trinchese, P.; Unzhakov, E. V.; Vacca, A.; Vázquez-Jáuregui, E.; Verducci, M.; Viant, T.; Villa, F.; Vishneva, A.; Vogelaar, B.; Wada, M.; Wahl, J.; Walding, J.; Wang, H.; Wang, Y.; Watson, A. W.; Westerdale, S.; Williams, R.; Wojcik, M. M.; Wu, S.; Xiang, X.; Xiao, X.; Yang, C.; Ye, Z.; Yllera de Llano, A.; Zappa, F.; Zappalà, G.; Zhu, C.; Zichichi, A.; Zullo, M.; Zullo, A.; Zuzel, G.

2018-03-01

Building on the successful experience in operating the DarkSide-50 detector, the DarkSide Collaboration is going to construct DarkSide-20k, a direct WIMP search detector using a two-phase Liquid Argon Time Projection Chamber (LAr TPC) with an active (fiducial) mass of 23 t (20 t). This paper describes a preliminary design for the experiment, in which the DarkSide-20k LAr TPC is deployed within a shield/veto with a spherical Liquid Scintillator Veto (LSV) inside a cylindrical Water Cherenkov Veto (WCV). This preliminary design provides a baseline for the experiment to achieve its physics goals, while further development work will lead to the final optimization of the detector parameters and an eventual technical design. Operation of DarkSide-50 demonstrated a major reduction in the dominant 39Ar background when using argon extracted from an underground source, before applying pulse shape analysis. Data from DarkSide-50, in combination with MC simulation and analytical modeling, shows that a rejection factor for discrimination between electron and nuclear recoils of >3 × 109 is achievable. This, along with the use of the veto system and utilizing silicon photomultipliers in the LAr TPC, are the keys to unlocking the path to large LAr TPC detector masses, while maintaining an experiment in which less than < 0.1 events (other than ν-induced nuclear recoils) is expected to occur within the WIMP search region during the planned exposure. DarkSide-20k will have ultra-low backgrounds than can be measured in situ, giving sensitivity to WIMP-nucleon cross sections of 1.2 × 10^{-47} cm2 (1.1 × 10^{-46} cm2) for WIMPs of 1 TeV/c 2 (10 TeV/c 2) mass, to be achieved during a 5 yr run producing an exposure of 100 t yr free from any instrumental background.

Development of an automatic reactor inspection system

International Nuclear Information System (INIS)

Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

2002-02-01

Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

Altın Fiyatlarının Yapay Sinir Ağları ile Tahmini ve Bir Uygulama

Directory of Open Access Journals (Sweden)

Rıdvan YÜKSEL

2016-03-01

Full Text Available Bu çalışmada altın fiyatlarını yapay sinir ağları ile öngörmek amacıyla, altın fiyatlarını etkileyebileceği düşünülen değişkenler olan Gümüş fiyatları, Brent Petrol fiyatları, ABD doları/ EUR paritesi, EuroNext100 endeksi, Amerika Dow Jones Endeksi, 13 Hafta vadeli ABD bonosu faiz oranı ve ABD TÜFE endeksi kullanılarak modeller kurulmuştur. Yapay sinir ağları ile kurulan modellerden elde edilen tahmin sonuçları, gerçek değerler ile R2, RMSE, MAE ve MAPE (% gibi performans kriterleri hesaplanarak karşılaştırılmıştır. Elde edilen bulgular yapay sinir ağlarının altın fiyatlarının tahmininde başarı ile kullanılabileceğini göstermektedir. Yapılan duyarlılık analizinin sonuçları değerlendirildiğinde altın fiyatlarını etkileyen faktörlerin başında gümüş ve petrol fiyatlarının geldiği tespit edilmiştir.

Genetic association analysis of LARS2 with type 2 diabetes

NARCIS (Netherlands)

E. Reiling (Erwin); B. Jafar-Mohammadi (Bahram); E. van 't Riet (Esther); M.N. Weedon (Michael); J.V. van Vliet-Ostaptchouk (Jana); T. Hansen (Torben); R. Saxena (Richa); T.W. van Haeften (Timon); P.P. Arp (Pascal); S. Das; M.G.A.A.M. Nijpels (Giel); M.J. Groenewoud (Marlous); E.C. van Hove (Els); A.G. Uitterlinden (André); J.W.A. Smit (Jan); A.D. Morris (Andrew); A.S.F. Doney (Alex); C.N.A. Palmer (Colin); C. Guiducci (Candace); A.T. Hattersley (Andrew); T.M. Frayling (Timothy); O. Pedersen (Oluf); P.E. Slagboom (Eline); D. Altshuler (David); L. Groop (Leif); J.A. Romijn; J.A. Maassen (Johannes); M.A. Hofker (Marten); J.M. Dekker (Jacqueline); M.I. McCarthy (Mark); L.M. 't Hart (Leen)

2010-01-01

textabstractAims/hypothesis: LARS2 has been previously identified as a potential type 2 diabetes susceptibility gene through the low-frequency H324Q (rs71645922) variant (minor allele frequency [MAF] 3.0%). However, this association did not achieve genome-wide levels of significance. The aim of this

Expert system for fast reactor diagnostic

International Nuclear Information System (INIS)

Parcy, J.P.

1982-09-01

A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented [fr

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Energy Technology Data Exchange (ETDEWEB)

Neil Todreas; Pavel Hejzlar

2008-06-30

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

Flexible Conversion Ratio Fast Reactor Systems Evaluation

International Nuclear Information System (INIS)

Neil Todreas; Pavel Hejzlar

2008-01-01

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

Principle of human system interface (HSI) design for new reactor console of PUSPATI TRIGA Reactor (RTP)

International Nuclear Information System (INIS)

Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Mohd Sabri Minhat; Izhar Abu Hussin

2013-01-01

Full-text: This paper will describe the principle of human system interface design for new reactor console in control room at TRIGA reactor facility. In order to support these human system interface challenges in digital reactor console. Software-based instrumentation and control (I and C) system for new reactor console could lead to new human machine integration. The proposed of Human System Interface (HSI) which included the large display panels which shows reactor status, compact and computer-based workstations for monitoring, control and protection function. The proposed Human System Interface (HIS) has been evaluated using various human factor engineering. It can be concluded that the Human System Interface (HIS) is designed as to address the safety related computer controlled system. (author)

Application of expert system to evaluating reactor water cleanup system performance

International Nuclear Information System (INIS)

Maeda, Katsuji; Nakamura, Masahiro; Nagasawa, Katsumi; Fushiki, Sumiyuki.

1991-01-01

Expert systems employing artificial intelligence (AI) have been developed for finding and elucidating causes of anomalies and malfunctions, presenting pertinent recommendation for countermeasures and for making precautionary diagnosis. On the other hand, further improvements in reliabilities for chemical control are required to promote BWR plant reliability and advancement. Especially, it is necessary to maintain the reactor water purity in high quality to minimize stress corrosion cracking (SCC) in primary cooling system, fuel performance degradation and radiation buildup. The reactor water quality is controlled by the reactor water cleanup (RWCU) system. So, it is very important to maintain the RWCU performance, in order to keep good reactor water quality. This paper describes an expert system used for evaluating RWCU system performance in BWR plants. (author)

Integral reactor system and method for fuel cells

Science.gov (United States)

Fernandes, Neil Edward; Brown, Michael S; Cheekatamarla, Praveen; Deng, Thomas; Dimitrakopoulos, James; Litka, Anthony F

2013-11-19

A reactor system is integrated internally within an anode-side cavity of a fuel cell. The reactor system is configured to convert hydrocarbons to smaller species while mitigating the lower production of solid carbon. The reactor system may incorporate one or more of a pre-reforming section, an anode exhaust gas recirculation device, and a reforming section.

Reactivity requirements and safety systems for heavy water reactors

International Nuclear Information System (INIS)

Kati, S.L.; Rustagi, R.S.

1977-01-01

The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

Reactor trip on turbine trip inhibit control system for nuclear power generating system

International Nuclear Information System (INIS)

Torres, J.M.; Musick, C.R.

1976-01-01

A reactor trip on turbine trip inhibit control system for a nuclear power generating system which utilizes steam bypass valves is described. The control system inhibits a normally automatic reactor trip on turbine trip when the bypass valves have the capability of bypassing enough steam to prevent reactor trip limits from being reached and/or to prevent opening of the secondary safety pressure valves. The control system generates a bypass valve capability signal which is continuously compared with the reactor power. If the capability is greater than the reactor power, then an inhibit signal is generated which prevents a turbine trip signal from tripping the nuclear reactor. 10 claims, 4 figures

Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-07-01

Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)

Brüssel Euroopa pealinnaks? / Ülar Mark

Index Scriptorium Estoniae

Mark, Ülar, 1968-

2009-01-01

Brüsseli Euroopa kvartali planeerimisvõistlusest, mille peakorraldaja oli Brüsseli pealinna piirkond koos Brüsseli linna ja Euroopa Komisjoniga. Eestist osales žüriis arhitekt Ülar Mark. Meeskonna Atelier Christian de Portzamparc võidutööst ja teiste teise vooru pääsenud nelja meeskonna (JDS / Julien De Smedt Architects, OMA / Office for Metropolitan Architecture, Xaveer De Geyter Architect, Fletcher Priest Architects) töödest

Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

International Nuclear Information System (INIS)

Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

1992-01-01

A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

Nuclear Reactor RA Safety Report, Vol. 8, Auxiliary system

International Nuclear Information System (INIS)

1986-11-01

This volume describes RA reactor auxiliary systems, as follows: special ventilation system, special drainage system, hot cells, systems for internal transport. Ventilation system is considered as part of the reactor safety and protection system. Its role is eliminate possible radioactive particles dispersion in the environment. Special drainage system includes pipes and reservoirs with the safety role, meaning absorption or storage of possible radioactive waste water from the reactor building. Hot cells existing in the RA reactor building are designed for production of sealed radioactive sources, including packaging and transport [sr

Reactor power system deployment and startup

International Nuclear Information System (INIS)

Wetch, J.R.; Nelin, C.J.; Britt, E.J.; Klein, G.; Rasor Associates, Inc., Sunnyvale, CA; California Institute of Technology, Pasadena)

1985-01-01

This paper addresses issues that should receive further examination in the near-term as concept selection for development of a U.S. space reactor power system is approached. The issues include: the economics, practicality and system reliability associated with transfer of nuclear spacecraft from low earth shuttle orbits to operational orbits, via chemical propulsion versus nuclear electric propulsion; possible astronaut supervised reactor and nuclear electric propulsion startup in low altitude Shuttle orbit; potential deployment methods for nuclear powered spacecraft from Shuttle; the general public safety of low altitude startup and nuclear safe and disposal orbits; the question of preferred reactor power level; and the question of frozen versus molten alkali metal coolant during launch and deployment. These issues must be considered now because they impact the SP-100 concept selection, power level selection, weight and size limits, use of deployable radiators, reliability requirements, and economics, as well as the degree of need for and the urgency of developing space reactor power systems. 5 references

Gaseous fuel reactors for power systems

International Nuclear Information System (INIS)

Helmick, H.H.; Schwenk, F.C.

1978-01-01

The Los Alamos Scientific Laboratory is participating in a NASA-sponsored program to demonstrate the feasibility of a gaseous uranium fueled reactor. The work is aimed at acquiring experimental and theoretical information for the design of a prototype plasma core reactor which will test heat removal by optical radiation. The basic goal of this work is for space applications, however, other NASA-sponsored work suggests several attractive applications to help meet earth-bound energy needs. Such potential benefits are small critical mass, on-site fuel processing, high fuel burnup, low fission fragment inventory in reactor core, high temperature for process heat, optical radiation for photochemistry and space power transmission, and high temperature for advanced propulsion systems. Low power reactor experiments using uranium hexafluoride gas as fuel demonstrated performance in accordance with reactor physics predictions. The final phase of experimental activity now in progress is the fabrication and testing of a buffer gas vortex confinement system

Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

International Nuclear Information System (INIS)

Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

2013-01-01

RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

Reactor technology: power conversion systems and reactor operation and maintenance

International Nuclear Information System (INIS)

Powell, J.R.

1977-01-01

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

Reactor Shutdown Mechanism by Top-mounted Hydraulic System

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Haun; Cho, Yeong Garp; Choi, Myoung Hwan; Lee, Jin Haeng; Huh, Hyung; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

There are two types of reactor shutdown mechanisms in HANARO. One is the mechanism driven by a hydraulic system, and the other is driven by a stepping motor. In HANARO, there are four Control Rod Drive Mechanisms (CRDMs) with an individual step motor and four Shutoff (SO) Units with an individual hydraulic system located at the top of reactor pool. The absorber rods in SO units are poised at the top of the core by the hydraulic force during normal operation. The rods of SO units drop by gravity as the first reactor showdown mechanism when a trip is commended by the reactor protection system (RPS). The rods in CRDMs also drop by gravity together as a redundant shutdown mechanism. When a trip is commended by the reactor regulating system (RRS), the absorber rods of CRDM only drop; while the absorber rods of SO units stay at the top of the core by the hydraulic system. The reactivity control mechanisms of in JRTR, one of the new research reactor with plate type fuels, consist of four CRDMs driven by an individual step motor and two second shutdown drive mechanisms (SSDMs) driven by an individual hydraulic system as shown in Fig. 1. The CRDMs act as the first reactor shutdown mechanism and reactor regulating as well. The top-mounted SSDM driven by the hydraulic system for the JRTR is under design in KAERI. The SSDM provides an alternate and independent means of reactor shutdown. The second shutdown rods (SSRs) of the SSDM are poised at the top of the core by the hydraulic system during the normal operation and drop by gravity for the reactor trip. Based on the proven technology of the design, operation and maintenance for HANARO, the SSDM for the JRTR has been optimized by the design improvement from the experience and test. This paper aims for the introduction of the SSDM in the process of the basic design. The major differences of the shutdown mechanisms by the hydraulic system are compared between HANARO and JRTR, and the design features, system, structure and

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

Distributed expert systems for nuclear reactor control

International Nuclear Information System (INIS)

Otaduy, P.J.

1992-01-01

A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

Seismic attenuation system for a nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Liszkai, Tamas; Cadell, Seth

2018-01-30

A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

Study on secondary shutdown systems in Tehran research reactor

Energy Technology Data Exchange (ETDEWEB)

Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

2015-09-15

Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

Cryogenic system design for a compact tokamak reactor

International Nuclear Information System (INIS)

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

1988-01-01

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

CRNL research reactor retrofit Emergency Filtration System

International Nuclear Information System (INIS)

Philippi, H.M.

1990-01-01

This paper presents a brief history of NRX and NRU research reactor effluent air treatment systems before describing the selection and design of an appropriate retrofit Emergency Filtration System (EFS) to serve these reactors and the future MX-10 isotope production reactor. The conceptual design of the EFS began in 1984. A standby concrete shielding filter-adsorber system, sized to serve the reactor with the largest exhaust flow, was selected. The standby system, bypassed under normal operating conditions, is equipped with normal exhaust stream shutoff and diversion valves to be activated manually when an emergency is anticipated, or automatically when emergency levels of gamma radiation are detected in the exhaust stream. The first phase of the EFS installation, that is the construction of the EFS and the connection of NRU to the system, was completed in 1987. The second phase of construction, which includes the connection of NRX and provisions for the future connection of MX-10, is to be completed in 1990

Safety system for reactor container

International Nuclear Information System (INIS)

Shimizu, Miwako; Seki, Osamu; Mano, Takio.

1995-01-01

A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

Thermionic nuclear reactor systems

International Nuclear Information System (INIS)

Kennel, E.B.

1986-01-01

Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

Micro processor based research reactor instrumentation and control system

International Nuclear Information System (INIS)

Hyde, W.K.

1987-01-01

The system consists of a Control System Computer (CSC) incorporated into a Reactor Control Console (RCC) and a Data Acquisition and Control Unit (DAC) adjacent to the reactor. The CSC has a high resolution color graphics CRT monitor which provides real-time graphic simulation of the reactor and a number of bar graphs displaying strategic parameters of the reactor system. In addition, abnormal or dangerous conditions are displayed. The CSC is equipped with two printers eliminating manual logging of reactor data. The reactor display and pulse mode display may also be printed. Historical data is saved in the system's large capacity memory and may be replayed and/or printed. Because of the CSC's inherent high speed math capability, raw reactor data will be quickly converted and displayed in real-time. Data can be presented in meaningful engineering units. The DAC provides a high speed data acquisition and control capability adjacent to the reactor. It continuously collects data from the reactor system, concentrates the data into a database and transmits it to the CSC when requested. Data transmission is over one of two data trunks to the CSC. The secondary trunk is used if the primary trunk fails. The data trunks drastically reduce the wiring requirements between the reactor and the Control Console. During steady-state operation of the reactor, operator commands to adjust the rod positions is transmitted from the CSC to the DAC which re-issues the commands to the drive mechanisms. In the automatic mode, the DAC will control the position of the rods via a PID algorithm. The system is independently monitored by two or more safety computers. Their function is to monitor the power level, the rate of change of power and fuel temperature of the reactor and to independently shut the reactor down in the event of a potentially dangerous (scram) condition. (author)

Materials degradation in fission reactors: Lessons learned of relevance to fusion reactor systems

International Nuclear Information System (INIS)

Was, Gary S.

2007-01-01

The management of materials in power reactor systems has become a critically important activity in assuring the safe, reliable and economical operation of these facilities. Over the years, the commercial nuclear power reactor industry has faced numerous 'surprises' and unexpected occurrences in materials. Mitigation strategies have sometimes solved one problem at the expense of creating another. Other problems have been solved successfully and have motivated the development of techniques to foresee problems before they occur. This paper focuses on three aspects of fission reactor experience that may benefit future fusion systems. The first is identification of parameters and processes that have had a large impact on the behavior of materials in fission systems such as temperature, dose rate, surface condition, gradients, metallurgical variability and effects of the environment. The second is the development of materials performance and failure models to provide a basis for assuring component integrity. Last is the development of proactive materials management programs that identify and pre-empt degradation processes before they can become problems. These aspects of LWR experience along with the growing experience with materials in the more demanding advanced fission reactor systems form the basis for a set of 'lessons learned' to aid in the successful management of materials in fusion reactor systems

Containment atmosphere cooling system for experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

1979-01-01

The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

Nuclear reactor monitoring system

International Nuclear Information System (INIS)

Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

1976-01-01

The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

Choice of thermal reactor systems: a report

Energy Technology Data Exchange (ETDEWEB)

1977-09-01

This is a report by the UK National Nuclear Corporation published by the UK Secretary of State for Energy (Mr. Benn) on 29th July 1977. It is concerned with the advantages and disadvantages of three thermal reactor systems -the AGR (advanced gas cooled reactor), the PWR (pressurised water reactor), and the SGHWR (steam generating heavy water reactor). The object was to help in the future choice of a thermal system for the UK to cover the next 25 years. The matter of export potential is also considered. A programme of four stations of 1100 to 1300 MW each over six years starting from 1979 was assumed. It is emphasised that a decision must be taken now both about reactor systems and actual orders. Headings are as follows: Extract from conclusions reached; Summary of main features of assessment; General conclusions regarding the following - safety, security of the investment, operational characteristics, development and launching requirements, effect on industry, and capital and generation costs. It is stated that in order to make an overall judgement on reactor choice the technical, commercial and social issues involved must be weighed in conjunction with cost differentials.

Peynirlere Kontamine Olan Küflerin Bazı Esansiyel Yağlar ile İnhibisyonu

Directory of Open Access Journals (Sweden)

Sibel Özçakmak

2015-02-01

Full Text Available Küf kontaminasyonu önemli kalite problemlerine ve peynir üreticilerine ekonomik kayıplara neden olmaktadır. Peynirlerden yaygın olarak izole edilen kontamine küf türleri Penicillium commune, P. verrucosum, P. roqueforti, P. palitans, Aspergillus flavus ve Geotrichum candidum olarak belirlenmiştir. Sorbik asit veya bunların tuzları, bozulma etmeni küflerin önlenmesi için peynir yüzeylerine uygulanmaktadır. Bazı sentetik gıda katkılarının kullanımıyla ilgili gıda endüstrisi ve mevzuatı oluşturan kuruluşlar tarafından getirilen kısıtlamalar doğal antimikrobiyel bileşikler üzerine araştırmaların doğmasına neden olmuştur. Bitki ve baharat, gıdaların mikrobiyolojik güvenliğini sağlayan etkili bileşiklere sahiptir. Salvia officinalis (ada çayı, Laurus nobilis (defne, Cinnamomum zeylanicum (tarçın, Thymus vulgaris (kekik, Cymbopogon citratus (limon otu, Origanum vulgare (keklik otu esansiyel yağları, bozulma etmeni küf şorasını inhibe edebilmiştir. Thymol ve karanfil esansiyel yağlarının Çedar peynirinde P. citrinum, Kaşar peynirinde A. flavus; zeytin yaprağı ekstraklarının, Tulum peynirinde A. niger, P. citrinum ve P. roqueforti, beyaz peynirde Mucor rocemosus’un gelişimini inhibe edebildiği bildirilmiştir.

ATLAS LAr Calorimeter Performance and Commissioning for LHC Run-2

CERN Document Server

Spettel, Fabian; The ATLAS collaboration

2015-01-01

The ATLAS detector was designed and built to study proton-proton colli- sions produced at the LHC at centre-of-mass energies up to 14 TeV and in- stantaneous luminosities up to $10^{34} \\text{cm}^{-2} \\text{s}^{-1}$. Liquid argon (LAr) sampling calorimeters are employed for all electromagnetic calorimetry in the pseudorapidity region $|\\eta|<3.2$, and for hadronic calorimetry in the region from $|\\eta|=1.5$ to $|\\eta|=4.9$. In the first LHC run a total luminosity of 27 $\\text{fb}^{-1}$ as been collected at center-of-mass energies of 7-8 TeV with very high operational efficiency of the LAr Calorimeters and excellent performance. The well calibrated and highly granular detector achieved its design values both in energy measurement as well as in direction resolution, which was a main ingredient for the successul discovery of a Higgs boson in the di-photon decay channel. The talk will give an overview of the procedures applied to calibrate the 180.000 read-out channels electronically as well as from using refe...

Power reactor information system (PRIS)

International Nuclear Information System (INIS)

1989-06-01

Since the very beginning of commercial operation of nuclear power plants, the nuclear power industry worldwide has accumulated more than 5000 reactor years of experience. The IAEA has been collecting Operating Experience data for Nuclear Power Plants since 1970 which were computerized in 1980. The Agency has undertaken to make Power Reactor Information System (PRIS) available on-line to its Member States. The aim of this publication is to provide the users of PRIS from their terminals with description of data base and communication systems and to show the methods of accessing the data

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

International Nuclear Information System (INIS)

Hohorst, J.K.; Allison, C.M.

2010-01-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

Energy Technology Data Exchange (ETDEWEB)

Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)

2010-07-01

The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)

Annexes to the lecture on reactor protection system including engineered features actuation system

International Nuclear Information System (INIS)

Palmaers, W.

1982-01-01

The present paper deals with the fundamentals for a reactor protection system and discusses the following topics: - System lay-out - Analog measured data acquisition - Analog measured data processing - Limit value generation and logical gating - Procesing of the reactor protection actuation signals - Decoupling of the reactor protection system - Mechanical lay-out - Monitoring system and - Emergency control station. (orig./RW)

A computer control system for a research reactor

International Nuclear Information System (INIS)

Crawford, K.C.; Sandquist, G.M.

1987-01-01

Most reactor applications until now, have not required computer control of core output. Commercial reactors are generally operated at a constant power output to provide baseline power. However, if commercial reactor cores are to become load following over a wide range, then centralized digital computer control is required to make the entire facility respond as a single unit to continual changes in power demand. Navy and research reactors are much smaller and simpler and are operated at constant power levels as required, without concern for the number of operators required to operate the facility. For navy reactors, centralized digital computer control may provide space savings and reduced personnel requirements. Computer control offers research reactors versatility to efficiently change a system to develop new ideas. The operation of any reactor facility would be enhanced by a controller that does not panic and is continually monitoring all facility parameters. Eventually very sophisticated computer control systems may be developed which will sense operational problems, diagnose the problem, and depending on the severity of the problem, immediately activate safety systems or consult with operators before taking action

The WA105-3x1x1 m3 dual phase LAr-TPC demonstrator

CERN Document Server

Murphy, Sebastien

2016-11-15

The dual phase Liquid Argon Time Projection Chamber (LAr TPC) is the state-of-art technology for neutrino detection thanks to its superb 3D tracking and calorimetry performance. Its main feature is the charge amplification in gas argon which provides excellent signal-to-noise ratio. Electrons produced in the liquid argon are extracted in the gas phase. Here, a readout plane based on Large Electron Multiplier detectors provides amplification of the charges before its collection onto an anode with strip readout. The charge amplification enables constructing fully homoge- nous giant LAr-TPCs with tuneable gain, excellent charge imaging performance and increased sensitivity to low energy events. Following a staged approach the WA105 collaboration is con- structing a dual phase LAr-TPC with an active volume of 3x1x1m3 that will soon be tested with cosmic rays. Its construction and operation aims to test scalable solutions for the crucial aspects of this technology: ultra high argon purity in non-evacuable tank, la...

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1978-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results presented in this paper can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

Plasma driving system requirements for commercial tokamak fusion reactors

International Nuclear Information System (INIS)

Brooks, J.N.; Kustom, R.C.; Stacey, W.M. Jr.

1977-01-01

The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration

SRAC: JAERI thermal reactor standard code system for reactor design and analysis

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

1983-01-01

The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

Radiolytic production of chemical fuels in fusion reactor systems

Energy Technology Data Exchange (ETDEWEB)

Fish, J D

1977-06-01

Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered.

Radiolytic production of chemical fuels in fusion reactor systems

International Nuclear Information System (INIS)

Fish, J.D.

1977-06-01

Miley's energy flow diagram for fusion reactor systems is extended to include radiolytic production of chemical fuel. Systematic study of the economics and the overall efficiencies of fusion reactor systems leads to a criterion for evaluating the potential of radiolytic production of chemical fuel as a means of enhancing the performance of a fusion reactor system. The ecumenicity of the schema is demonstrated by application to (1) tokamaks, (2) mirror machines, (3) theta-pinch reactors, (4) laser-heated solenoids, and (5) inertially confined, laser-pellet devices. Pure fusion reactors as well as fusion-fission hybrids are considered

Cable handling system for use in a nuclear reactor

International Nuclear Information System (INIS)

Crosgrove, R.O.; Larson, E.M.; Moody, E.

1982-01-01

A cable handling system for use in an installation such as a nuclear reactor is disclosed herein along with relevant portions of the reactor which, in a preferred embodiment, is a liquid metal fast breeder reactor. The cable handling system provides a specific way of interconnecting certain internal reactor components with certain external components, through an assembly of rotatable plugs. Moreover, this is done without having to disconnect these components from one another during rotation of the plugs and yet without interfering with other reactor components in the vicinity of the rotating plugs and cable handling system

Power conditioning system for a nuclear reactor

International Nuclear Information System (INIS)

Higashigawa, Yuichi; Joge, Toshio.

1981-01-01

Purpose: To provide a power conditioning system for a BWR type reactor which has a function to be automatically operated within a range that the relationship between the heat power of the reactor and the electric power of an electric generator does not lose the safety of fuel by eliminating the unnecessary fluctuation of the power of the reactor. Constitution: A load request error signal fed from a conventional turbine control system to recirculation flow regulator is eliminated, and a reactor power conditioning system is newly provided, to which an electric generator power signal, a reactor average power area monitor signal and a load request signal are inputted. Thus, the load request signal is compared directly with the electric power of the electric generator, the recirculation flow rate is controlled by the compared result, and whether the correlation between the heat power of the reqctor and the electric power of the generator satisfies the correlation determined to prove the safety of fuel or not is checked. If this correlation is satisfied, the recirculation flow rate is merely automatically controlled. (Yoshino, Y.)

Development of multi-functional telerobotic systems for reactor dismantlement

International Nuclear Information System (INIS)

Fujii, Yoshio; Usui, Hozumi; Shinohara, Yoshikuni

1992-01-01

This report summarizes technological features of advanced telerobotic systems for reactor dismantling application developed at the Japan Atomic Energy Research Institute. Taking into consideration the special environmental conditions in reactor dismantling, major effort was made to develop multifunctional telerobotic system of high reliability which can be used to perform various complex tasks in an unstructured environment and operated in an easy and flexible manner. The system development was carried out through constructing three systems in seccession; a light-duty and a heavy-duty system as a prototype system for engineering test in cold environment, and a demonstration system for practical on-site application to dismantling highly radioactive reactor internals of an experimental boiling water reactor JPDR (Japan Power Demonstration Reactor). Each system was equipped with one or two amphibious manipulators which can be operated in either a push-button manual, a bilateral master-slave, a teach-and-playback or a programmed control mode. Different scheme was adopted in each system at designing the manipulator, transporter and man-machine interface so as to compare their advantages and disadvantages. According to the JPDR decommissioning program, the demonstration system was successfully operated to dismantle a portion of the radioactive reactor internals of the JPDR, which used underwater plasma arc cutting method and proved the usefulness of the multi-functional telerobotic system for reducing the occupational hazards and enhancing the work efficiency in the course of dismantling highly radioactive reactor components. (author)

Management system requirements for small reactors

Energy Technology Data Exchange (ETDEWEB)

Jones, K.A., E-mail: kenneth.jones@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2013-07-01

This abstract identifies the management system requirements for the life cycle of small reactors from initial conception through completion of decommissioning. For small reactors, the requirements for management systems remain the same as those for 'large' reactors regardless of the licensee' business model and objectives. The CSA N-Series of standards provides an interlinked set of requirements for the management of nuclear facilities. CSA N286 provides overall direction to management to develop and implement sound management practices and controls, while other CSA nuclear standards provide technical requirements and guidance that support the management system. CSA N286 is based on a set of principles. The principles are then supported by generic requirements that are applicable to the life cycle of nuclear facilities. CNSC regulatory documents provide further technical requirements and guidance. (author)

Microchannel Reactor System for Catalytic Hydrogenation

Energy Technology Data Exchange (ETDEWEB)

Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

2010-12-22

We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

Creation of reactor's reliable system of emergency energy supply

International Nuclear Information System (INIS)

Batyrbekov, G.A.; Brovkin, A.Yu.; Petukhov, V.K.; Chekushin, A.I.; Chernyaev, V.P.; Yagotinets, N.A.

1998-01-01

System of reliable power supply of the WWR-K reactor complex is described, which completely provides safety operation of reactor equipment in the case of total voltage loss from external power transmission lines as well as under destruction of accumulation batteries by earthquake more than 6 balls. Switching on in operation of diesel-generators and system of constant current supply from accumulator batteries is occurred automatically under cessation of voltage supply from centralized power system. Reliable reactor dampening in case it work on capacity has been ensured. Reactor cooling under its emergency shutdown during both the partial or the total loss of coolant in first counter has been carried out. Under full coolant loss the system of emergency reactor cooling has been switched on in operation

The combined hybrid system: A symbiotic thermal reactor/fast reactor system for power generation and radioactive waste toxicity reduction

International Nuclear Information System (INIS)

Hollaway, W.R.

1991-08-01

If there is to be a next generation of nuclear power in the United States, then the four fundamental obstacles confronting nuclear power technology must be overcome: safety, cost, waste management, and proliferation resistance. The Combined Hybrid System (CHS) is proposed as a possible solution to the problems preventing a vigorous resurgence of nuclear power. The CHS combines Thermal Reactors (for operability, safety, and cost) and Integral Fast Reactors (for waste treatment and actinide burning) in a symbiotic large scale system. The CHS addresses the safety and cost issues through the use of advanced reactor designs, the waste management issue through the use of actinide burning, and the proliferation resistance issue through the use of an integral fuel cycle with co-located components. There are nine major components in the Combined Hybrid System linked by nineteen nuclear material mass flow streams. A computer code, CHASM, is used to analyze the mass flow rates CHS, and the reactor support ratio (the ratio of thermal/fast reactors), IFR of the system. The primary advantages of the CHS are its essentially actinide-free high-level radioactive waste, plus improved reactor safety, uranium utilization, and widening of the option base. The primary disadvantages of the CHS are the large capacity of IFRs required (approximately one MW e IFR capacity for every three MW e Thermal Reactor) and the novel radioactive waste streams produced by the CHS. The capability of the IFR to burn pure transuranic fuel, a primary assumption of this study, has yet to be proven. The Combined Hybrid System represents an attractive option for future nuclear power development; that disposal of the essentially actinide-free radioactive waste produced by the CHS provides an excellent alternative to the disposal of intact actinide-bearing Light Water Reactor spent fuel (reducing the toxicity based lifetime of the waste from roughly 360,000 years to about 510 years)

Computer System Analysis for Decommissioning Management of Nuclear Reactor

International Nuclear Information System (INIS)

Nurokhim; Sumarbagiono

2008-01-01

Nuclear reactor decommissioning is a complex activity that should be planed and implemented carefully. A system based on computer need to be developed to support nuclear reactor decommissioning. Some computer systems have been studied for management of nuclear power reactor. Software system COSMARD and DEXUS that have been developed in Japan and IDMT in Italy used as models for analysis and discussion. Its can be concluded that a computer system for nuclear reactor decommissioning management is quite complex that involved some computer code for radioactive inventory database calculation, calculation module on the stages of decommissioning phase, and spatial data system development for virtual reality. (author)

Safety Analysis for Power Reactor Protection System

International Nuclear Information System (INIS)

Eisawy, E.A.; Sallam, H.

2012-01-01

The main function of a Reactor Protection System (RPS) is to safely shutdown the reactor and prevents the release of radioactive materials. The purpose of this paper is to present a technique and its application for used in the analysis of safety system of the Nuclear Power Plant (NPP). A more advanced technique has been presented to accurately study such problems as the plant availability assessments and Technical Specifications evaluations that are becoming increasingly important. The paper provides the Markov model for the Reactor Protection System of the NPP and presents results of model evaluations for two testing policies in technical specifications. The quantification of the Markov model provides the probability values that the system will occupy each of the possible states as a function of time.

Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

International Nuclear Information System (INIS)

Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

2003-01-01

The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

Sampling system for a boiling reactor NPP

International Nuclear Information System (INIS)

Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

1976-01-01

Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

Fault-tolerant reactor protection system

International Nuclear Information System (INIS)

Gaubatz, D.C.

1997-01-01

A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Each division performs independently of the others (asynchronous operation). All communications between the divisions are asynchronous. Each chassis substitutes its own spare sensor reading in the 2/3 vote if a sensor reading from one of the other chassis is faulty or missing. Therefore the presence of at least two valid sensor readings in excess of a set point is required before terminating the output to the hardware logic of a scram inhibition signal even when one of the four sensors is faulty or when one of the divisions is out of service. 16 figs

Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

International Nuclear Information System (INIS)

Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

1991-01-01

The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

Passive Decay Heat Removal System for Micro Modular Reactor

Energy Technology Data Exchange (ETDEWEB)

Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

2015-10-15

Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

New measuring and protection system at VR-1 training reactor

International Nuclear Information System (INIS)

Kropik, M.; Jurickova, M.

2006-01-01

The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

Emergency reactor cooling systems for the experimental VHTR

International Nuclear Information System (INIS)

Mitake, Susumu; Suzuki, Katsuo; Miyamoto, Yoshiaki; Tamura, Kazuo; Ezaki, Masahiro.

1983-03-01

Performances and design of the panel cooling system which has been proposed to be equipped as an emergency reactor cooling system for the experimental multi purpose very high temperature gas-cooled reactor are explained. Effects of natural circulation flow which would develop in the core and temperature transients of the panel in starting have been precisely investigated. Conditions and procedures for settling accidents with the proposed panel cooling system have been also studied. Based on these studies, it has been shown that the panel cooling system is effective and useful for the emergency reactor cooling of the experimental VHTR. (author)

Nuclear reactor power control system based on flexibility model

International Nuclear Information System (INIS)

Li Gang; Zhao Fuyu; Li Chong; Tai Yun

2011-01-01

Design the nuclear reactor power control system in this paper to cater to a nonlinear nuclear reactor. First, calculate linear power models at five power levels of the reactor as five local models and design controllers of the local models as local controllers. Every local controller consists of an optimal controller contrived by the toolbox of Optimal Controller Designer (OCD) and a proportion-integration-differentiation (PID) controller devised via Genetic Algorithm (GA) to set parameters of the PID controller. According to the local models and controllers, apply the principle of flexibility model developed in the paper to obtain the flexibility model and the flexibility controller at every power level. Second, the flexibility model and the flexibility controller at a level structure the power control system of this level. The set of the whole power control systems corresponding to global power levels is to approximately carry out the power control of the reactor. Finally, the nuclear reactor power control system is simulated. The simulation result shows that the idea of flexibility model is feasible and the nuclear reactor power control system is effective. (author)

Aladağlar da Bitki Formasyonları ve Dağılışları

OpenAIRE

ÜNALDI, Ülkü ESER; TOROĞLU, Emin

2007-01-01

Bulunduğu konum itibariyle Türkiye\\'de yetişme şartlarının bitki gelişimine ve çeşitliliğine uygun koşullar taşıdığı yerlerden biri de Aladağlar\\'dır. Nitekim dağda orman, çalı ve ot formasyonları yer almaktadır. Orman formasyonu kuru ormanlar ve yarı nemli ormanlar olmak üzere ikiye ayrılmaktadır. Kuru ormanlar güney yamaçta 250-1300mler, kuzey yamaçta 1400–2400 mler arasında uzanmaktadır. Kuru ormanların hakim elemanları güney yamaçlarda kızılçam, kuzey yamaçlarda i...

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

In equilibrium symbiotic power plant system containing both thermal reactors and fast breeders, excess plutonium produced by the fast breeders is used to enrich the fuel of the thermal reactors. In plutonium deficient symbiotic power plant system plutonium is supplied both by thermal plants and fast breeders. Mathematical models were constructed and different equations solved to characterize the fuel utilization of both systems if they contain only a single thermal type and a single fast type reactor. The more plutonium is produced in the system, the higher output ratio of thermal to fast reactors is achieved in equilibrium symbiotic power plant system. Mathematical equations were derived to calculate the doubling time and the breeding gain of the equilibrium symbiotic system. (V.N.) 2 figs.; 2 tabs

Steam generating system in LMFBR type reactors

International Nuclear Information System (INIS)

Kurosawa, Katsutoshi.

1984-01-01

Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

The digital reactor protection system for the instrumentation and control of reactor TRIGA PUSPATI (RTP)

International Nuclear Information System (INIS)

Nurfarhana Ayuni Joha; Izhar Abu Hussin; Mohd Idris Taib; Zareen Khan Abdul Jalil Khan

2010-01-01

Reactor Protection System (RPS) is important for Reactor Instrumentation and Control System. The RPS comprises all redundant electrical devices and circuitry involved in the generation of those initiating signals associated to the trip protective function. The instrumentation system for the RPS provides automatic protection signals against unsafe and improper reactor operation. The physical separation is provided for all of the redundant instrumentation systems to preserve redundancy. The safety protection systems using circuits composed of analog instruments and relays with relay contacts is difficult to realize from various reasons. Therefore, an application of digital technology can be said a logical conclusion also in the light of its functional superiority. (author)

Event tree analysis for the system of hybrid reactor

International Nuclear Information System (INIS)

Yang Yongwei; Qiu Lijian

1993-01-01

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

Fuel assembly transfer and storage system for nuclear reactors

International Nuclear Information System (INIS)

Allain, Albert; Thomas, Claude.

1981-01-01

Transfer and storage system on a site comprising several reactors and at least one building housing the installations common to all these reactors. The system includes: transfer and storage modules for the fuel assemblies comprising a containment capable of containing several assemblies carried on a transport vehicle, a set of tracks for the modules between the reactors and the common installations, handling facilities associated with each reactor for moving the irradiated assemblies from the reactor to a transfer module placed in loading position on a track serving the reactor and conversely to move the new assemblies from the transfer module to the reactor, and at least one handling facility located in the common installation building for loading the modules with new assemblies [fr

Cooling system for auxiliary reactor component

International Nuclear Information System (INIS)

Fujihira, Tomoko.

1991-01-01

A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

Nõmme maja "Kergus" : Kerese tn., Tallinn / Ülar Mark, Kalle Komissarov

Index Scriptorium Estoniae

Mark, Ülar

1999-01-01

Kahekorruselise, tänavapoolse pika umbse seinaga eramu projekteeris Arhitektuuribüroo Mark & Tamm, arhitekt Ülar Mark, kaasautor Kalle Komissarov. Konstruktiivne osa: Randväli & Karema Inseneribüroo. Projekt 1997, hoone valmis 1999. Eramu tähenduse ümbermõtestamisest. 13 illustratsiooni: korruste plaanid, välisvaated, sisevaade.Mark & Tamm (arhitektibüroo)

The IAEA power reactor information system - PRIS

International Nuclear Information System (INIS)

Laue, H.J.; Qureshi, A.; Skjoeldebrand, R.; White, D.

1983-01-01

The IAEA Power Reactor Information System, PRIS, is based on a collection of basic design data and operating experience data which the IAEA started in 1970. PRIS is used for annual publications on 'Power Reactors in Member States', 'Operating Experience with Nuclear Power Stations in Member States', which gives annual operating information for individual plants, and a 'Performance Analysis Report' summarizing each year's and earlier experience. Since 1973 information has been collected in a systematic manner on significant plant outages (= more than 10 full power hours). There is now information on more than 10,000 outages in the system which permits some conclusions to be drawn both in regard to individual plants and to categories of plants on the significance of different outage reasons and different types of equipment failures. PRIS has not been intended to be a component reliability information system as an international data collection must stop short of the level of detail which would be needed for that purpose. The objectives of PRIS have been to provide a factual background for assumptions on parameters which are essential for economic evaluations and for systems operation planning (load factor and availability). The outage information does, however, lend itself to conclusions about generic problems in different categories of plants and it can be used by an individual operator to find other plants where information about particular problems can be obtained. It would also now be possible to use PRIS for setting availability goals based on experience and not only on theoretical design considerations. The paper demonstrates the conclusions which can be drawn from 662 reactor years of operation of light and heavy water pressurized reactors and 390 reactor years of boiling water reactors and, in particular, the role that the main heat removal system and its components have played in the equipment failure category

IAEA data base system for nuclear research reactors (RRDB)

International Nuclear Information System (INIS)

Lipscher, P.

1986-01-01

The IAEA Data Base System for Nuclear Research Reactors (RRDB) User's Guide is intended for the user who wishes to understand the concepts and operation of the RRDB system. The RRDB is a computerized system recording administrative, operational and technical data on all the nuclear research reactors currently operating, under construction, planned or shut down in IAEA Member States. The data is received by the IAEA from reactor centres on magnetic tapes or as responses to questionnaires. All the data on research, training, test and radioactive isotope production reactors and critical assemblies is stored on the RRDB system. A full set of RRDB programs (in NATURAL) are contained at the back of this Guide

Ageing investigation and upgrading of components/systems of Kartini research reactor

Energy Technology Data Exchange (ETDEWEB)

Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

1998-10-01

Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

Survey of thorium utilization in power reactor systems

International Nuclear Information System (INIS)

Schwartz, M.H.; Schleifer, P.; Dahlberg, R.C.

1976-01-01

It is clear that thorium-fueled thermal power reactor systems based on current technology can play a vital role in serving present and long-term energy needs. Advanced thorium converters and thermal breeders can provide an expanded resource base from which the world's growing energy demands can be met. Utilization of a symbiotic system of fast breeders and thorium-fueled thermal reactors can be particularly effective in providing low cost power while conserving uranium resources. Breeder reactors are characterized by high capital costs and very low fuel costs since they produce more fuel than they consume. This excess fuel can be used to fuel thermal converter reactors whose capital costs are low. This symbiosis is optimized when 233 U is bred in the fast breeders and then used to fuel high-conversion-ratio thermal converter reactors operating on the thorium-uranium fuel cycle. The thorium-cycle HTGR, after undergoing more than fifteen years of development in both the United States and Europe, provides for the optimum utilization of our limited uranium resources. Other thermal reactor systems, previously operating on the uranium cycle, also show potential in their capability to utilize the thorium cycle effectively

Reactor inventory monitoring system for Angra-1 reactor

International Nuclear Information System (INIS)

S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

1996-01-01

This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

Molten-salt reactor information system

International Nuclear Information System (INIS)

Haubenreich, P.N.; Cardwell, D.W.; Engel, J.R.

1975-06-01

The Molten-Salt Reactor Information System (MSRIS) is a computer-based file of abstracts of documents dealing with the technology of molten-salt reactors. The file is stored in the IBM-360 system at ORNL, and may be searched through the use of established interactive computer programs from remote terminals connected to the computer via telephone lines. The system currently contains 373 entries and is subject to updating and expansion as additional information is developed. The nature and general content of the data file, a general approach for obtaining information from it, and the manner in which material is added to the file are described. Appendixes provide the list of keywords currently in use, the subject categories under which information is filed, and simplified procedures for searching the file from remote terminals. (U.S.)

Electrical system regulations of the IEA-R1 reactor

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2013-01-01

The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

Comparative analysis of nuclear reactor control system designs

International Nuclear Information System (INIS)

Russcher, G.E.

1975-01-01

Control systems are vital to the safe operation of nuclear reactors. Their seismic design requirements are some of the most important criteria governing reactor system design evaluation. Consequently, the seismic analysis for nuclear reactors is directed to include not only the mechanical and structural seismic capabilities of a reactor, but the control system functional requirements as well. In the study described an alternate conceptual design of a safety rod system was compared with a prototypic system design to assess their relative functional reliabilities under design seismic conditions. The comparative methods utilized standard success tree and decision tree techniques to determine the relative figures of merit. The study showed: (1) The methodology utilized can provide both qualitative and quantitative bases for design decisions regarding seismic functional capabilities of two systems under comparison, (2) the process emphasizes the visibility of particular design features that are subject to common mode failure while under seismic loading, and (3) minimal improvement was shown to be available in overall system seismic performance of an independent conceptual design, however, it also showed the system would be subject to a new set of operational uncertainties which would have to be resolved by extensive development programs

Development of a system model for advanced small modular reactors.

Energy Technology Data Exchange (ETDEWEB)

Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

2014-01-01

This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

Reactor protection system including engineered features actuation system

International Nuclear Information System (INIS)

Palmaers, W.

1982-01-01

The safety concept requires to ensure that - the reactor protection system - the active engineered safeguard - and the necessary auxiliary systems are so designed and interfaced in respect of design and mode of action that, in the event of single component failure reliable control of the consequences of accidents remains ensured at all times and that the availability of the power plant is not limited unnecessarily. In order to satisfy these requirements due, importance was attached to a consistent spacial separation of the mutually redundant subsystems of the active safety equipment. The design and layout of the reactor protection system, of the power supply (emergency power supply), and of the auxiliary systems important from the safety engineering point of view, are such that their subsystems also largely satisfy the requirements of independence and spacial separation. (orig./RW)

Stack Monitoring System At PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Zamrul Faizad Omar; Mohd Sabri Minhat; Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Izhar Abu Hussin

2014-01-01

This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

Nuclear reactor vessel decontamination systems

International Nuclear Information System (INIS)

McGuire, P. J.

1985-01-01

There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

Experimental evaluation of an expert system for nuclear reactor operators

International Nuclear Information System (INIS)

Nelson, W.R.

1984-10-01

The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

Systemization of Design and Analysis Technology for Advanced Reactor

International Nuclear Information System (INIS)

Kim, Keung Koo; Lee, J.; Zee, S. K.

2009-01-01

The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

Tandem Mirror Reactor Systems Code (Version I)

International Nuclear Information System (INIS)

Reid, R.L.; Finn, P.A.; Gohar, M.Y.

1985-09-01

A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

Additional reactor protection system of RBMK-1500

International Nuclear Information System (INIS)

1999-01-01

Analysis of anticipated transients without scram of RBMK-1500 reactor showed that additional reactor protection system is required. Data of accident analysis in the case of loose of external electric power and loose of vacuum in condensers of turbines are provided

IDAS-RR: an incident data base system for research reactors

International Nuclear Information System (INIS)

Matsumoto, Kiyoshi; Kohsaka, Atsuo; Kaminaga, Masanori; Murayama, Youji; Ohnishi, Nobuaki; Maniwa, Masaki.

1990-03-01

An Incident Data Base System for Research Reactors, IDAS-RR, has been developed. IDAS-RR has information about abnormal incidents (failures, transients, accidents, etc.) of research reactors in the world. Data reference, input, editing and other functions of IDAS-RR are menu driven. The routine processing and data base management functions are performed by the system software and hardware. PC-9801 equipment was selected as the hardware because of its portability and popularity. IDAS-RR provides effective reference information for the following activities. 1) Analysis of abnormal incident of research reactors, 2) Detail analysis of research reactor behavior in the abnormal incident for building the knowledge base of the reactor emergency diagnostic system for research reactor, 3) Planning counter-measure for emergency situation in the research reactor. This report is a user's manual of IDAS-RR. (author)

Metrology/viewing system for next generation fusion reactors

International Nuclear Information System (INIS)

Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M.; Dagher, M.A.

1997-01-01

Next generation fusion reactors require accurate measuring systems to verify sub-millimeter alignment of plasma-facing components in the reactor vessel. A metrology system capable of achieving such accuracy must be compatible with the vessel environment of high gamma radiation, high vacuum, elevated temperature, and magnetic field. This environment requires that the system must be remotely deployed. A coherent, frequency modulated laser radar system is being integrated with a remotely operated deployment system to meet these requirements. The metrology/viewing system consists of a compact laser transceiver optics module which is linked through fiber optics to the laser source and imaging units that are located outside of the harsh environment. The deployment mechanism is a telescopic-mast positioning system. This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision ranging and surface mapping system

Boiling water reactor liquid radioactive waste processing system

International Nuclear Information System (INIS)

Anon.

1977-01-01

The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-11-15

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Nuclear reactor sealing system

International Nuclear Information System (INIS)

McEdwards, J.A.

1983-01-01

A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

BWR reactor management system

International Nuclear Information System (INIS)

Makino, Kakuji; Kawamura, Atsuo; Yoshioka, Ritsuo; Neda, Toshikatsu.

1979-01-01

It is necessary to grasp the delicate state of operation in reactor cores in view of the control of burn-up and power output at the time of the operation management of BWRs. Enormous labor has been required for the collection, processing and evaluation of the data. It is desirable to obtain the safer, more efficient and faster method of operation control by predicting the states in cores including the change of xenon and reflecting them to operation plans as well as by tracing with high accuracy the past burn-up history for a long period. At present, the on-line evaluation of the states in cores is carried out with the process computers attached to respective units, but the amount of data required for core operation management of high degree far exceeds their capacity. From such viewpoints, the research and development on the reactor management system were carried out. The data processing concerning core operation management is performed with newly installed computers utilizing the data from existing process computers, and the operation of reactor cores, the qualitative improvement of management works, labor saving, and fast, efficient operation control are feasible with it. This system was installed in an actual plant in October, 1977. The composition of the system, the prediction of the change in local output distribution accompanying control rod operation, the prediction of the change in the states in cores due to the flow rate of coolant, and the function of collecting plant data are explained. (Kako, I.)

Advanced robotic remote handling system for reactor dismantlement

International Nuclear Information System (INIS)

Shinohara, Yoshikuni; Usui, Hozumi; Fujii, Yoshio

1991-01-01

An advanced robotic remote handling system equipped with a multi-functional amphibious manipulator has been developed and used to dismantle a portion of radioactive reactor internals of an experimental boiling water reactor in the program of reactor decommissioning technology development carried out by the Japan Atomic Energy Research Institute. (author)

REACTOR - a Concept for establishing a System-of-Systems

Science.gov (United States)

Haener, Rainer; Hammitzsch, Martin; Wächter, Joachim

2014-05-01

REACTOR is a working title for activities implementing reliable, emergent, adaptive, and concurrent collaboration on the basis of transactional object repositories. It aims at establishing federations of autonomous yet interoperable systems (Systems-of-Systems), which are able to expose emergent behaviour. Following the principles of event-driven service-oriented architectures (SOA 2.0), REACTOR enables adaptive re-organisation by dynamic delegation of responsibilities and novel yet coherent monitoring strategies by combining information from different domains. Thus it allows collaborative decision-processes across system, discipline, and administrative boundaries. Interoperability is based on two approaches that implement interconnection and communication between existing heterogeneous infrastructures and information systems: Coordinated (orchestration-based) communication and publish/subscribe (choreography-based) communication. Choreography-based communication ensures the autonomy of the participating systems to the highest possible degree but requires the implementation of adapters, which provide functional access to information (publishing/consuming events) via a Message Oriented Middleware (MOM). Any interconnection of the systems (composition of service and message cascades) is established on the basis of global conversations that are enacted by choreographies specifying the expected behaviour of the participating systems with respect to agreed Service Level Agreements (SLA) required by e.g. national authorities. The specification of conversations, maintained in commonly available repositories also enables the utilisation of systems for purposes (evolving) other than initially intended. Orchestration-based communication additionally requires a central component that controls the information transfer via service requests or event processing and also takes responsibility of managing business processes. Commonly available transactional object repositories are

Reactor safety systems

International Nuclear Information System (INIS)

Kafka, P.

1975-01-01

The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

Reactor coolant purification system circulation pumps (CUW pumps)

International Nuclear Information System (INIS)

Tsutsui, Toshiaki

1979-01-01

Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

Modeling and simulation of CANDU reactor and its regulating system

Science.gov (United States)

Javidnia, Hooman

Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different

Simplified safety and containment systems for the iris reactor

International Nuclear Information System (INIS)

Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

2001-01-01

The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

Applications of plasma core reactors to terrestrial energy systems

International Nuclear Information System (INIS)

Lantham, T.S.; Biancardi, F.R.; Rodgers, R.J.

1974-01-01

Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrail applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times

Reactor alarm system development and application issues

Energy Technology Data Exchange (ETDEWEB)

Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

1997-09-01

The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

Reactor alarm system development and application issues

International Nuclear Information System (INIS)

Drexler, J.E.; Oicese, G.O.

1997-01-01

The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

Dosimetry system of the RB reactor; Dozimetarski sistem reaktora RB

Energy Technology Data Exchange (ETDEWEB)

Lolic, B; Vukadin, D [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1962-07-01

Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor.

Electron Attenuation Measurement using Cosmic Ray Muons at the MicroBooNE LArTPC

Energy Technology Data Exchange (ETDEWEB)

Meddage, Varuna [Kansas State U., Manhattan

2017-10-01

The MicroBooNE experiment at Fermilab uses liquid argon time projection chamber (LArTPC) technology to study neutrino interactions in argon. A fundamental requirement for LArTPCs is to achieve and maintain a low level of electronegative contaminants in the liquid to minimize the capture of drifting ionization electrons. The attenuation time for the drifting electrons should be long compared to the maximum drift time, so that the signals from particle tracks that generate ionization electrons with long drift paths can be detected efficiently. In this talk we present MicroBooNE measurement of electron attenuation using cosmic ray muons. The result yields a minimum electron 1/e lifetime of 18 ms under typical operating conditions, which is long compared to the maximum drift time of 2.3 ms.

TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

International Nuclear Information System (INIS)

Solbrig, C.W.; Stevens, W.W.

1990-01-01

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

Core cooling system for reactor

International Nuclear Information System (INIS)

Kondo, Ryoichi; Amada, Tatsuo.

1976-01-01

Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

Naval application of battery optimized reactor integral system

International Nuclear Information System (INIS)

Kim, N. H.; Kim, T. W.; Son, H. M.; Suh, K. Y.

2007-01-01

Past civilian N.S. Savanna (80 MW t h), Otto-Hahn (38 MW t h) and Mutsu (36 MW t h) experienced stable operations under various sea conditions to prove that the reactors were stable and suitable for ship power source. Russian nuclear icebreakers such as Lenin (90 MW t h x2), Arukuchika (150 MW t h x2) showed stable operations under severe conditions during navigation on the Arctic Sea. These reactor systems, however, should be made even more efficient, compact, safe and long life, because adding support from the land may not be available on the sea. In order to meet these requirements, a compact, simple, safe and innovative integral system named Naval Application Vessel Integral System (NAVIS) is being designed with such novel concepts as a primary liquid metal coolant, a secondary supercritical carbon dioxide (SCO 2 ) coolant, emergency reactor cooling system, safety containment and so on. NAVIS is powered by Battery Optimized Reactor Integral System (BORIS). An ultra-small, ultra-long-life, versatile-purpose, fast-spectrum reactor named BORIS is being developed for a multi-purpose application such as naval power source, electric power generation in remote areas, seawater desalination, and district heating. NAVIS aims to satisfy special environment on the sea with BORIS using the lead (Pb) coolant in the primary system. NAVIS improves the economical efficiency resorting to the SCO 2 Brayton cycle for the secondary system. BORIS is operated by natural circulation of Pb without needing pumps. The reactor power is autonomously controlled by load-following operation without an active reactivity control system, whereas B 4 C based shutdown control rod is equipped for an emergency condition. SCO 2 promises a high power conversion efficiency of the recompression Brayton cycle due to its excellent compressibility reducing the compression work at the bottom of the cycle and to a higher density than helium or steam decreasing the component size. Therefore, the SCO 2 Brayton

Power Reactor Information System (PRIS)

International Nuclear Information System (INIS)

Spiegelberg, R.

1992-01-01

The IAEA has been collecting Operating Experience data for Nuclear Power Plants of the IAEA Member States since 1970. In order to facilitate an analysis of nuclear power plant performance as well as to produce relevant publications, all previously collected data supplied from the questionnaires were computerized in 1980 and the Power Reactor Information System was implemented. PRIS currently contains production records for the years up to and including 1990 and about 98% of the reactors-years operating experience in the world is contained in PRIS. (orig.)

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

A gas-cooled reactor surface power system

International Nuclear Information System (INIS)

Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

1999-01-01

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

A Gas-Cooled Reactor Surface Power System

Energy Technology Data Exchange (ETDEWEB)

Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

1998-11-09

A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

Microprocessor tester for the treat upgrade reactor trip system

International Nuclear Information System (INIS)

Lenkszus, F.R.; Bucher, R.G.

1984-01-01

The upgrading of the Transient Reactor Test (TREAT) Facility at ANL-Idaho has been designed to provide additional experimental capabilities for the study of core disruptive accident (CDA) phenomena. In addition, a programmable Automated Reactor Control System (ARCS) will permit high-power transients up to 11,000 MW having a controlled reactor period of from 15 to 0.1 sec. These modifications to the core neutronics will improve simulation of LMFBR accident conditions. Finally, a sophisticated, multiply-redundant safety system, the Reactor Trip System (RTS), will provide safe operation for both steady state and transient production operating modes. To insure that this complex safety system is functioning properly, a Dedicated Microprocessor Tester (DMT) has been implemented to perform a thorough checkout of the RTS prior to all TREAT operations

User's manual for the reactor burnup system, REBUS

International Nuclear Information System (INIS)

Olson, A.P.; Regis, J.P.; Meneley, D.A.; Hoover, L.J.

1972-01-01

A user's manual for the REBUS System (REactor BUrnup System) is presented. Its primary purpose is to provide sufficient information about the REBUS capability to the user to ensure its efficient utilization. The current REBUS System either solves for the infinite time (equilibrium) operating conditions of a recycle system under fixed conditions, or solves for operating conditions during a single time step (non-equilibrium). The capability of studying various in-reactor fuel management and ex-reactor fuel management schemes has been included. REBUS has been operated with one- and two-dimensional diffusion theory neutronics solutions up to the present time. The model was specifically designed for extension to other neutronics models such as three-dimensional diffusion or transport theory and direct or synthesis solutions

Gas-cooled reactor power systems for space

International Nuclear Information System (INIS)

Walter, C.E.

1987-01-01

Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

Nuclear reactor cavity floor passive heat removal system

Science.gov (United States)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

2018-03-06

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

Nuclear reactor safety system

International Nuclear Information System (INIS)

Sato, Takashi.

1979-01-01

Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

FISS: a computer program for reactor systems studies

International Nuclear Information System (INIS)

Tamm, H.; Sherman, G.R.; Wright, J.H.; Nieman, R.E.

1979-08-01

ΣFISSΣ is a computer code for use in investigating alternative fuel cycle strategies for Canadian and world nuclear programs. The code performs a system simulation accounting for dynamic effects of growing nuclear systems. Facilities in the model include storage for irradiated fuel, mines, plants for enrichment, fuel fabrication, fuel reprocessing and heavy water, and reactors. FISS is particularly useful for comparing various reactor strategies and studying sensitivities of resource consumption, capital investment and energy costs with changes in fuel cycle parameters, reactor parameters and financial variables. (author)

Introduction to nuclear power reactors and their health physics systems

International Nuclear Information System (INIS)

Brtis, J.S.

1982-01-01

This paper provides an introduction to: (1) the major systems of Boiling Water Reactors (BWR's) and Pressurized Water Reactors (PWR's), (2) the production and distribution of radiation sources in BWR's and PWR's, (3) the regulatory and functional requirements for nuclear power reactor design from a health physics standpoint, (4) the health physics systems provided to meet such requirements, and (5) a bibliography of documents germane to power reactor health physics design

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

Science.gov (United States)

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

2015-11-01

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

Recirculation system for nuclear reactors

International Nuclear Information System (INIS)

Braun, H. E.; Dollard, W. J.; Tower, S. N.

1980-01-01

A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

MATLAB/SIMULINK platform for simulation of CANDU reactor control system

International Nuclear Information System (INIS)

Javidnia, H.; Jiang, J.

2007-01-01

In this paper a simulation platform for CANDU reactors' control system is presented. The platform is built on MATLAB/SIMULINK interactive graphical interface. Since MATLAB/SIMULINK are powerful tools to describe systems mathematically, all the subsystems in a CANDU reactor are represented in MATLAB's language and are implemented in SIMULINK graphical representation. The focus of the paper is on the flux control loop of CANDU reactors. However, the ideas can be extended to include other parts in CANDU power plants and the same technique can be applied to other types of nuclear reactors and their control systems. The CANDU reactor model and xenon feedback model are also discussed in this paper. (author)

Multivariate statistical pattern recognition system for reactor noise analysis

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

1976-01-01

A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system

Multivariate statistical pattern recognition system for reactor noise analysis

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

1975-01-01

A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system. 19 references

Fuel transfer system for a nuclear reactor

International Nuclear Information System (INIS)

Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

1977-01-01

Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

New technology for reactor protection system of CAREM reactor

International Nuclear Information System (INIS)

Dezzutti, J.C.; Verrastro, C.; Estryk, D.

2009-01-01

The use of FPGA in safety functions in a nuclear power plant, increase the reliability of software based systems, without loose any of the function required by the supervision and control systems. In this work the architecture of a Reactor Protection System is described, it use four independent measurement channels in 2 oo 4 configuration, each channel is based on diverse approach in 1 oo 2 configuration, the reliability of this system is near the same than the hardwired logic, with full performance like software based system. (author)

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Design study of plant system for the fusion experimental reactor (FER)

International Nuclear Information System (INIS)

Iida, Hiromasa; Kuroda, Hideo; Yamada, Masao; Suzuki, Tatsushi; Honda, Tsutomu; Ohmura, Hiroshi; Itoh, Shinichi.

1986-11-01

This report describes design study results of the FER plant system. The purpose of this study is to have an image of the FER plant system as a whole by designing major auxiliary systems, reactor building and maintenance and radwaste desposal systems. The major auxiliary systems include tritium, cooling, evacuation and fueling systems. For these each systems, flowdiagrams are studied and designs of devices and pipings are conducted. In the reactor building design, layout of the above auxiliary systems in the building is studied with careful zoning concept by the radiation level. Structural integrity of the reactor building is also studied including seismic analysis. In the design of the maintenance and radwaste system flowdiagram of failed reactor components is developed and transfer vehicles and buildings are designed. Finally assuming JAERI Naka site as the reactor site layout of the whole FER plant system is developed. (author)

Reactor structure and superconducting magnet system of ITER

International Nuclear Information System (INIS)

Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

1993-01-01

Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

D. E. Shropshire

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

International Nuclear Information System (INIS)

Shropshire, D.E.

2009-01-01

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

Lichenised and Lichenicolous Fungi of Aladağlar National Park (Niğde, Kayseri and Adana Provinces) in Turkey

OpenAIRE

HALICI, Mehmet Gökhan; AKSOY, Ahmet

2014-01-01

Three hundred and two lichenised fungi taxa belonging to 90 genera and 45 lichenicolous fungi taxa belonging to 24 genera are reported from Aladağlar National Park. Of these, 290 lichenised fungi taxa and 21 lichenicolous fungi taxa are reported for the first time from Aladağlar National Park. Nine species of lichenised fungi, namely Arthopyrenia fraxini A.Massal., Aspicilia obscurata (Fr.) Arnold, Cephalophysis leucospila (Anzi) H.Kilias & Scheid., Chaenotheca ferruginea (Turner ex S...

Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

International Nuclear Information System (INIS)

Conklin, J.C.

1990-04-01

The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

New reactor technology: safety improvements in nuclear power systems.

Science.gov (United States)

Corradini, M L

2007-11-01

Almost 450 nuclear power plants are currently operating throughout the world and supplying about 17% of the world's electricity. These plants perform safely, reliably, and have no free-release of byproducts to the environment. Given the current rate of growth in electricity demand and the ever growing concerns for the environment, nuclear power can only satisfy the need for electricity and other energy-intensive products if it can demonstrate (1) enhanced safety and system reliability, (2) minimal environmental impact via sustainable system designs, and (3) competitive economics. The U.S. Department of Energy with the international community has begun research on the next generation of nuclear energy systems that can be made available to the market by 2030 or earlier, and that can offer significant advances toward these challenging goals; in particular, six candidate reactor system designs have been identified. These future nuclear power systems will require advances in materials, reactor physics, as well as thermal-hydraulics to realize their full potential. However, all of these designs must demonstrate enhanced safety above and beyond current light water reactor systems if the next generation of nuclear power plants is to grow in number far beyond the current population. This paper reviews the advanced Generation-IV reactor systems and the key safety phenomena that must be considered to guarantee that enhanced safety can be assured in future nuclear reactor systems.

Recycling systems for BWR type reactors

International Nuclear Information System (INIS)

Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

1986-01-01

Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

Power Trip Set-points of Reactor Protection System for New Research Reactor

International Nuclear Information System (INIS)

Lee, Byeonghee; Yang, Soohyung

2013-01-01

This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

Directory of Open Access Journals (Sweden)

Alexander Labutin

2017-02-01

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

Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor

International Nuclear Information System (INIS)

Raisic, N.

1962-09-01

In addition to the safety analysis for the zero power RB reactor, this report contains a general description of the reactor, reactor components, auxiliary equipment and the reactor building. Reactor Rb has been reconstructed during 1961-1962 and supplied with new safety-control system as well as with a complete dosimetry instrumentation. Since RB reactor was constructed without shielding special attention is devoted to safety and protection of the staff performing experiments. Due to changed circumstances in the Institute ( start-up of the RA 7 MW power reactor) the role of the RB reactor was redefined

System and prospect assesment of the small innovative reactor IRIS-50

International Nuclear Information System (INIS)

Lumbanraja, Sahala M.; Wibowo

2002-01-01

System and prospect of the small innovative reactor IRIS-50 in Indonesia have been studied. IRIS-50 (International Reactor Innovative and Secure) is an advanced light water cooled modular reactor being developed by an international consortium led by Westinghouse. This reactor is specifically developed to match market demands, or to developing country. This reactor is based on simplified operation and maintenance, enhanced and safety, easy to inspect, short construction time, small investment cost, competitive generating cost, and easily suited to the infrastructures. IRIS main characteristic is integral reactor concept, being all the major reactor coolant system components located inside the pressure vessel

Validation of reactor core protection system

International Nuclear Information System (INIS)

Lee, Sang-Hoon; Bae, Jong-Sik; Baeg, Seung-Yeob; Cho, Chang-Ho; Kim, Chang-Ho; Kim, Sung-Ho; Kim, Hang-Bae; In, Wang-Kee; Park, Young-Ho

2008-01-01

Reactor COre Protection System (RCOPS), an advanced core protection calculator system, is a digitized one which provides core protection function based on two reactor core operation parameters, Departure from Nucleate Boiling Ratio (DNBR) and Local Power Density (LPD). It generates a reactor trip signal when the core condition exceeds the DNBR or LPD design limit. It consists of four independent channels adapted a two-out-of-four trip logic. System configuration, hardware platform and an improved algorithm of the newly designed core protection calculator system are described in this paper. One channel of RCOPS was implemented as a single channel facility for this R and D project where we performed final integration software testing. To implement custom function blocks, pSET is used. Software test is performed by two methods. The first method is a 'Software Module Test' and the second method is a 'Software Unit Test'. New features include improvement of core thermal margin through a revised on-line DNBR algorithm, resolution of the latching problem of control element assembly signal and addition of the pre-trip alarm generation. The change of the on-line DNBR calculation algorithm is considered to improve the DNBR net margin by 2.5%-3.3%. (author)

High-Fat Diet Augments VPAC1 Receptor-Mediated PACAP Action on the Liver, Inducing LAR Expression and Insulin Resistance

Directory of Open Access Journals (Sweden)

Masanori Nakata

2016-01-01

Full Text Available Pituitary adenylate cyclase-activating polypeptide (PACAP acts on multiple processes of glucose and energy metabolism. PACAP potentiates insulin action in adipocytes and insulin release from pancreatic β-cells, thereby enhancing glucose tolerance. Contrary to these effects at organ levels, PACAP null mice exhibit hypersensitivity to insulin. However, this apparent discrepancy remains to be solved. We aimed to clarify the mechanism underlying the antidiabetic phenotype of PACAP null mice. Feeding with high-fat diet (HFD impaired insulin sensitivity and glucose tolerance in wild type mice, whereas these changes were prevented in PACAP null mice. HFD also impaired insulin-induced Akt phosphorylation in the liver in wild type mice, but not in PACAP null mice. Using GeneFishing method, HFD increased the leukocyte common antigen-related (LAR protein tyrosine phosphatase in the liver in wild type mice. Silencing of LAR restored the insulin signaling in the liver of HFD mice. Moreover, the increased LAR expression by HFD was prevented in PACAP null mice. HFD increased the expression of VPAC1 receptor (VPAC1-R, one of three PACAP receptors, in the liver of wild type mice. These data indicate that PACAP-VPAC1-R signaling induces LAR expression and insulin resistance in the liver of HFD mice. Antagonism of VPAC1-R may prevent progression of HFD-induced insulin resistance in the liver, providing a novel antidiabetic strategy.

Uneventful octreotide LAR therapy throughout three pregnancies, with favorable delivery and anthropometric measures for each newborn: a case report

Directory of Open Access Journals (Sweden)

Naccache Deeb

2011-08-01

Full Text Available Abstract Introduction The safety of octreotide use, in its short-acting preparation, in pregnancy is still unclear. This report provides the first documentation of uneventful octreotide LAR use during three pregnancies in a woman with bronchial carcinoid-associated adrenocorticotropic hormone-dependent Cushing's syndrome. Case presentation A 25-year-old Arabic woman presented to our emergency department with rapid onset of headache, flaring acne and hirsutism, facial puffiness, weight gain and paroxysmal myopathy, and paranoiac thoughts of rape and sexual intimidation. After undergoing surgical removal of a mass by left lower lung lobectomy, her residual lung disease medical therapy failed. Chronic octreotide LAR injections were initiated as indicated by a positive octreoscan. Follow-up revealed a long-lasting positive response to octreotide. Avidity of octreotide to somatostatin receptor sub-type 2 was later confirmed by a positive somatostatin receptor sub-type 2 in the resected tumor specimen. Against our instructions, the patient had three spontaneous pregnancies leading to delivery of three full-term healthy children while her octreotide LAR therapy continued. Conclusion This case adds more data supporting the potential for the safe use of octreotide and the feasibility of octreotide LAR use during pregnancy, making compliance with the patient's preference not to withdraw octreotide therapy as soon as her pregnancy is confirmed a thoughtful option.

Reliability analysis of reactor protection systems

International Nuclear Information System (INIS)

Alsan, S.

1976-07-01

A theoretical mathematical study of reliability is presented and the concepts subsequently defined applied to the study of nuclear reactor safety systems. The theory is applied to investigations of the operational reliability of the Siloe reactor from the point of view of rod drop. A statistical study conducted between 1964 and 1971 demonstrated that most rod drop incidents arose from circumstances associated with experimental equipment (new set-ups). The reliability of the most suitable safety system for some recently developed experimental equipment is discussed. Calculations indicate that if all experimental equipment were equipped with these new systems, only 1.75 rod drop accidents would be expected to occur per year on average. It is suggested that all experimental equipment should be equipped with these new safety systems and tested every 21 days. The reliability of the new safety system currently being studied for the Siloe reactor was also investigated. The following results were obtained: definite failures must be detected immediately as a result of the disturbances produced; the repair time must not exceed a few hours; the equipment must be tested every week. Under such conditions, the rate of accidental rod drops is about 0.013 on average per year. The level of nondefinite failures is less than 10 -6 per hour and the level of nonprotection 1 hour per year. (author)

A CULPA É DA MULHER: O Anticristo, de Lars von Trier

Directory of Open Access Journals (Sweden)

João Nunes Silva

2016-08-01

Full Text Available RESUMO O Anticristo, filme de Lars von Trier, lançado em 2009, mostra o desespero de um casal ao perder seu único filho. Extremamente polêmico e repleto de referências bíblicas - a começar pelo título -, é um filme que choca pelo incessante desespero de uma mãe em luto que parece carregar o peso do mundo em sua condição de mulher. A proposta deste artigo é fazer uma análise deste filme iluminando a culpa cristã historicamente atribuída à mulher e seus desdobramentos imediatos, como o feminicídio.   PALAVRAS-CHAVE: Análise Fílmica; Cristianismo; Feminicídio; O Anticristo.   ABSTRACT Antichrist, film of Lars von Trier, released on 2009, shows the the despair of a couple to lose their only son. Extremely controversial and fraught with biblical references - beginning with the title -, it’s a film that shocked the unyielding despair a bereaved mother that seems to carry the weight of the world on his wife's condition. The purpose of this paper is to analyze of this film illuminating the christian guilt historically attributed to the woman and their immediate consequences, like the feminicide.   KEYWORDS: Filmic Analysis; Christianity; Feminicide; Antichrist.     RESUMEN Anticristo, pelicula de Lars von Trier, lanzada em 2009, muestra el desespero de una pareja cuando pierde su único hijo. Extremadamente polémica y llena de referencias bíblicas - empezando con el título -, es una película que sorprende al inquebrantable desesperación de una madre em luto que parece cargar el peso del mundo por ser mujer. El objetivo de este artículo es analizar esta película iluminando la culpa cristiana históricamente asignada a las mujeres y sus consecuencias inmediatas, como el feminicidio.   PALABRAS CLAVE: Análisis fílmica; Cristianismo; Feminicidio; Anticristo.

R&D Studies of the ATLAS LAr Calorimeter Readout Electronics for super-LHC

CERN Document Server

Chen, H

2010-01-01

The ATLAS Liquid Argon (LAr) calorimeters are high precision, high sensitivity and high granularity detectors, total about 180,000 signals are digitized and processed real-time on detector, to provide energy and time deposited in each detector element at every occurrence of the L1-trigger. A luminosity upgrade (x10) of the LHC will occur ~2017, the current readout electronics will have to be upgraded to sustain the higher radiation levels. A completely innovative readout scheme is being developed. The front-end readout will send out data continuously at each bunch crossing through high speed radiation resistant optical links, the data will be processed real-time with the possibility of implementing trigger algorithms. This article is an overview of the R&D activities and architectural studies the ATLAS LAr collaboration is developing: front-end analog and mixed-signal ASIC design, radiation resistance optical-links in SOS, high-speed back-end processing units based on FPGA architectures and power supply d...

A Design of Alarm System in a Research Reactor Facility

International Nuclear Information System (INIS)

Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk

2013-01-01

The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants

Reactor feedwater system

International Nuclear Information System (INIS)

Hikabe, Katsumi.

1978-01-01

Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

Design Requirements of an Advanced HANARO Reactor Core Cooling System

International Nuclear Information System (INIS)

Park, Yong Chul; Ryu, Jeong Soo

2007-12-01

An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

Fluid Flow Patterns During Production from Gas Hydrates in the Laboratory compared to Field Settings: LARS vs. Mallik

Science.gov (United States)

Strauch, B.; Heeschen, K. U.; Priegnitz, M.; Abendroth, S.; Spangenberg, E.; Thaler, J.; Schicks, J. M.

2015-12-01

The GFZ's LArge Reservoir Simulator LARS allows for the simulation of the 2008 Mallik gas hydrate production test and the comparison of fluid flow patterns and their driving forces. Do we see the gas flow pattern described for Mallik [Uddin, M. et al., J. Can. Petrol Tech, 50, 70-89, 2011] in a pilot scale test? If so, what are the driving forces? LARS has a network of temperature sensors and an electric resistivity tomography (ERT) enabling a good spatial resolution of gas hydrate occurrences, water and gas distribution, and changes in temperature in the sample. A gas flow meter and a water trap record fluid flow patterns and a backpressure valve has controlled the depressurization equivalent to the three pressure stages (7.0 - 5.0 - 4.2 MPa) applied in the Mallik field test. The environmental temperature (284 K) and confining pressure (13 MPa) have been constant. The depressurization induced immediate endothermic gas hydrate dissociation until re-establishment of the stability conditions by a consequent temperature decrease. Slight gas hydrate dissociation continued at the top and upper lateral border due to the constant heat input from the environment. Here transport pathways were short and permeability higher due to lower gas hydrate saturation. At pressures of 7.0 and 5.0 MPa the LARS tests showed high water flow rates and short irregular spikes of gas production. The gas flow patterns at 4.2 MPa and 3.0MPa resembled those of the Mallik test. In LARS the initial gas surges overlap with times of hydrate instability while water content and lengths of pathways had increased. Water production was at a minimum. A rapidly formed continuous gas phase caused the initial gas surges and only after gas hydrate dissociation decreased to a minimum the single gas bubbles get trapped before slowly coalescing again. In LARS, where pathways were short and no additional water was added, a transport of microbubbles is unlikely to cause a gas surge as suggested for Mallik.

Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

Energy Technology Data Exchange (ETDEWEB)

Qualls, A L; Cetiner, M S; Wilson, Jr, T L

2012-04-30

The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

Light water reactor safeguards system evaluation

International Nuclear Information System (INIS)

Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

1978-01-01

A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

Ventilation system in the RA reactor building - design specifications

International Nuclear Information System (INIS)

Badrljica, R.

1984-09-01

Protective role of the ventilation system of nuclear facilities involve construction of ventilation barriers which prevent release of radioactive particulates or gases, elimination od radioactive particulates and gases from the air which is released from contaminated zones into the reactor environment. Ventilation barriers are created by dividing the building into a number of ventilation zones with different sub pressure compared to the atmospheric pressure. The RA reactor building is divided into four ventilation zones. First zone is the zone of highest risk. It includes reactor core with horizontal experimental channels, underground rooms of the primary coolant system (D 2 O), helium system, hot cells and the space above the the reactor core. Second zone is the reactor hall and the room for irradiated fuel storage. The third zone includes corridors in the basement, ground floor and first floor where the probability of contamination is small. The fourth zone includes the annex where the contamination risk is low. There is no have natural air circulation in the reactor building. Ventilators for air input and outlet maintain the sub pressure in the building (pressure lower than the atmospheric pressure). This prevents release of radioactivity into the atmosphere [sr

A space vehicle rotating with a uniform angu- lar velocity about a ...

Indian Academy of Sciences (India)

IAS Admin

A space vehicle rotating with a uniform angu- lar velocity about a vertical axis fixed to it is falling freely vertically downwards, say, with its engine shut off. It carries two astronauts inside it. One astronaut throws a tiny tool towards the other astronaut. The motion of the tiny tool with reference to a rotating frame rigidly fixed.

Fast breeder reactor reference system classification for the ENEA data bank

International Nuclear Information System (INIS)

Righini, R.

1988-01-01

This report contains the Reference System Classification (RSC) of fast breeder reactors: it provides a functional system breakdown of the reactor. For each system the following important characteristics are reported: the main function, the mode of operation, its location in the reactor, the main interface system, its main components and the component working environment (fluid and/or atmosphere type). The RSC represent a basic step in organizing the ENEA data bank for the registration and processing of reliability data on typical fast reactor components; it provides a functional component breakdown and represent a plant-unique identification in the process of omogenization of event-data coming from different reactors. In this report it was tried to take into account different generations of nuclear power plants, different plant layouts and solutions: in particular loop and pool reactors are separately treated

Performance of the prism reactor's passive decay heat removal system

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

SP-100 Program: space reactor system and subsystem investigations

International Nuclear Information System (INIS)

Harty, R.B.

1983-01-01

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

Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

Energy Technology Data Exchange (ETDEWEB)

Josephson, Gary B.; Tonkyn, Russell G.; Frye, J. G.; Riley, Brian J.; Rappe, Kenneth G.

2011-04-06

Pacific Northwest National Laboratory (PNNL) has performed an assessment of a Hybrid Plasma/Filter system as an alternative to conventional methods for collective protection. The key premise of the hybrid system is to couple a nonthermal plasma (NTP) reactor with reactive adsorption to provide a broader envelope of protection than can be provided through a single-solution approach. The first step uses highly reactive species (e.g. oxygen radicals, hydroxyl radicals, etc.) created in a nonthermal plasma (NTP) reactor to destroy the majority (~75% - 90%) of an incoming threat. Following the NTP reactor an O3 reactor/filter uses the O3 created in the NTP reactor to further destroy the remaining organic materials. This report summarizes the laboratory development of the Hybrid Plasma Reactor/Filter to protect against a ‘worst-case’ simulant, methyl bromide (CH3Br), and presents a preliminary engineering assessment of the technology to Joint Expeditionary Collective Protection performance specifications for chemical vapor air purification technologies.

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

International Nuclear Information System (INIS)

Syahrudin Yusuf; M Subhan; Ikhsan Shobari; Sutomo Budihardjo

2010-01-01

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

Small reactor power systems for manned planetary surface bases

Energy Technology Data Exchange (ETDEWEB)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options.

Small reactor power systems for manned planetary surface bases

International Nuclear Information System (INIS)

Bloomfield, H.S.

1987-12-01

A preliminary feasibility study of the potential application of small nuclear reactor space power systems to manned planetary surface base missions was conducted. The purpose of the study was to identify and assess the technology, performance, and safety issues associated with integration of reactor power systems with an evolutionary manned planetary surface exploration scenario. The requirements and characteristics of a variety of human-rated modular reactor power system configurations selected for a range of power levels from 25 kWe to hundreds of kilowatts is described. Trade-off analyses for reactor power systems utilizing both man-made and indigenous shielding materials are provided to examine performance, installation and operational safety feasibility issues. The results of this study have confirmed the preliminary feasibility of a wide variety of small reactor power plant configurations for growth oriented manned planetary surface exploration missions. The capability for power level growth with increasing manned presence, while maintaining safe radiation levels, was favorably assessed for nominal 25 to 100 kWe modular configurations. No feasibility limitations or technical barriers were identified and the use of both distance and indigenous planetary soil material for human rated radiation shielding were shown to be viable and attractive options

Coolant cleanup system for a nuclear reactor

International Nuclear Information System (INIS)

Shiina, Atsushi; Usui, Naoshi; Yamamoto, Michiyoshi; Osumi, Katsumi.

1983-01-01

Purpose: To maintain the electric conductivity of reactor water lower and to minimize the heat loss in the cleanup system by providing a low temperature cleanup system and a high temperature cleanup system together. Constitution: A low temperature cleanup system using ion exchange resins as filter aids and a high temperature cleanup system using inorganic ion exchange materials as filter aids are provided in combination. A part of the reactor water in a reactor pressure vessel is passed through a conductivity meter, one portion of which flows into the high temperature cleanup system having no heat exchanger and filled with inorganic ion exchange materials by way of a first flow rate control valve and the other portion of which flows into the low temperature cleanup system having heat exchangers and filled with the ion exchange materials by way of a second control valve. The first control valve is adjusted so as to flow, for example, about more than 15% of the feedwater flow rate to the high temperature cleanup system and the second control valve is adjusted with its valve opening degree depending on the indication of the conductivity meter so as to flow about 2 - 7 % of the feedwater flow rate into the low temperature cleanup system, to thereby control the electric conductivity to between 0.055 - 0.3 μS/cm. (Moriyama, K.)

An intelligent safety system concept for future CANDU reactors

International Nuclear Information System (INIS)

Hinds, H.W.

1980-01-01

A review of the current Regional Over-power Trip (ROPT) system employed on the Bruce NGS-A reactors confirmed the belief that future reactors should have an improved ROPT system. We are developing such an 'intelligent' safety system. It uses more of the available information on reactor status and employs modern computer technology. Fast triplicated safety computers compute maps of fuel channel power, based on readings from prompt-responding flux detectors. The coefficients for this calculation are downloaded periodically from a fourth supervisor computer. These coefficients are based on a detailed 3-D flux shape derived from physics data and other plant information. A demonstration of one of three safety channels of such a system is planned. (auth)

Method of controlling ECCS system in reactors

International Nuclear Information System (INIS)

Oohashi, Hideaki; Ikehara, Morihiko.

1982-01-01

Purpose: To eliminate the risk of misoperation and thereby improve the reliability of ECCS system upon accident. Method: ECCS system for nuclear reactor is automatically started by either of signals from a water level detector in a pressure vessel or from a pressure detector in a reactor container. Further, the ECCS system is started or stopped by the manual operation irrespective of the signals, and the signals from the pressure detector are isolated from the ECCS-starting signal by the contacts which actuate interlocked with the stopping operation of the manual operation switch. Then, after stopping the ECCS system by the manual operation, the ECCS system is started by the signals from the water level detector irrespective of the signals from the pressure detector. (Seki, T.)

Control device for start-up of reactor depressurization system

International Nuclear Information System (INIS)

Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

1998-01-01

The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

The flow measurement methods for the primary system of integral reactors

International Nuclear Information System (INIS)

Lee, J.; Seo, J. K.; Lee, D. J.

2001-01-01

It is the common features of the integral reactors that the main components of the primary system are installed within the reactor vessel, and so there are no any flow pipes connecting the reactor coolant pumps or steam generators. Due to no any flow pipes, it is impossible to measure the differential pressure at the primary system of the integral reactors, and it also makes impossible measure the primary coolant flow rate. The objective of the study is to draw up the flow measurement methods for the primary system of integral reactors. As a result of the review, we have made a selection of the flow measurement method by pump speed, bt HBM, and by pump motor power as the flow measurement methods for the primary system of integral reactors. Peculiarly, we did not found out a precedent which the direct pump motor power-flow rate curve is used as the flow measurement method in the existing commercial nuclear power reactors. Therefore, to use this method for integral reactors, it is needed to bear the follow-up measures in mind. The follow-up measures is included in this report

Automated ultrasonic examination of light water reactor systems

International Nuclear Information System (INIS)

Walter, J.H.

1975-01-01

An automated ultrasonic examination system has been developed to meet the pre- and inservice inspection requirements of light water reactors. This system features remotely-controlled travelling instrument carriers, computerized collection and storage or inspection data in a manner providing real time comparison against code standards, and computer control over the positioning of the instrument carriers to provide precise location data. The system is currently being utilized in the field for a variety of reactor inspections. The principal features of the system and the recent inspection experience are discussed. (author)

In-service inspections of the reactor cooling system of pressurized water reactors

International Nuclear Information System (INIS)

Fuerste, W.; Hohnerlein, G.; Werden, B.

1982-01-01

In order to guarantee constant safety of the components of the reactor cooling system, regular in-service inspections are carried out after commissioning of the nuclear power plant. This contribution is concerned with the components of the reactor cooling system, referring to the legal requirements, safety-related purposes and scope of the in-service inspections during the entire period of operation of a nuclear power plant. Reports are made with respect to type, examination intervals, examination technique, results and future development. The functional tests which are carried out within the scope of the in-service inspections are not part of this contribution. (orig.) [de

Scanning tunneling microscope assembly, reactor, and system

Science.gov (United States)

Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

2014-11-18

An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

Breeder reactor fuel fabrication system development

International Nuclear Information System (INIS)

Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

1981-01-01

Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

Geometrik Yapıların İnşasında Pergel ve Çizgecin Kullanımı

OpenAIRE

Erduran, Ayten; Yeşildere, Sibel

2010-01-01

Bu yapıda üç matematik öğretinin pergel ve çizgeçmenin geometrik yapıları süreçleri incelenmektedir. Öğretmenlerin geometri yapı yapımı ve ilgili dersleri video kamera ile kaydedilmiş ve derslerdeki öğretmen-öğrenci-araç üçlüsü arasındaki etkileşim incelenmiştir. Ders Kitaplığı içeren öğretmenlerle görüşmeler yapıldı. Çalışmada üç matematik öğretmeninin pergel ve çizgeci geometrik yapılar yapı mühendisliği. Araştırma pergel ve çizgeçle geometrik yapıların inşasına ezbere bir anlayışla öğretme...

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Distributed computer control system for reactor optimization

International Nuclear Information System (INIS)

Williams, A.H.

1983-01-01

At the Oldbury power station a prototype distributed computer control system has been installed. This system is designed to support research and development into improved reactor temperature control methods. This work will lead to the development and demonstration of new optimal control systems for improvement of plant efficiency and increase of generated output. The system can collect plant data from special test instrumentation connected to dedicated scanners and from the station's existing data processing system. The system can also, via distributed microprocessor-based interface units, make adjustments to the desired reactor channel gas exit temperatures. The existing control equipment will then adjust the height of control rods to maintain operation at these temperatures. The design of the distributed system is based on extensive experience with distributed systems for direct digital control, operator display and plant monitoring. The paper describes various aspects of this system, with particular emphasis on: (1) the hierarchal system structure; (2) the modular construction of the system to facilitate installation, commissioning and testing, and to reduce maintenance to module replacement; (3) the integration of the system into the station's existing data processing system; (4) distributed microprocessor-based interfaces to the reactor controls, with extensive security facilities implemented by hardware and software; (5) data transfer using point-to-point and bussed data links; (6) man-machine communication based on VDUs with computer input push-buttons and touch-sensitive screens; and (7) the use of a software system supporting a high-level engineer-orientated programming language, at all levels in the system, together with comprehensive data link management

BEACON TSM application system to the operation of PWR reactors

International Nuclear Information System (INIS)

Lozano, J. A.; Mildrum, C.; Serrano, J. F.

2012-01-01

BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

PRISM reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.

1991-01-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

Implementation of digital control and protection systems of China advanced research reactor

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

Reactor system safety assurance

International Nuclear Information System (INIS)

Mattson, R.J.

1984-01-01

The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

Approach to developing reliable space reactor power systems

International Nuclear Information System (INIS)

Mondt, J.F.; Shinbrot, C.H.

1991-01-01

The Space Reactor Power System Project is in the engineering development phase of a three-phase program. During Phase II, the Engineering Development Phase, the SP-100 Project has defined and is pursuing a new approach to developing reliable power systems. The approach to developing such a system during the early technology phase is described in this paper along with some preliminary examples to help explain the approach. Developing reliable components to meet space reactor power system requirements is based on a top down systems approach which includes a point design based on a detailed technical specification of a 100 kW power system

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1980-01-01

This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

Reactor coolant pressure boundary leakage detection system

International Nuclear Information System (INIS)

Dissing, E.; Svansson, L.

1979-08-01

The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

Synthesis of relay control systems for nuclear reactors

International Nuclear Information System (INIS)

Postnikov, N.S.

1996-01-01

The problem on stabilizing an oscillatory-unstable reactor by a single-link relay system, the characteristics whereof have a dead zone and hysteresis loop, is considered. The methodology of synthesis of feedback law, providing for stochastic steady-state mode of reactor operation with the minimum frequency of control impact introduction is proposed. This methodology is applicable to general-type relay systems with arbitrary oscillatory-unstable objects. 6 refs., 5 figs

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

Performance Test for Neutron Detector and Associated System using Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

NEPTUNE: a modular system for light-water reactor calculation

International Nuclear Information System (INIS)

Bouchard, J.; Kanevoky, A.; Reuss, P.

1975-01-01

A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Kim, Seong Gu; Cho, Seongkuk; Yu, Hwanyeal; Kim, Yonghee; Jeong, Yong Hoon; Lee, Jeong Ik

2014-01-01

Small modular reactor (SMR) systems that have advantages of little initial capital cost and small restriction on construction site are being developed by many research organizations around the world. Existing SMR concepts have the same objective: to achieve compact size and a long life core. Most of small modular reactors have much smaller size than the large nuclear power plant. However, existing SMR concepts are not fully modularized. This paper suggests a complete modular reactor with an innovative concept for reactor cooling by using a supercritical carbon dioxide. The authors propose the supercritical CO_2 Brayton cycle (S-CO_2 cycle) as a power conversion system to achieve small volume of power conversion unit (PCU) and to contain the reactor core and PCU in one vessel. A conceptual design of the proposed small modular reactor was developed, which is named as KAIST Micro Modular Reactor (MMR). The supercritical CO_2 Brayton cycle for the S-CO_2 cooled reactor core was optimized and the size of turbomachinery and heat exchanger were estimated preliminary. The nuclear fuel composed with UN was proposed and the core lifetime was obtained from a burnup versus reactivity calculation. Furthermore, a system layout with fully passive safety systems for both normal operation and emergency operation was proposed. (author)

Plenum separator system for pool-type nuclear reactors

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1983-01-01

This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

Considerations on nuclear reactor passive safety systems

International Nuclear Information System (INIS)

2016-01-01

After having indicated some passive safety systems present in electronuclear reactors (control bars, safety injection system accumulators, reactor cooling after stoppage, hydrogen recombination systems), this report recalls the main characteristics of passive safety systems, and discusses the main issues associated with the assessment of new passive systems (notably to face a sustained loss of electric supply systems or of cold water source) and research axis to be developed in this respect. More precisely, the report comments the classification of safety passive systems as it is proposed by the IAEA, outlines and comments specific aspects of these systems regarding their operation and performance. The next part discusses the safety approach, the control of performance of safety passive systems, issues related to their reliability, and the expected contribution of R and D (for example: understanding of physical phenomena which have an influence of these systems, capacities of simulation of these phenomena, needs of experimentations to validate simulation codes)

Reactor cooling system

International Nuclear Information System (INIS)

Kato, Etsuji.

1979-01-01

Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

System modeling and reactor design studies of the Advanced Thermionic Initiative space nuclear reactor

International Nuclear Information System (INIS)

Lee, H.H.; Abdul-Hamid, S.; Klein, A.C.

1996-01-01

In-core thermionic space reactor design concepts that operate at a nominal power output range of 20 to 50 kW(electric) are described. Details of the neutronic, thermionic, thermal hydraulics, and shielding performance are presented. Because of the strong absorption of thermal neutrons by natural tungsten and the large amount of natural tungsten within the reactor core, two designs are considered. An overall system design code has been developed at Oregon State University to model advanced in-core thermionic energy conversion-based nuclear reactor systems for space applications. The results show that the driverless single-cell Advanced Thermionic Initiative (ATI) configuration, which does not have driver fuel rods, proved to be more efficient than the driven core, which has driver rods. The results also show that the inclusion of the true axial and radial power distribution decrease the overall conversion efficiency. The flattening of the radial power distribution by three different methods would lead to a higher efficiency. The results show that only one TFE works at the optimum emitter temperature; all other TFEs are off the optimum performance and result in a 40% decrease of the efficiency of the overall system. The true axial profile is significantly different as there is a considerable amount of neutron leakage out of the top and bottom of the reactor. The analysis reveals that the axial power profile actually has a chopped cosine shape. For this axial profile, the reactor core overall efficiency for the driverless ATI reactor version is found to be 5.84% with a total electrical power of 21.92 kW(electric). By considering the true axial power profile instead of the uniform power profile, each TFE loses ∼80 W(electric)

Modelling aerosol behavior in reactor cooling systems

International Nuclear Information System (INIS)

McDonald, B.H.

1990-01-01

This paper presents an overview of some of the areas of concern in using computer codes to model fission-product aerosol behavior in the reactor cooling system (RCS) of a water-cooled nuclear reactor during a loss-of-coolant accident. The basic physical processes that require modelling include: fission product release and aerosol formation in the reactor core, aerosol transport and deposition in the reactor core and throughout the rest of the RCS, and the interaction between aerosol transport processes and the thermalhydraulics. In addition to these basic physical processes, chemical reactions can have a large influence on the nature of the aerosol and its behavior in the RCS. The focus is on the physics and the implications of numerical methods used in the computer codes to model aerosol behavior in the RCS

Assessment of core protection and monitoring systems for an advanced reactor SMART

International Nuclear Information System (INIS)

In, Wang Kee; Hwang, Dae Hyun; Yoo, Yeon Jong; Zee, Sung Qunn

2002-01-01

Analogue and digital core protection/monitoring systems were assessed for the implementation in an advanced reactor. The core thermal margins to nuclear fuel design limits (departure from nucleate boiling and fuel centerline melting) were estimated using the design data for a commercial pressurized water reactor and an advanced reactor. The digital protection system resulted in a greater power margin to the fuel centerline melting by at least 30% of rated power for both commercial and advanced reactors. The DNB margin with the digital system is also higher than that for the analogue system by 8 and 12.1% of rated power for commercial and advanced reactors, respectively. The margin gain with the digital system is largely due to the on-line calculations of DNB ratio and peak local power density from the live sensor signals. The digital core protection and monitoring systems are, therefore, believed to be more appropriate for the advanced reactor

Nonlinear dynamic analysis of nuclear reactor primary coolant systems

International Nuclear Information System (INIS)

Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

1979-01-01

The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

Development of intellectual reactor design system IRDS

International Nuclear Information System (INIS)

Kugo, T.; Tsuchihashi, K.; Nakagawa, M.; Mori, T.

1993-01-01

An intellectual reactor design system IRDS has been developed to support feasibility study and conceptual design of new type reactors in the fields of reactor core design including neutronics, thermal-hydraulics and fuel design. IRDS is an integrated software system in which a variety of computer codes in the different fields are installed. An integration of simulation modules are performed by the information transfer between modules through design model in which the design information of the current design work is stored. An object oriented architecture is realized in frame representation of core configuration in a design data base. The knowledge relating to design tasks to be performed are encapsulated, to support the conceptual design work. The system is constructed on an engineering workstation, and supports efficiently design work through man-machine interface adopting the advanced information processing technologies. Optimization methods for design parameters with use of the artificial intelligence technique are now under study, to reduce the parametric study work. A function to search design window in which design is feasible is realized in the fuel pin design. (orig.)

System for unattended surveillance of nuclear reactor behavior

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.

1977-01-01

A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and dimensionality reduction capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns and to recognize deviations from these patterns were evaluated by experiments at the ORNL High-Flux Isotope Reactor. Power perturbations of less than 0.1% of the mean value in selected frequency ranges were readily detected by the pattern recognition system

Closed Brayton cycle power conversion systems for nuclear reactors :

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lipinski, Ronald J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vernon, Milton E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Travis [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2006-04-01

This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at

Reactor noise analysis applications in NPP I and C systems

Energy Technology Data Exchange (ETDEWEB)

Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

2006-07-01

Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

Prism reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.

1991-08-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

PRISM reactor system design and analysis of postulated unscrammed events

International Nuclear Information System (INIS)

Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

1991-01-01

Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

A protection system of low temperature thermo-supply nuclear reactor

International Nuclear Information System (INIS)

Jiang Binsen

1988-09-01

A Protection system of low temperature thermo-supply nuclear reactor is introduced. It is the first protection system, which is designed and manufactred on the basis of Chinese National Standard GB 4083-83 'General Safety Principle of Nuclear Reactor Protection System', to be considered under the circumstances of industry level in China. Advantages of the protection system are as follows: 1)The single failure criteria can fully be fulfilled by the protection system. 2) On-line testing system can be used for detecting all of failure components and quick identifying the failure points in the system. 3) It is convenience for maintenacnce of the system. To complete this project is very important and helpful in promoting the development of the protection system and safety operation of nuclear reactor in China

System for chemical decontamination of nuclear reactor primary systems

International Nuclear Information System (INIS)

Schlonski, J.S.; McGiure, M.F.; Corpora, G.J.

1992-01-01

This patent describes a method of chemically decontaminating a nuclear reactor primary system, having a residual heat removal system with one or more residual heat removal heat exchangers, each having an upstream and a downstream side, at or above ambient pressure. It comprises: injecting decontamination chemicals using an injection means; circulating the injected decontamination chemicals throughout the primary system; directing the circulated decontamination chemicals and process fluids to a means for removing suspended solids and dissolved materials after the circulated chemicals and process fluids have passed through the residual heat removal heat exchanger; decontaminating the process fluids; and feeding the decontaminated process fluids to the injection means. This patent also describes a chemical decontamination system for use at, or above, ambient pressure in a nuclear reactor primary system having a residual heat removal system. It comprises: means for injecting decontamination chemicals into the primary system; means for removing dissolved and suspended materials and decontamination chemicals from the primary system; one or more residual heat removal pumps; means located downstream of one of the residual heat removal heat exchangers; and a return line connecting the means

Internet and Its Contributions to Library Services Internet: Kütüphane Hizmetlerine Katkıları

Directory of Open Access Journals (Sweden)

Yaşar Çelik

1995-06-01

Full Text Available This paper examines the major Internet services such as email, telnet and ftp and describes various information discovery and retrieval tools such as gopher, WAIS, and WWW. It also discusses the contributions of the Internet to library and information services. Bu makalede elektronik posta, telnet ve ftp gibi belli başlı Internet hizmetleri tanıtılmış, çeşitli bilgi keşfetme ve erişim araçları (gopher, WAIS, WWW açıklanmıştır. İnternet'in kütüphane hizmetlerine katkıları kısaca tartışılmıştır.

Multi-objective optimization of the reactor coolant system

International Nuclear Information System (INIS)

Chen Lei; Yan Changqi; Wang Jianjun

2014-01-01

Background: Weight and size are important criteria in evaluating the performance of a nuclear power plant. It is of great theoretical value and engineering significance to reduce the weight and volume of the components for a nuclear power plant by the optimization methodology. Purpose: In order to provide a new method for the optimization of nuclear power plant multi-objective, the concept of the non-dominated solution was introduced. Methods: Based on the parameters of Qinshan I nuclear power plant, the mathematical models of the reactor core, the reactor vessel, the main pipe, the pressurizer and the steam generator were built and verified. The sensitivity analyses were carried out to study the influences of the design variables on the objectives. A modified non-dominated sorting genetic algorithm was proposed and employed to optimize the weight and the volume of the reactor coolant system. Results: The results show that the component mathematical models are reliable, the modified non-dominated sorting generic algorithm is effective, and the reactor inlet temperature is the most important variable which influences the distribution of the non-dominated solutions. Conclusion: The optimization results could provide a reference to the design of such reactor coolant system. (authors)

Research Reactor Power Control System Design by MATLAB/SIMULINK

International Nuclear Information System (INIS)

Baang, Dane; Suh, Yong Suk; Kim, Young Ki; Im, Ki Hong

2013-01-01

In this study it is presented that MATLAB/SIMULINK can be efficiently used for modeling and power control system design for research reactors. The presented power control system deals with various functions including reactivity control, signals processing, reactivity calculation, alarm request generation, etc., thus it is required to test all the software logic using proper model for reactor, control rods, and field instruments. In MATLAB/SIMULINK tool, point kinetics, thermal model, control absorber rod model, and other instrument models were developed based on reactor parameters and known properties of each component or system. The software for power control system was invented and linked to the model to test each function. From the simulation result it is shown that the power control performance and other functions of the system can be easily tested and analyzed in the proposed simulation structure

LArGe: active background suppression using argon scintillation for the GERDA 0νββ-experiment

International Nuclear Information System (INIS)

Agostini, M.; Budjas, D.; Schoenert, S.; Barnabe-Heider, M.; Cattadori, C.; Gangapshev, A.; Gusev, K.; Heisel, M.; Smolnikov, A.; Junker, M.; Klimenko, A.; Lubashevskiy, A.; Pelczar, K.; Zuzel, G.

2015-01-01

LArGe is a GERDA low-background test facility to study novel background suppression methods in a low-background environment, for future application in the GERDA experiment. Similar to GERDA, LArGe operates bare germanium detectors submersed into liquid argon (1 m 3 , 1.4tons), which in addition is instrumented with photomultipliers to detect argon scintillation light. The scintillation signals are used in anti-coincidence with the germanium detectors to effectively suppress background events that deposit energy in the liquid argon. The background suppression efficiency was studied in combination with a pulse shape discrimination (PSD) technique using a BEGe detector for various sources, which represent characteristic backgrounds to GERDA. Suppression factors of a few times 10 3 have been achieved. First background data of LArGe with a coaxial HPGe detector (without PSD) yield a background index of (0.12 - 4.6) x 10 -2 cts/(keV kg year) (90 % C.L.), which is at the level of GERDA Phase I. Furthermore, for the first time we monitor the natural 42 Ar abundance (parallel to GERDA), and have indication for the 2νββ-decay in natural germanium. These results show the effectivity of an active liquid argon veto in an ultra-low background environment. As a consequence, the implementation of a liquid argon veto in GERDA Phase II is pursued. (orig.)

LArGe: active background suppression using argon scintillation for the Gerda 0ν β β -experiment

Science.gov (United States)

Agostini, M.; Barnabé-Heider, M.; Budjáš, D.; Cattadori, C.; Gangapshev, A.; Gusev, K.; Heisel, M.; Junker, M.; Klimenko, A.; Lubashevskiy, A.; Pelczar, K.; Schönert, S.; Smolnikov, A.; Zuzel, G.

2015-10-01

LArGe is a Gerda low-background test facility to study novel background suppression methods in a low-background environment, for future application in the Gerda experiment. Similar to Gerda, LArGe operates bare germanium detectors submersed into liquid argon (1 m^3, 1.4 tons), which in addition is instrumented with photomultipliers to detect argon scintillation light. The scintillation signals are used in anti-coincidence with the germanium detectors to effectively suppress background events that deposit energy in the liquid argon. The background suppression efficiency was studied in combination with a pulse shape discrimination (PSD) technique using a BEGe detector for various sources, which represent characteristic backgrounds to Gerda. Suppression factors of a few times 10^3 have been achieved. First background data of LArGe with a coaxial HPGe detector (without PSD) yield a background index of (0.12-4.6)× 10^{-2} cts/(keV kg year) (90 % C.L.), which is at the level of Gerda Phase I. Furthermore, for the first time we monitor the natural ^{42}Ar abundance (parallel to Gerda), and have indication for the 2ν β β -decay in natural germanium. These results show the effectivity of an active liquid argon veto in an ultra-low background environment. As a consequence, the implementation of a liquid argon veto in Gerda Phase II is pursued.

LArGe: active background suppression using argon scintillation for the GERDA 0νββ-experiment

Energy Technology Data Exchange (ETDEWEB)

Agostini, M.; Budjas, D.; Schoenert, S. [Technische Universitaet Muenchen, Munich (Germany); Barnabe-Heider, M. [Technische Universitaet Muenchen, Munich (Germany); Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Cattadori, C. [Universita degli Studi di Milano, Milan (Italy); INFN, Milan (Italy); Gangapshev, A. [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Institut for Nuclear Research, Moscow (Russian Federation); Gusev, K. [Technische Universitaet Muenchen, Munich (Germany); Joint Institut for Nuclear Research, Dubna (Russian Federation); National Research Center Kurchatov Institut, Moscow (Russian Federation); Heisel, M.; Smolnikov, A. [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Junker, M. [Laboratori Nazionali del Gran Sasso, Assergi (Italy); Klimenko, A.; Lubashevskiy, A. [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Joint Institut for Nuclear Research, Dubna (Russian Federation); Pelczar, K. [Jagellonian University, Cracow (Poland); Zuzel, G. [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany); Jagellonian University, Cracow (Poland)

2015-10-15

LArGe is a GERDA low-background test facility to study novel background suppression methods in a low-background environment, for future application in the GERDA experiment. Similar to GERDA, LArGe operates bare germanium detectors submersed into liquid argon (1 m{sup 3}, 1.4tons), which in addition is instrumented with photomultipliers to detect argon scintillation light. The scintillation signals are used in anti-coincidence with the germanium detectors to effectively suppress background events that deposit energy in the liquid argon. The background suppression efficiency was studied in combination with a pulse shape discrimination (PSD) technique using a BEGe detector for various sources, which represent characteristic backgrounds to GERDA. Suppression factors of a few times 10{sup 3} have been achieved. First background data of LArGe with a coaxial HPGe detector (without PSD) yield a background index of (0.12 - 4.6) x 10{sup -2} cts/(keV kg year) (90 % C.L.), which is at the level of GERDA Phase I. Furthermore, for the first time we monitor the natural {sup 42}Ar abundance (parallel to GERDA), and have indication for the 2νββ-decay in natural germanium. These results show the effectivity of an active liquid argon veto in an ultra-low background environment. As a consequence, the implementation of a liquid argon veto in GERDA Phase II is pursued. (orig.)

Vibration test on KMRR reactor structure and primary cooling system piping

Energy Technology Data Exchange (ETDEWEB)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

Vibration test on KMRR reactor structure and primary cooling system piping

International Nuclear Information System (INIS)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

Development of the next generation reactor analysis code system, MARBLE

International Nuclear Information System (INIS)

Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

2011-03-01

A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

Osmanlı Dönemi Başlıklı Ortakent Mezar Taşları Ottoman Period Headed Gravestones of Ortakent

Directory of Open Access Journals (Sweden)

H.Kamil BİÇİCİ

2012-12-01

Full Text Available This concerns, “Ortakent Gravestones.”. Among the cemetery of Ortakent, The dates of the gravestones are between 18.th-20.th centuries and 20 samples of 30 are from 18.th. century, 9 are from 19.th., 1 is from 20.th. century. Gravestones were of marble. The lenghts of the stones are between 140 cm.-45 cm., their widths are between 36 cm.-14 cm., their thicknesses are between 14 cm.-3.5 cm. Most of the samples are in rectangular and shapes and some are rectangular-prizmal. 3 of them have Foot gravestones. 30 head gravestones have large wadded turban (kavuk, turban (sarık, fez (fes and kerchief. All of the gravestones have inscriptions. 5 of them are without inscriptions. Inscriptions on 3 samples were laid diagonal, and 27 of them were laid in a linear system, 30 of head ve foot gravesones are in good condition. 3 samples are broken. 19 of 30 gravestones are men’s and 11 are of women’s. While scarping was used for making gravestones are scarping, and painting were used as the maindecoration tecnique is seen on foot stone with the painting technique,inscriptions of stones were coloured. Most of the inscriptions anddecorations are relieved. Decoration subjects seen on gravestones arephytomorphic, geometric, objective and calligraphic. “Osmanlı Dönemi Başlıklı Ortakent Mezar Taşları” konulu bu incelemede Ortakent Mezarlığında bulunan 30 mezar taşı araştırılarak, çalışmamızda yer almıştır. Mezar taşlarında tarihi bilinen bütün örnekler XVII. yy. ikinci çeyreği ile XX. yy. ilk çeyreği arasındadır. 30 örnekten 20 tanesi XVIII. yy., 9 tanesi XIX. yy., 1 tanesi XX. yy.dır. Mezar taşlarının hepsi mermer malzemeden yapılmıştır. Mezar taşlarının büyükten küçüğe doğru boyları 140 cm. ile 45 cm., genişlikleri 36 cm. ile 14 cm., kalınlıkları ise 14 ile 3.5 cm. arasında değişmektedir. Ortakent’te bulunan mezar taşları şahideli tiptedir. Örneklerin çoğu dikdörtgen gövdelidir. 3

Fuel management of mixed reactor type power plant systems

International Nuclear Information System (INIS)

Csom, Gyula

1988-01-01

Breeding gain in symbiotic nuclear power plant system consisting of both thermal and fast breeder reactors depends on the characteristics and the ratio of thermal and fast reactors. The composition of the symbiotic power plant systems was determined for equilibrium and plutonium deficient systems. According to natural uranium utilization, symbiotic power plant systems are not less efficient than the systems containing only fast breeders. Depleted uranium can be applied in both types of systems. Reprocessing demands of the symbiotic power plant sytems were determined. (V.N.) 23 figs.; 1 tab

Lightning protection system analysis at Multipurpose Reactor G A. Siwabessy building

International Nuclear Information System (INIS)

Teguh-Sulistyo

2003-01-01

Analysis to the part of lightning protection system at Multi Purpose Reactor GA Siwabessy (RSG-GAS) have been done. Observation examined the damage of some part of the earthing system caused by human error of chemically system. The analysis performed some assumptions and simulations to the points of lightning stroke. From this analysis obtained that the reactor building do not have vertical finial which can protect effectively to the whole reactor building and auxiliary building. Installing some new finials at some places are needed to protect building therefore the reactor building and auxiliary building well safe from lighting stroke

Real-time numerical simulation with high efficiency for an experimental reactor system

International Nuclear Information System (INIS)

Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan

2006-01-01

The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)

Review of the treat upgrade reactor scram system reliability analysis

International Nuclear Information System (INIS)

Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

1984-10-01

In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

International Nuclear Information System (INIS)

Morrone, A.; Brussalis, W.G.

1975-01-01

The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de