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Sample records for chooz reactor

  1. From double Chooz to triple Chooz - neutrino physics at the Chooz reactor complex

    International Nuclear Information System (INIS)

    Huber, Patrick; Kopp, Joachim; Lindner, Manfred; Rolinec, Mark; Winter, Walter

    2006-01-01

    We discuss the potential of the proposed Double Chooz reactor experiment to measure the neutrino mixing angle sin 2 2θ 13 . We especially consider systematical uncertainties and their partial cancellation in a near and far detector operation, and we discuss implications of a delayed near detector startup. Furthermore, we introduce Triple Chooz, which is a possible upgrade scenario assuming a second, larger far detector, which could start data taking in an existing cavern five years after the first far detector. We review the role of the Chooz reactor experiments in the global context of future neutrino beam experiments. We find that both Double Chooz and Triple Chooz can play a leading role in the search for a finite value of sin 2 2θ 13 . Double Chooz could achieve a sensitivity limit of ∼ 2.10 -2 at the 90% confidence level after 5 years while the Triple Chooz setup could give a sensitivity below 10 -2

  2. The double chooz reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Botella, I Gil [CIEMAT, Basic Research Department, Avenida Complutense, 22, 28040 Madrid (Spain)], E-mail: ines.gil@ciemat.es

    2008-05-15

    The Double Chooz reactor neutrino experiment will be the next detector to search for a non vanishing {theta}{sub 13} mixing angle with unprecedented sensitivity, which might open the way to unveiling CP violation in the leptonic sector. The measurement of this angle will be based in a precise comparison of the antineutrino spectrum at two identical detectors located at different distances from the Chooz nuclear reactor cores in France. Double Chooz is particularly attractive because of its capability to measure sin{sup 2} (2{theta}{sub 13}) to 3{sigma} if sin{sup 2}(2{theta}{sub 13}) > 0.05 or to exclude sin{sup 2}(2{theta}{sub 13}) down to 0.03 at 90% C.L. for {delta}m{sup 2} = 2.5 x 10{sup -3} eV{sup 2} in three years of data taking with both detectors. The construction of the far detector starts in 2008 and the first neutrino results are expected in 2009. The current status of the experiment, its physics potential and design and expected performance of the detector are reviewed.

  3. Chooz B

    International Nuclear Information System (INIS)

    Barillot, Pascale; Baize, Jean-Marc

    1997-01-01

    This EDF press communique give information related to the exploitation of the Chooz B NPP. A calendar of the Chooz B1 and B2 NPPs exploitation is given as well as information about the local economic impact. The exploitation of the PWR reactors of the French nuclear sector corresponds to a accumulated experience of 600 year-reactor. Significant technological evolution has been recorded, namely in the test-control system, the turbo-alternator group 'Arabelle' and in the vapor generators. The reactor safety is based on the high professionalism of the exploitation personnel, on the computer-assisted behaviour allowing the choice of operators and on the conception based upon the experience accumulated by the French nuclear power plants, equivalent to 600 year-reactor operation. EDF operates a system of continual surveillance which allows the monitoring the environmental effects caused by the NPP exploitation. The following issues concerning the environment impact are reported in this document: - The effluent releases in the environment; - Health studies conducted in the NPPs' neighbourhood; - New authorizations for waste release; - Radioactive waste management. The report also mentions the French-Belgian partnership in the PWR construction, the socio-economic regional impact of the EDF activities related with the Chooz NPP operation, and the partnership with the associated service companies. Six appendices are attached to the report containing the following information: - A general layout of Chooz NPP; - Chooz B key figures; Chooz B key data; - Security and public information; - Evolution of PWR system in France; - World's nuclear systems

  4. Chooz A: a model for the dismantling of water-cooled reactors

    International Nuclear Information System (INIS)

    Anon.

    2017-01-01

    The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

  5. Chooz A, First Pressurized Water Reactor to be Dismantled in France - 13445

    Energy Technology Data Exchange (ETDEWEB)

    Boucau, Joseph [Westinghouse Electric Company, 43 rue de l' Industrie, Nivelles (Belgium); Mirabella, C. [Westinghouse Electric France, Orsay (France); Nilsson, Lennart [Westinghouse Electric Sweden, Vaesteraas (Sweden); Kreitman, Paul J. [Westinghouse Electric Company, Lake Bluff, IL 60048 (United States); Obert, Estelle [EDF - DPI - CIDEN, Lyon (France)

    2013-07-01

    Nine commercial nuclear power plants have been permanently shut down in France to date, of which the Chooz A plant underwent an extensive decommissioning and dismantling program. Chooz Nuclear Power Station is located in the municipality of Chooz, Ardennes region, in the northeast part of France. Chooz B1 and B2 are 1,500 megawatt electric (MWe) pressurized water reactors (PWRs) currently in operation. Chooz A, a 305 MWe PWR implanted in two caves within a hill, began operations in 1967 and closed in 1991, and will now become the first PWR in France to be fully dismantled. EDF CIDEN (Engineering Center for Dismantling and Environment) has awarded Westinghouse a contract for the dismantling of its Chooz A reactor vessel (RV). The project began in January 2010. Westinghouse is leading the project in a consortium with Nuvia France. The project scope includes overall project management, conditioning of the reactor vessel (RV) head, RV and RV internals segmentation, reactor nozzle cutting for lifting the RV out of the pit and seal it afterwards, dismantling of the RV thermal insulation, ALARA (As Low As Reasonably Achievable) forecast to ensure acceptable doses for the personnel, complementary vacuum cleaner to catch the chips during the segmentation work, needs and facilities, waste characterization and packaging, civil work modifications, licensing documentation. The RV and RV internals will be segmented based on the mechanical cutting technology that Westinghouse applied successfully for more than 13 years. The segmentation activities cover the cutting and packaging plan, tooling design and qualification, personnel training and site implementation. Since Chooz A is located inside two caves, the project will involve waste transportation from the reactor cave through long galleries to the waste buffer area. The project will end after the entire dismantling work is completed, and the waste storage is outside the caves and ready to be shipped either to the ANDRA (French

  6. International Cooperation for the Dismantling of Chooz A Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Grenouillet, J.J.; Posivak, E.

    2009-01-01

    Chooz A is the first PWR that is being decommissioned in France. The main issue that is conditioning the success of the project is the Reactor Pressure Vessel (RPV) and Reactor Vessel Internals (RVI) segmentation. Whereas Chooz A is the first and unique RPV and RVI being dismantled in France, there are many similar experiences available in the world. Thus the project team was eager to cooperate with other teams facing or being faced with the same issue. A cooperation programme was established in two separate ways: - Benefiting from experience feedback from completed RPV and RVI dismantling projects, - Looking for synergy with future RPV dismantling projects for activities such as segmentation tools design, qualification and manufacturing for example. This paper describes the implementation of this programme and how the outcome of the cooperation was used for the implementation of Chooz-A RPV and RVI segmentation project. It shows also the limits of such a cooperation. (authors)

  7. Status of the double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, L.F.G.; Kemp, E. [Universidade Estadual de Campinas (IFGW/UNICAMP), SP (Brazil). Inst. de Fisica Gleb Wataghin. Dept. de Raios Cosmicos e Cronologia; Anjos, J.C. dos; Lima Junior, H.P.; Gama, R.; Abrahao, T.; Pepe, I.M. [Centro Brasileiro de Pesquisas Fisicas (MCT/CBPF), Rio de Janeiro, RJ (Brazil)

    2012-07-01

    Full text: The neutrino oscillation experiment Double Chooz, located at the nuclear power plant of Chooz (Ardennes, France), has as purpose to measure the mixing angle {theta}{sub 13} with greater precision than the measurement of its predecessor, the Chooz experiment. For this, the Double Chooz experiment will use antineutrinos generated in both Chooz plant reactors and will make high accuracy measurements of the flux in two different baselines with identical gadolinium-doped-scintillator detectors, that uses the inverse-beta-decay positron and neutron detection as an antineutrino signature. This signature is given by a prompt signal from the positron and a delayed neutron signal, which allow to decrease significantly the uncorrelated background. Other forms of background, in particular the cosmic correlated one, are managed using different technic. The use of two equal detectors at different baselines will allow to better control our systematic errors, in particular the ones associated with the reactor and with the reactor anomaly effect. The first detector (known as the 'far detector', 1050m from the reactors) is already in operation since April 2011 and the second one (known as the 'near detector', 400m from the reactor) is under construction. Thus, in this work we present the main aspects of this neutrino oscillation experiment and results update. (author)

  8. The double chooz experiment: simulation of reactor antineutrino spectra

    International Nuclear Information System (INIS)

    Mueller, T.

    2010-01-01

    The Double Chooz experiment aims to study the oscillations of electron antineutrinos produced by the Chooz nuclear power station, located in France, in the Ardennes region. It will lead to an unprecedented accuracy on the value of the mixing angle θ 13 . Improving the current knowledge on this parameter, given by the Chooz experiment, requires a reduction of both statistical and systematic errors, that is to say not only observing a large data sample, but also controlling the experimental uncertainties involved in the production and detection of electron antineutrinos. The use of two identical detectors will cancel most of the experimental systematic uncertainties limiting the sensitivity to the value of the mixing angle. We present in this thesis, simulations of reactor antineutrino spectra that were carried out in order to control the sources of systematic uncertainty related to the production of these particles by the plant. We also discuss our work on controlling the normalization error of the experiment through the precise determination of the number of target protons by a weighing measurement and through the study of the fiducial volume of the detectors which requires an accurate modeling of neutron physics. After three years of data taking with two detectors, Double Chooz will be able to disentangle an oscillation signal for sin 2 2θ 13 ≥ 0.05 (at 3σ) or, if no oscillations are observed, to put a limit of sin 2 2θ 13 ≤ 0.03 at 90% C.L. (author) [fr

  9. Double Chooz

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Christian [Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, D-69117 Heidelberg (Germany)

    2006-05-15

    The goal of the Double Chooz reactor neutrino experiment is to search for the neutrino mixing parameter {theta}{sub 13}. Double Chooz will use two identical detectors at 150 m and 1.05 km distance from the reactor cores. The near detector is used to monitor the reactor {nu}-bar {sub e} flux while the second is dedicated to the search for a deviation from the expected (1/distance){sup 2} behavior. This two detector concept will allow a relative normalization systematic error of ca. 0.6 %. The expected sensitivity for sin{sup 2}2{theta}{sub 13} is then in the range 0.02 - 0.03 after three years of data taking. The antineutrinos will be detected in a liquid scintillator through the capture on protons followed by a gamma cascade, produced by the neutron capture on Gd.

  10. Re-examining reactor vessel embrittlement at Chooz A

    International Nuclear Information System (INIS)

    Guilleret, J.-C.

    1988-01-01

    The Chooz A PWR experienced an extended shutdown in 1987/88 following indications that the reactor vessel was embrittling more rapidly than expected. Discrepancies between the expected rate and estimates of the actual rate were not easily explained. The huge body of work done since then to establish safety margins and support restart of the plant should provide a model for the owners of other older PWRs grappling with the embrittlement issue. (author)

  11. Coherent neutrino scattering with low temperature bolometers at Chooz reactor complex

    International Nuclear Information System (INIS)

    Billard, J; Gascon, J; Jesus, M De; Carr, R; Formaggio, J A; Heine, S T; Johnston, J; Leder, A; Sibille, V; Winslow, L; Dawson, J; Lasserre, T; Figueroa-Feliciano, E; Palladino, K J; Vivier, M

    2017-01-01

    We present the potential sensitivity of a future recoil detector for a first detection of the process of coherent elastic neutrino nucleus scattering (CE ν NS). We use the Chooz reactor complex in France as our luminous source of reactor neutrinos. Leveraging the ability to cleanly separate the rate correlated with the reactor thermal power against (uncorrelated) backgrounds, we show that a 10 kg cryogenic bolometric array with 100 eV threshold should be able to extract a CE ν NS signal within one year of running. (paper)

  12. The double Chooz hardware trigger system

    Energy Technology Data Exchange (ETDEWEB)

    Cucoanes, Andi; Beissel, Franz; Reinhold, Bernd; Roth, Stefan; Stahl, Achim; Wiebusch, Christopher [RWTH Aachen (Germany)

    2008-07-01

    The double Chooz neutrino experiment aims to improve the present knowledge on {theta}{sub 13} mixing angle using two similar detectors placed at {proportional_to}280 m and respectively 1 km from the Chooz power plant reactor cores. The detectors measure the disappearance of reactor antineutrinos. The hardware trigger has to be very efficient for antineutrinos as well as for various types of background events. The triggering condition is based on discriminated PMT sum signals and the multiplicity of groups of PMTs. The talk gives an outlook to the double Chooz experiment and explains the requirements of the trigger system. The resulting concept and its performance is shown as well as first results from a prototype system.

  13. Double Chooz Improved Multi-Detector Measurements

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    The Double Chooz experiment (DC) is a reactor neutrino oscillation experiment running at Chooz nuclear power plant (2 reactors) in France. In 2011, DC first reported indication of non-zero θ13 with the far detector (FD) located at the maximum of oscillation effects (i.e. disappearance), thus challenging the CHOOZ non-observation limit. A robust observation of θ13 followed in 2012 by the Daya Bay experiments with multiple detector configurations. Since 2015 DC runs in a multi-detector configuration making thus the impact of several otherwise dominating systematics reduce strongly. DC’s unique almost "iso-flux" site, allows the near detector (ND) to become a direct accurate non-oscillation reference to the FD. Our first multi-detector results at MORIOND-2016 showed an intriguing deviation of θ13 with respect to the world average. We will address this issue in this seminar. The combined "reactor-θ13" measurement is expected to ...

  14. Sterile Neutrino Search with the Double Chooz Experiment

    Science.gov (United States)

    Hellwig, D.; Matsubara, T.; Double Chooz Collaboration

    2017-09-01

    Double Chooz is a reactor antineutrino disappearance experiment located in Chooz, France. A far detector at a distance of about 1 km from reactor cores is operating since 2011; a near detector of identical design at a distance of about 400 m is operating since begin 2015. Beyond the precise measurement of θ 13, Double Chooz has a strong sensitivity to so called light sterile neutrinos. Sterile neutrinos are neutrino mass states not taking part in weak interactions, but may mix with known neutrino states. In this paper, we present an analysis method to search for sterile neutrinos and the expected sensitivity with the baselines of our detectors.

  15. Sterile neutrino search with the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Hellwig, Denise; Bekman, Ilja; Kampmann, Philipp; Schoppmann, Stefan; Soiron, Michael; Stahl, Achim; Wiebusch, Christopher [III. Physikalisches Institut B, RWTH Aachen (Germany)

    2016-07-01

    The Double Chooz experiment is a reactor neutrino disappearance experiment located at the Chooz nuclear power plant, France. It measures the electron-antineutrino flux of the two nuclear reactors with two detectors of identical design. A far detector at a distance of about 1 km is operating since 2011; a near detector at a distance of about 400 m is operating since the end of 2014. The combination of the two detectors offers sensitivity to sterile neutrino mixing parameters. Sterile neutrinos are neutrino mass states not taking part in weak interactions, but may mix with known neutrino states. This induces additional mixing angles and mass differences. This talk describes the search for sterile neutrinos and the sensitivity of Double Chooz to the mixing angle θ{sub 14}.

  16. Status of the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, L.F.G. [Universidade Estadual de Campinas (IFGW/UNICAMP), SP (Brazil). Inst. de Fisica Gleb Wataghin. Dept. de Raios Cosmicos e Cronologia; Universite Oaris Diderot (France); Kemp, E. [Universidade Estadual de Campinas (IFGW/UNICAMP), SP (Brazil). Inst. de Fisica Gleb Wataghin. Dept. de Raios Cosmicos e Cronologia; Anjos, J.C. dos; Barbosa, A.F.; Lima Junior, H.P.; Gama, R.; Abrahao, T.; Pepe, I.M. [Centro Brasileiro de Pesquisas Fisicas (CBPF), Rio de Janeiro, RJ (Brazil); Chimenti, P. [Universidade Federal do ABC (UFABC), SP (Brazil)

    2011-07-01

    Full text: The neutrino oscillation experiment Double Chooz, located at the nuclear power plant of Chooz (Ardennes, France), has as purpose to measure the mixing angle {theta}{sub 13} with greater precision than the measurement of its predecessor, the Chooz experiment. In four years of data acquisition, we expect to decrease the current limit of sin{sup 2}(2{theta}{sub 13}) from 0.2 to 0.03, contributing for the determination of this parameter that is still open in the neutrino mixing matrix. For this, the Double Chooz experiment will use antineutrinos generated in both Chooz plant reactors and will make high accuracy measurements of the flux in two different distances with identical detectors. The first detector (the 'far detector') is already in operation since January 2011 and the construction of the second one (the'{sup n}ear detector') should be completed on the beginning of 2012. Thus, in this work we present the main aspects of this neutrino oscillation experiment and preliminary results of the detector performance. (author)

  17. Neutrino oscillations - the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Th. [CEA Saclay, Dept. d' Astrophysique, de Physique des Particules de Physique Nucleaire et de l' Instrumentation Associee (DSM/DAPNIA/SPP/APC), 91- Gif sur Yvette (France)

    2007-07-01

    {theta}{sub 13} is the mixing angle that couples the field of the neutrino number 3 (the heaviest) to the electron field. The Double Chooz experiment will use 2 identical detectors, near the Chooz nuclear reactor cores to measure the last undetermined mixing angle {theta}{sub 13}. The basic principle of the multi-detector concept is the cancellation of the reactor-induced systematic errors. The first detector will be installed in the existing underground laboratory (1050 meters away from the plant station) that was used in the first Chooz experiment in the nineties. The second detector will be constructed from 2009 in a new neutrino laboratory, located down a 45 m well that will be excavated 300 m away from the reactors. An average visible neutrino rate of 55 (550) events per day is expected to be detected inside the far (near) detector, taking into account the various inefficiencies, if no oscillations. The near detector will perform a measurement of the anti-neutrino flux and its energy spectrum with an unprecedented accuracy and for a long period (3 years). These huge statistics will also be exploited to monitor changes in the relative amounts of U{sup 235} and Pu{sup 239} in the core, paving the way to use neutrino detection for safeguards applications. (A.C.)

  18. Measurement of the mixing leptonic parameter θ13 at the Double Chooz reactor antineutrino experiment

    International Nuclear Information System (INIS)

    Durand, V.

    2012-01-01

    The Double Chooz experiment aims at measuring the neutrino mixing parameter θ13 by studying the oscillations of de ν-bar e produced by the Chooz nuclear reactors located in France. The experimental concept consists in comparing the signal of two identical 10.3 m 3 detectors, allowing to cancel most of the experimental systematic uncertainties. The near detector, whose goal is the flux normalization and a measurement without oscillation, is expected to be delivered in 2013. The farthest detector from the source is taking data since April 2011 and is sensitive to θ 13 , which is expected to affect both the rate and the shape of the measured de ν-bar e . In this thesis, are first presented the Double Chooz experiment, with its ν-bar e source, its detection method, and the expected signal and backgrounds. In order to perform a selection, important quantities have to be reconstructed, calibrated, and saved in data files. The channel time offsets determination, the energy and vertex reconstruction algorithm CocoReco, the reconstruction packages of the Common Trunk, and the light trees maker Cheetah are especially presented. Concerning the data analysis, all the selection cuts and results for signal and backgrounds are discussed, particularly the multiplicity cut, the multiple off time window method, the lithium veto cut, and the cosmogenic 9 Li background studies. The Double Chooz experiment observed 8,249 de ν-bar e candidates in 227.93 days in its far detector only. The reactor antineutrino flux prediction used the Bugey 4 flux measurement after correction for differences in core composition. The expectation in case of no-oscillation is 8,937 events and this deficit is interpreted as evidence for ν-bar e disappearance. From a rate and shape analysis, is found sin 2 2θ = 0,109± 0,030 (stat) ± 0,025 (syst), with Δm 2 31 = 2,32 x 10 -3 eV 2 , while the no-oscillation hypothesis is even excluded at 2.9 σ. (author) [fr

  19. Results from the Double Chooz experiment

    Directory of Open Access Journals (Sweden)

    Kaneda Michiru

    2016-01-01

    Full Text Available Recent results from the Double Chooz experiment on the neutrino mixing angle θ13 are presented. Two detectors are located at distances of 400m and 1050m from the reactor cores of the Chooz nuclear power plant, to measure the original neutrino flux from the reactor cores and the disappearance of neutrinos, respectively. The Far Detector has taken data since 2011 while the Near Detector started the data taking in 2014. The latest far detector only result with gadolinium capture events is sin2 2θ13=0.090+0.032.-0.029. Studies using hydrogen capture events also have been improved and the combined result of gadolinium and hydrogen capture events is obtained as sin2 2θ13 = 0.088 ± 0.033.

  20. The U238 antineutrino spectrum in the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Haag, Nils; Oberauer, Lothar; Potzel, Walter; Schreckenbach, Klaus [Technische Universitaet, Muenchen (Germany); Lachenmaier, Tobias [Eberhard Karls Universitaet, Tuebingen (Germany)

    2011-07-01

    The DoubleChooz experiment aims at the determination of the unknown neutrino mixing parameter {Theta}{sub 13}. Two liquid scintillator detectors will measure an electron antineutrino disappearance at the Chooz site in the French ardennes. In order to improve the sensitivity, the antineutrino spectrum emitted by the Chooz reactor cores has to be determined with high accuracy. This talk focusses on the U238 spectrum, which is the only contributing spectrum, that was not measured until now. The final U238 beta spectrum is presented, and its implementation into the analysis framework is shown.

  1. Latest results from the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Goeger-Neff, Marianne [Physik Department E15, Technische Universitaet Muenchen (Germany); Collaboration: Double Chooz-Collaboration

    2016-07-01

    Double Chooz aims at a precise measurement of the neutrino mixing angle θ{sub 13} through the disappearance of reactor electron antineutrinos. The experiment relies on the measurement of neutrino flux and spectrum with two identical detectors at 400 m and 1000 m from the reactor cores of two nuclear power reactors. anti ν{sub e} are detected by inverse beta decay on free protons in 8.3 tons of Gd-loaded liquid scintillator, providing a unique delayed coincidence signature. Double Chooz has been running since 2011 with the far detector only, providing the first indication for non-zero θ{sub 13} with reactor antineutrinos. With a rate+shape analysis of 467.90 live days from 2011-2013 we obtain a value of sin {sup 2} 2 θ{sub 13}=0.090{sup +0.032}{sub -0.029}. Data taking with the near detector has started beginning of 2015, allowing a significant reduction of both reactor and detector related systematic uncertainties. The talk reviews the most recent results obtained with the far detector only and discuss first data from the near detector.

  2. Double-Chooz: a search for {theta}{sub 13}

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Th. [DSM/DAPNIA/SPP, CEA/Saclay, 91191 Gif-sur-Yvette (France)

    2005-12-15

    The Double-Chooz experiment goal is to search for a non-vanishing value of the {theta}{sub 13} neutrino mixing angle. This is the last step to accomplish prior moving towards a new era of precision measurements in the lepton sector. The current best constraint on the third mixing angle comes from the CHOOZ reactor neutrino experiment sin{sup 2}(2{theta}{sub 13})<0.2 (90% C.L., {delta}m{sub atm}{sup 2}=2.0eV{sup 2}). Double-Chooz will explore the range of sin{sup 2}(2{theta}{sub 13}) from 0.2 to 0.03-0.02, within three years of data taking. The improvement of the CHOOZ result requires an increase in the statistics, a reduction of the systematic error below one percent, and a careful control of the backgrounds. Therefore, Double-Chooz will use two identical detectors, one at 150 m and another at 1.05 km distance from the Chooz nuclear cores. In addition, we will use the near detector as a ''state of the art'' prototype to investigate the potential of neutrinos for monitoring the civil nuclear power plants. The plan is to start operation with two detectors in 2008, and to reach a sensitivity sin{sup 2}(2{theta}{sub 13}) of 0.05 in 2009, and 0.03-0.02 in 2011.

  3. Measurement of the leptonic mixing parameter θ13 at the reactor antineutrino experiment Double Chooz

    International Nuclear Information System (INIS)

    Durand, V.

    2014-01-01

    The Double Chooz experiment aims at measuring the neutrino mixing parameter θ 13 by studying the oscillations of ν-bar e produced by the Chooz nuclear reactors located in France. The experimental concept consists in comparing the signal of 2 identical 10.3 m 3 detectors, allowing the cancellation of most of the experimental systematic uncertainties. The near detector, whose goal is the flux normalization and a measurement without oscillation, is expected to be delivered in 2013. The farthest detector from the source is taking data since april 2011 and is sensitive to θ 13 , which is expected to affect both the rate and the shape of measured ν-bar e . In 102 days with far detector only, 4121 ν-bar e candidates have been observed, while the expectation in case of no-oscillation is 4344 ± 165 events. This deficit is interpreted as evidence for ν-bar e disappearance. From a rate and shape analysis, it is found that sin 2 (θ 13 ) = [0.086±0.041(stat.)±0.030(syst)] with Δm 31 2 =2.4*10 -3 eV 2 , while the no-oscillation hypothesis is even excluded at 94.6%. We present in this document the concept and detection method of the Double Chooz experiment, along with its analysis strategy and first results on θ 13 . (author)

  4. Double-Chooz: a search for {theta}{sub 13}

    Energy Technology Data Exchange (ETDEWEB)

    Mention, G. [APC/PCC, College de France, 75005 Paris (France)

    2005-08-15

    The Double-Chooz experiment goal is to search for a non-vanishing value of the {theta}{sub 13} neutrino mixing angle. This is the last step to accomplish prior moving towards a new era of precision measurements in the lepton sector. The current best constraint on the third mixing angle comes from the CHOOZ reactor neutrino experiment sin{sup 2}(2{theta}{sub 13}) < 0.2-0.14 (90% C.L., {delta}m{sub atm}{sup 2}=2.0-2.410{sup -3} eV{sup 2}). Double-Chooz will explore the range of sin{sup 2}(2{theta}{sub 13}) from 0.2 to 0.03-0.02, within three years of data taking. The improvement of the CHOOZ result requires an increase in the statistics, a reduction of the systematic error below one percent, and a careful control of the backgrounds. Therefore, Double-Chooz will use two identical detectors, one at 150 m and another at 1.05 km distance from the Chooz nuclear cores. In addition, we will use the near detector as a 'state of the art' prototype to investigate the potential of neutrinos for monitoring the civil nuclear power plants. The plan is to start operation with two detectors in 2008, and to reach a sin{sup 2}(2{theta}{sub 13}) sensitivity of 0.05 in 2009, and 0.03-0.02 in 2011.

  5. Double Chooz: Results and Perspectives

    Science.gov (United States)

    Franke, A. J.

    2013-04-01

    The Double Chooz experiment has been observing ν from two reactor cores at the Chooz nuclear power station in Ardennes, France, with a single 10.3 m fiducial volume Gd-doped liquid scintillator detector at a flux-weighted average baseline of ˜1050 m. This article reviews results achieved with a detector live time of 227.93 days and exposure of 33.71 GW-ton-years. A total of 8,249 candidate ν events have been observed, compared to an expected 8,937 events in the null-oscillation case: this deficit is interpreted as evidence for ν disappearance. A fit to the observed neutrino rate and spectral shape gives a best-fit value of sin2(2θ13)=0.109±0.030 (stat.)±0.025 (syst.) at a mass-squared splitting of Δm312=2.32×10-3 eV. The null-oscillation hypothesis is excluded by the data at 99.8% CL (2.9σ). The Double Chooz Near Detector is under construction, and analysis efforts to measure neutrino directionality, test Lorentz violation, and measure backgrounds in situ are underway.

  6. Background-independent measurement of θ13 in Double Chooz

    International Nuclear Information System (INIS)

    Abe, Y.; Anjos, J.C. dos; Barriere, J.C.; Baussan, E.; Bekman, I.; Bergevin, M.; Bezerra, T.J.C.; Bezrukov, L.; Blucher, E.; Buck, C.; Busenitz, J.; Cabrera, A.; Caden, E.; Camilleri, L.; Carr, R.; Cerrada, M.; Chang, P.-J.; Chauveau, E.

    2014-01-01

    The oscillation results published by the Double Chooz Collaboration in 2011 and 2012 rely on background models substantiated by reactor-on data. In this analysis, we present a background-model-independent measurement of the mixing angle θ 13 by including 7.53 days of reactor-off data. A global fit of the observed antineutrino rates for different reactor power conditions is performed, yielding a measurement of both θ 13 and the total background rate. The results on the mixing angle are improved significantly by including the reactor-off data in the fit, as it provides a direct measurement of the total background rate. This reactor rate modulation analysis considers antineutrino candidates with neutron captures on both Gd and H, whose combination yields sin 2 (2θ 13 )=0.102±0.028(stat.)±0.033(syst.). The results presented in this study are fully consistent with the ones already published by Double Chooz, achieving a competitive precision. They provide, for the first time, a determination of θ 13 that does not depend on a background model

  7. Double Chooz: Results and Perspectives

    Energy Technology Data Exchange (ETDEWEB)

    Franke, A.J., E-mail: ajfranke@nevis.columbia.edu [Columbia University, New York, New York 10027 (United States)

    2013-04-15

    The Double Chooz experiment has been observing ν{sup ¯}{sub e} from two reactor cores at the Chooz nuclear power station in Ardennes, France, with a single 10.3 m{sup 3} fiducial volume Gd-doped liquid scintillator detector at a flux-weighted average baseline of ∼1050 m. This article reviews results achieved with a detector live time of 227.93 days and exposure of 33.71 GW-ton-years. A total of 8,249 candidate ν{sup ¯}{sub e} events have been observed, compared to an expected 8,937 events in the null-oscillation case: this deficit is interpreted as evidence for ν{sup ¯}{sub e} disappearance. A fit to the observed neutrino rate and spectral shape gives a best-fit value of sin{sup 2}(2θ{sub 13})=0.109±0.030(stat.)±0.025(syst.) at a mass-squared splitting of Δm{sub 31}{sup 2}=2.32×10{sup −3} eV{sup 2}. The null-oscillation hypothesis is excluded by the data at 99.8% CL (2.9σ). The Double Chooz Near Detector is under construction, and analysis efforts to measure neutrino directionality, test Lorentz violation, and measure backgrounds in situ are underway.

  8. Design studies for the Double Chooz trigger

    International Nuclear Information System (INIS)

    Cucoanes, Andi Sebastian

    2009-01-01

    The main characteristic of the neutrino mixing effect is assumed to be the coupling between the flavor and the mass eigenstates. Three mixing angles (θ 12 , θ 23 , θ 13 ) are describing the magnitude of this effect. Still unknown, θ 13 is considered very small, based on the measurement done by the CHOOZ experiment. A leading experiment will be Double Chooz, placed in the Ardennes region, on the same site as used by CHOOZ. The Double Chooz goal is the exploration of ∝80% from the currently allowed θ 13 region, by searching the disappearance of reactor antineutrinos. Double Chooz will use two similar detectors, located at different distances from the reactor cores: a near one at ∝150 m where no oscillations are expected and a far one at 1.05 km distance, close to the first minimum of the survival probability function. The measurement foresees a precise comparison of neutrino rates and spectra between both detectors. The detection mechanism is based on the inverse β-decay. The Double Chooz detectors have been designed to minimize the rate of random background. In a simplified view, two optically separated regions are considered. The target, filled with Gd-doped liquid scintillator, is the main antineutrino interaction volume. Surrounding the target, the inner veto region aims to tag the cosmogenic muon background which hits the detector. Both regions are viewed by photomultipliers. The Double Chooz trigger system has to be highly efficient for antineutrino events as well as for several types of background. The trigger analyzes discriminated signals from the central region and the inner veto photomultipliers. The trigger logic is fully programmable and can combine the input signals. The trigger conditions are based on the total energy released in event and on the PMT groups multiplicity. For redundancy, two independent trigger boards will be used for the central region, each of them receiving signals from half of the photomultipliers. A third trigger board

  9. Double Chooz experiment

    International Nuclear Information System (INIS)

    Palomares, C.

    2009-01-01

    Double Chooz will use two identical detectors at different distances from the Chooz nuclear power station to search for a non-vanishing value of θ 13 , and, hopefully, to open the way to experiments aspiring to discover CP violation in the leptonic sector. The far detector is expected to be operative by the end of 2009. Installation of the near detector will occur in 2010. Double Chooz has the capacity to exclude sin 2 (2θ 13 ) 31 2 = 2.5 x 10 -3 eV 2 with three years of data running both near and far detectors. (author)

  10. Design studies for the Double Chooz trigger

    Energy Technology Data Exchange (ETDEWEB)

    Cucoanes, Andi Sebastian

    2009-07-24

    The main characteristic of the neutrino mixing effect is assumed to be the coupling between the flavor and the mass eigenstates. Three mixing angles ({theta}{sub 12}, {theta}{sub 23}, {theta}{sub 13}) are describing the magnitude of this effect. Still unknown, {theta}{sub 13} is considered very small, based on the measurement done by the CHOOZ experiment. A leading experiment will be Double Chooz, placed in the Ardennes region, on the same site as used by CHOOZ. The Double Chooz goal is the exploration of {proportional_to}80% from the currently allowed {theta}{sub 13} region, by searching the disappearance of reactor antineutrinos. Double Chooz will use two similar detectors, located at different distances from the reactor cores: a near one at {proportional_to}150 m where no oscillations are expected and a far one at 1.05 km distance, close to the first minimum of the survival probability function. The measurement foresees a precise comparison of neutrino rates and spectra between both detectors. The detection mechanism is based on the inverse {beta}-decay. The Double Chooz detectors have been designed to minimize the rate of random background. In a simplified view, two optically separated regions are considered. The target, filled with Gd-doped liquid scintillator, is the main antineutrino interaction volume. Surrounding the target, the inner veto region aims to tag the cosmogenic muon background which hits the detector. Both regions are viewed by photomultipliers. The Double Chooz trigger system has to be highly efficient for antineutrino events as well as for several types of background. The trigger analyzes discriminated signals from the central region and the inner veto photomultipliers. The trigger logic is fully programmable and can combine the input signals. The trigger conditions are based on the total energy released in event and on the PMT groups multiplicity. For redundancy, two independent trigger boards will be used for the central region, each of

  11. Nuclear safety and radiation protection report of the Chooz nuclear facilities - 2010; Rapport sur la surete nucleaire et la radioprotection des installations nucleaires de Chooz - 2010

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-06-15

    This safety report was established under the article 21 of the French law no. 2006-686 of June 13, 2006 relative to nuclear safety and information transparency. It presents, first, the facilities of the Chooz nuclear power plant (Ardennes (FR)): 2 PWR reactors in operation (Chooz B, INB 139 and 144) and one partially dismantled PWR reactor (Chooz A, INB 163). Then, the nuclear safety and radiation protection measures taken regarding the facilities are reviewed: nuclear safety definition, radiation protection of intervening parties, safety and radiation protection improvement paths, crisis management, external and internal controls, technical situation of facilities, administrative procedures in progress. The incidents and accidents which occurred in 2010, are reported as well as the radioactive and non-radioactive (chemical, thermal) effluents discharge in the environment. Finally, The radioactive materials and wastes generated by the facilities are presented and sorted by type of waste, quantities and type of conditioning. Other environmental impacts (noise, microbial proliferation in cooling towers) are presented with their mitigation measures. Actions in favour of transparency and public information are presented as well. The document concludes with a glossary and a list of recommendations from the Committees for health, safety and working conditions. (J.S.)

  12. Nuclear safety and radiation protection report of the Chooz nuclear facilities - 2011

    International Nuclear Information System (INIS)

    2012-01-01

    This safety report was established under the article 21 of the French law no. 2006-686 of June 13, 2006 relative to nuclear safety and information transparency. It presents, first, the facilities of the Chooz nuclear power plant (Ardennes (FR)): 2 PWR reactors in operation (Chooz B, INB 139 and 144) and one partially dismantled PWR reactor (Chooz A, INB 163). Then, the nuclear safety and radiation protection measures taken regarding the facilities are reviewed: nuclear safety definition, radiation protection of intervening parties, safety and radiation protection improvement paths, crisis management, external and internal controls, technical situation of facilities, administrative procedures in progress. The incidents and accidents which occurred in 2011, are reported as well as the radioactive and non-radioactive (chemical, thermal) effluents discharge in the environment. Finally, The radioactive materials and wastes generated by the facilities are presented and sorted by type of waste, quantities and type of conditioning. Other environmental impacts (noise, microbial proliferation in cooling towers) are presented with their mitigation measures. Actions in favour of transparency and public information are presented as well. The document concludes with a glossary followed by the viewpoint of the Committees for health, safety and working conditions. (J.S.)

  13. Nuclear safety and radiation protection report of the Chooz nuclear facilities - 2010

    International Nuclear Information System (INIS)

    2011-06-01

    This safety report was established under the article 21 of the French law no. 2006-686 of June 13, 2006 relative to nuclear safety and information transparency. It presents, first, the facilities of the Chooz nuclear power plant (Ardennes (FR)): 2 PWR reactors in operation (Chooz B, INB 139 and 144) and one partially dismantled PWR reactor (Chooz A, INB 163). Then, the nuclear safety and radiation protection measures taken regarding the facilities are reviewed: nuclear safety definition, radiation protection of intervening parties, safety and radiation protection improvement paths, crisis management, external and internal controls, technical situation of facilities, administrative procedures in progress. The incidents and accidents which occurred in 2010, are reported as well as the radioactive and non-radioactive (chemical, thermal) effluents discharge in the environment. Finally, The radioactive materials and wastes generated by the facilities are presented and sorted by type of waste, quantities and type of conditioning. Other environmental impacts (noise, microbial proliferation in cooling towers) are presented with their mitigation measures. Actions in favour of transparency and public information are presented as well. The document concludes with a glossary and a list of recommendations from the Committees for health, safety and working conditions. (J.S.)

  14. Chooz B1 start-up marked by total computer control

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The start-up of Chooz B1, the first of EdF's latest generation 1450 MWe N4 reactors, marks the first use in a nuclear power plant of a fully computerised control room. Working in partnership with EdF, Sema Group designed and supplied the advanced command and control system for this plant. (Author)

  15. The neutrino experiment Double Chooz and data analysis with the near detector

    Energy Technology Data Exchange (ETDEWEB)

    Franke, Michael Werner

    2016-03-07

    During the last years there has been a huge progress in the field of neutrino physics. Neutrino oscillations are well established and almost all parameters, except a possible CP-violating phase, are determined to high precision. One experiment providing a precise measurement of the neutrino mixing angle θ{sub 13} is the Double Chooz reactor antineutrino experiment. The reactor antineutrinos are detected via the inverse beta decay in two identical liquid scintillator based detectors. A few years ago, the value of θ{sub 13} was unknown and only an upper limit existed. Double Chooz was the first reactor antineutrino experiment presenting a result for a nonzero value of θ{sub 13}. The value for sin{sup 2}2θ{sub 13} from the latest Double Chooz publication is 0.090{sup +0.032}{sub -0.029}. As part of this thesis, an infrastructure for filling the Double Chooz near detector was established and 190 m{sup 3} of detector liquids were prepared successfully. The filling process was optimized to allow an efficient filling of the near detector. The total operation time was reduced to only 22 days. Compared to the far detector filling time of 2 months, this is a great improvement. The development of a completely new level measurement system was as well part of this thesis. Due to the excellent performance of the level measurement system, the hard restrictions for the safety of the Double Chooz detector were met during the entire filling process. Several power glitches and network failures did not harm the system and did not result in any loss of data. These irregularities and the simple maintenance and repair possibilities certify the success of the design concept for the new level measurement system. For this thesis, data from the Double Chooz near detector with a total live time of 110.4 days was used. The mass concentrations of uranium and thorium in the near detector were determined using BiPo coincidences. These events originate from the β-decay of {sup 214}Bi and {sup

  16. From the measurement of the θ13 mixing angle to the search for geo-neutrinos: studying νe-bare with Double Chooz and Borexino

    International Nuclear Information System (INIS)

    Roncin, Romain

    2014-01-01

    Double Chooz is a reactor neutrino oscillation experiment which aims at measuring the θ 13 mixing angle thanks to two identical detectors located at different distances from the two reactors of the Chooz nuclear power plant, in the French Ardennes. While the near detector will start taking data in fall 2014 to normalize the flux of the neutrinos emitted by the nuclear reactors, the far detector is running since April 2011 and allows to observe the neutrinos disappearance through the neutrino oscillation phenomenon. This thesis is also dedicated to the Borexino experiment which was designed to observe solar neutrinos. Due to its low background level as well as its position in a nuclear free country, Italy, Borexino is also sensitive to geo-neutrinos. This thesis presents both the Double Chooz and Borexino experiments, from the description of the detectors to the main results, with a special attention to the background and its rejection. Studies on the neutrino directionality with these two experiments are also detailed. In the case of Double Chooz, since the neutrinos are coming from the two nuclear reactors, the precision of the analysis method can be assessed. This thesis presents also for the first time the possibility to retrieve the initial direction of the neutrinos when the neutrons created in the inverse beta decay reactions are captured on hydrogen. In the case of Borexino, neutrino directionality information could facilitate the discrimination between geo-neutrinos and neutrinos from nuclear reactors. (author) [fr

  17. Chooz-A Steam Generators Characterization

    International Nuclear Information System (INIS)

    Aitammar, Laurie

    2016-01-01

    EDF nuclear waste management requires a deep understanding of characterization, classification and waste sorting operations. In fact, French nuclear waste management defines several classes with specific management, treatment and storage facilities. Based on particular criteria, the more the radiological risk of the nuclear waste is important, the more its management will be complex and expensive. During the dismantling of the first French pressurized reactor Chooz-A, decontamination of the primary water circuits (not including the reactor vessel), the steam generators and the pressurizer have been carried out in order to reduce their activity levels. Thanks to these decontamination operations, and a specific characterization methodology, EDF was able to Re-classify the 4 steam generators and store them in one piece at the ANDRA Very Low Level Activity disposal facilities instead of the Low and Intermediate Level activity one. This re-classification allowed EDF to avoid important cutting and packaging processes. To characterize and declare Chooz-A SG activity, EDF-CIDEN used a methodology defined by the French institute of atomic energy, CEA. The method is based on external gamma spectrometry measurements performed with NaI collimated detectors, associated with MERCURAD simulations providing the transfer functions for the detectors and activity sources. Internal measurements are carried out with a CZT (CdZnTe) probe inside the SG tubes to refine the 3D model. In fact, the primary side represents the main source of activity, and understanding its contamination distribution is important to reduce the model and calculation uncertainties. Measurements eventually provided SG 60 Co global activity, from which the activities of other radionuclides of the spectrum were determined using scaling factors. The final activity declaration takes into account the standard deviation of the measurements in order to cover the uncertainties of the methodology. Thereby, the declaration

  18. Background estimation of cosmic-ray induced neutrons in Chooz site water veto tank for possible future Ricochet Deployment

    Science.gov (United States)

    Silva, James

    2017-09-01

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CE νNS) using metallic superconducting and germanium semi-conducting detectors with sub-keV thresholds placed near a neutrino source such as the Chooz Nuclear Reactor Complex. In this poster, we present an estimate of the flux of cosmic-ray induced neutrons, which represent an important background in any (CE νNS) search, based on reconstructed cosmic ray data from the Chooz Site. We have simulated a possible Ricochet deployment at the Chooz site in GEANT4 focusing on the spallation neutrons generated when cosmic rays interact with the water tank veto that would surround our detector. We further simulate and discuss the effectiveness of various shielding configurations for optimizing the background levels for a future Ricochet deployment.

  19. Progress in RPV-examination of the Chooz-A vessel (and the French procedures, its new developments (MIS5))

    Energy Technology Data Exchange (ETDEWEB)

    Samman, J; Martin, E; Lacroix, R [Electricite de France (EDF), 93 - Saint-Denis (France). Groupe des Labs.

    1988-12-31

    This document deals with the French Chooz-A reactor. It describes the method used for in-service inspection of Reactor Pressure Vessels (RPV). The ultrasonic testing procedure is described, showing its advantages and limitations. The supplementary ultrasonic examination is also described, as well as the validation of underclad cracks detection and sizing. Historical data is also provided. (TEC).

  20. Chooz-B1, the new Electricite de France PWR: calculation scheme of neutron leakages from the reactor cavity

    International Nuclear Information System (INIS)

    Champion, G.; Thiriet, A.; Vergnaud, T.; Bourdet, L.; Nimal, J.C.; Brandicourt, G.

    1987-04-01

    A new calculation scheme has been set up to assess the neutron field characteristics inside French PWR. In order to take into account multiple neutron scattering and the complexity of the reactor geometry, the use of Monte-Carlo methods have been heavily increased. They are coupled with classical SN.-methods. The main goal aimed at was to find out the neutron field characteristics at the level of the reactor pit openings. These radiation reference sources will be used to check the neutron shielding efficiencies. The new calculation scheme has been applied to CHOOZ-B1, the first unit of the new N4 program. The former results have been compared with the measurement results related to PALUEL-I and II PWR, two units of the previous P4 program. Although the core and the geometry are not entirely similar, it is possible to check with confidence the calculation results along the vessel and at the core midplane level with the measurement results at the same locations. It appears that they are in good agreement. Consequently, the new calculation scheme appears reliable

  1. Building a nuclear power plant from A to Z: Chooz B as an example

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    From the design studies to the tests and reactor divergence, building a nuclear power plant involves strictly planned operations. This issue deals with the operations carried out first in the design offices and then on the site, in order to achieve such a vast and sophisticated industrial project: organizing, designing, engineering, assembling, testing, manufacturing and transporting heavy components. These various phases are shown through the example of Chooz B, a nuclear power plant with two 1.450 MW units, which is being built in the Ardennes, and whose first reactor will diverge in 1995. (author)

  2. Chasing theta-13 with the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Thierry [CEA/DSM/IRFU/SPP, Gif-sur-Yvette (France)

    2009-07-01

    Neutrino oscillation physics is entering a precision measurement area. The smallness of the theta-13 neutrino mixing angle is still enigmatic and should be resolved. Double Chooz will use two identical detectors near the Chooz nuclear power station to search for a non vanishing theta-13, and hopefully open the way to experiments aspiring to discover CP violation in the leptonic sector.

  3. Double Chooz and non Asian efforts towards {theta}{sub 13}

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Thierry [CEA/Saclay and Laboratoire Astroparticle and Cosmologie Institut de Recherche des Lois Fondamentales de l' Univers Service de Physique des Particules 91191 Gif-s-Yvette (France)], E-mail: thierry.lasserre@cea.fr

    2008-11-01

    Neutrino oscillation physics is entering a precision measurement area. The smallness of the {theta}{sub 13} neutrino mixing angle is still enigmatic and should be resolved. Double Chooz will use two identical detectors near the Chooz nuclear power station to search for a non vanishing {theta}{sub 13}, and hopefully open the way to experiments aspiring to discover CP violation in the leptonic sector. The Angra project aims to prevail over the Double Chooz experiment if the third neutrino oscillation channel is not discovered in the forthcoming years.

  4. Measuring the θ13 mixing angle with the two Double Chooz detectors

    International Nuclear Information System (INIS)

    Sibille, Valerian

    2016-01-01

    The Double Chooz experiment aims at accurately measuring the value of the θ 13 leptonic mixing angle. To this intent, the experiment makes the most of two identical detectors - filled with gadolinium-loaded liquid scintillator - observing ν e -bar's released by the two 4.25 GWth nuclear reactors of the French Chooz power plant. The so-called 'far detector' - located at an average distance of 1050 m from the two nuclear cores - has been taking data since April 2011. The 'near detector' - at an average distance of 400 m from the cores - has monitored the reactor since December 2014. The θ 13 mixing parameter leads to an energy dependent disappearance of ν e -bar's as they propagate from the nuclear cores to the detection sites, which allows for a fit of the sin 2 2θ 13 value. By reason of correlations between the detectors and an iso-flux layout, the detection systematics and the ν e -bar flux uncertainty impairing the θ 13 measurement are dramatically suppressed. In consequence, the precision of the θ 13 measurement is dominated by the uncertainty on the backgrounds and the relative normalisation of the ν e -bar-rates. The main background originates from the decay of β n -emitters - generated by μ-spallation - within the detector itself. The energy spectra of these cosmogenic isotopes have been simulated and complemented by a diligent error treatment. These predictions have been successfully compared to the corresponding data spectra, extracted by means of an active veto, whose performance has been studied at both sites. The rate of cosmogenic background remaining within the ν e -bar candidates has also been assessed. Additionally, the normalisation of the ν e -bar rates, bound to the number of target protons within each detector, has been evaluated. All this work was part of the first Double Chooz multi-detector results, yielding sin 2 2θ 13 =0.111 ± 0.018. (author) [fr

  5. The DoubleChooz DAQ systems.

    CERN Multimedia

    CERN. Geneva

    2012-01-01

    The Double Chooz (DC) reactor anti-neutrino experiment consists of a neutrino detector and a large area Outer Veto detector. A custom data-acquisition (DAQ) system written in Ada language for all the sub-detector in the neutrino detector systems and a generic object oriented data acquisition system for the Outer Veto detector were developed. Generic object-oriented programming was also used to support several electronic systems to be readout providing a simple interface for any new electronics to be added given its dedicated driver. The core electronics of the experiment is based on FADC electronics (500MHz sampling rate), therefore a data-reduction scheme has been implemented to reduce the data volume per trigger. A dynamic data-format was created to allow dynamic reduction of each trigger before data is written to disk. The decision is based on low level information that determines the relevance of each trigger. The DAQ is structured internally into two types of processors: several read-out processors readi...

  6. Characterization of Magnetic Field Immersed Photomultipliers from Double Chooz Experiment. Design and Construction of their Magnetic Shields; Caracterizacion de los fotomultiplicadores del experimento Double Chooz bajo campo magnetico y diseno y construccion de sus blindajes magneticos

    Energy Technology Data Exchange (ETDEWEB)

    Valdivia Valero, F J

    2007-12-28

    Flavour oscillations of neutrinos are a quantum-mechanical effect widely demonstrated. It is explained through interferences of their mass eigenstates, therefore, belonging to the physical area beyond the Standard Model. This work deals with the CIEMAT collaboration in the neutrino experiment Double Chooz. Such an experiment aims to measure the mixture angle {theta}{sub 1}3, one of the PMNS leptonic mixture matrix, with a un reached-before sensibility by decrease of systematic errors. For this, two identical scintillator detectors, equipped with PMT's, will be sited at different distances from two reactors located in the nuclear power plant CHOOZ B (France). The electronic neutrino flux from these reactors will be compared, explaining its deficit by flavour oscillations of these particles. The identity of both detectors will be diminished by the magnetic field effects on the PMT's response. Therefore, this study serves as for quantifying such an effects as for fitting the magnetic shields design that minimize them. Shielding measurements and final design of magnetic shields as much as the effect these ones cause in the PMT's response immersed in a monitored magnetic field are presented. (Author) 85 refs.

  7. Characterization of Magnetic Field Immersed Photomultipliers from Double Chooz Experiment. Design and Construction of their Magnetic Shields; Caracterizacion de los fotomultiplicadores del experimento Double Chooz bajo campo magnetico y diseno y construccion de sus blindajes magneticos

    Energy Technology Data Exchange (ETDEWEB)

    Valdivia Valero, F. J.

    2007-12-28

    Flavour oscillations of neutrinos are a quantum-mechanical effect widely demonstrated. It is explained through interferences of their mass eigenstates, therefore, belonging to the physical area beyond the Standard Model. This work deals with the CIEMAT collaboration in the neutrino experiment Double Chooz. Such an experiment aims to measure the mixture angle {theta}{sub 1}3, one of the PMNS leptonic mixture matrix, with a un reached-before sensibility by decrease of systematic errors. For this, two identical scintillator detectors, equipped with PMT's, will be sited at different distances from two reactors located in the nuclear power plant CHOOZ B (France). The electronic neutrino flux from these reactors will be compared, explaining its deficit by flavour oscillations of these particles. The identity of both detectors will be diminished by the magnetic field effects on the PMT's response. Therefore, this study serves as for quantifying such an effects as for fitting the magnetic shields design that minimize them. Shielding measurements and final design of magnetic shields as much as the effect these ones cause in the PMT's response immersed in a monitored magnetic field are presented. (Author) 85 refs.

  8. Characterization of Magnetic Field Immersed Photomultipliers from Double Chooz Experiment. Design and Construction of their Magnetic Shields

    International Nuclear Information System (INIS)

    Valdivia Valero, F. J.

    2007-01-01

    Flavour oscillations of neutrinos are a quantum-mechanical effect widely demonstrated. It is explained through interferences of their mass eigenstates, therefore, belonging to the physical area beyond the Standard Model. This work deals with the CIEMAT collaboration in the neutrino experiment Double Chooz. Such an experiment aims to measure the mixture angle θ 1 3, one of the PMNS leptonic mixture matrix, with a un reached-before sensibility by decrease of systematic errors. For this, two identical scintillator detectors, equipped with PMT's, will be sited at different distances from two reactors located in the nuclear power plant CHOOZ B (France). The electronic neutrino flux from these reactors will be compared, explaining its deficit by flavour oscillations of these particles. The identity of both detectors will be diminished by the magnetic field effects on the PMT's response. Therefore, this study serves as for quantifying such an effects as for fitting the magnetic shields design that minimize them. Shielding measurements and final design of magnetic shields as much as the effect these ones cause in the PMT's response immersed in a monitored magnetic field are presented. (Author) 85 refs

  9. Installation, commissioning and performance of the trigger system of the Double Chooz experiment and the analysis of hydrogen capture neutrino events

    Energy Technology Data Exchange (ETDEWEB)

    Lucht, Sebastian

    2013-11-18

    The Double Chooz experiment is a reactor antineutrino experiment located at Chooz, a small town in the Ardennes region in the north of France close to the Belgium border. The aim of the experiment is to measure the leptonic mixing angle θ{sub 13}. The antineutrino flux is measured by two identical detectors at different distances from the reactor cores used as neutrino source, in a so called ''disappearance'' experiment. Double Chooz is a precision experiment because previous experiments indicated a small value of θ{sub 13}. Therefore, the systematic uncertainties introduced by background events and detector related components have to be as small as possible. The detector and all electronic components have been designed accordingly. The first part of this thesis describes the trigger and timing system of the Double Chooz experiment. This system triggers the data acquisition of the detector. It continuously monitors the signals of the photomultiplier tubes (PMTs) of the detector. These signals are summed for groups of PMTs (group signal) and for all PMTs (sum signal). The group signals are the input signals to the trigger system. They are discriminated by one threshold resulting in a multiplicity condition on the number of active group signal discriminators. The sum signal is discriminated by four thresholds. The default trigger configuration for the Double Chooz experiment is based on a combination on the sum signal discriminators and the multiplicity condition. In addition, the trigger system provides a common clock signal for all data acquisition components and an online event classification to allow an online data reduction. The trigger system was installed and commissioned in 2011. In this thesis the commissioning of the trigger system and its performance is presented. Furthermore the development and tests of possible improvements for the trigger system are presented and discussed. The second part of this thesis introduces a complementary

  10. CHOOZ-A expert assessment program

    International Nuclear Information System (INIS)

    Bouat, M.; Godin, R.

    1993-01-01

    CHOOZ-A Nuclear Power Plant, the first French-Belgian PWR unit (300 MWe) was definitively shut down at the end of October 1991, after 24 years in operation. Since summer 1991, the steering committee of the French (EDF) Lifetime Project has initiated a large inquiry to the different technical specialists of EDF and external organizations, trying to define a wide expert assessment program on this plant. The aim is to improve the knowledge of aging mechanisms such as those observed on the 52 PWR French nuclear power plants (900 and 1,300 MWe), and contribute to the validation of non-destructive in-service testing methods. This paper presents the retained CHOOZ-A expert assessment program and technical lines followed during its set up. First major project stages are described, then technical choices are explained, and at last the final program is presented with the specific content of each expert assessment. The definitive program is scheduled for a three year period starting at the moment of final shutdown license acquisition, with a provisional total budget of more than US $10 million

  11. Electromagnetic Compatibility (EMC) and Noise Background Testing for Double Chooz PMT System

    International Nuclear Information System (INIS)

    Cepero, J. R.; Encabo Fernandez, F. J.; Pepe, I.; Verdugo, A.

    2009-01-01

    The Double Chooz PMT system is a HV/signal splitter. In this report is presented an electromagnetic compatibility and background noise testing for the Double Chooz PMT system. It was possible to proceed the EMC testing on different grounding configurations of PMT splitter due to its special PCB design, endowed of jumping points and a metal box ground electrode. (Author)

  12. Electromagnetic Compatibility (EMC) and Noise Background Testing for Double Chooz PMT System

    Energy Technology Data Exchange (ETDEWEB)

    Cepero, J. R.; Encabo Fernandez, F. J.; Pepe, I.; Verdugo, A.

    2009-05-21

    The Double Chooz PMT system is a HV/signal splitter. In this report is presented an electromagnetic compatibility and background noise testing for the Double Chooz PMT system. It was possible to proceed the EMC testing on different grounding configurations of PMT splitter due to its special PCB design, endowed of jumping points and a metal box ground electrode. (Author)

  13. The Chooz power station: ten years of operation

    International Nuclear Information System (INIS)

    Teste du Bailler, Andre

    1977-01-01

    The switching into actual service of the Chooz plant, the first pressurized water reactor ever built in France, occurred on 3rd april 1967. Ten years later, one can establish a highly positive balance schedule of plant's operation whose availability is satisfactory, except the mechanical failure which occurred during the startup. The behavior of the equipment, in particular of the components of the primary loop, was satisfactory in its whole since it allowed the gradual increase in capacity by 15% with respect to the initial design. It allowed also the achievment of noticeable progress in the design of equipment intended for the new power stations. Interesting results have also been obtained in radioprotection, working conditions of the staff and environment protection fields. Finally, the training of the operating teams has been closely followed, whether it concerned the operators directly affected by plant operation or the trainees gathered in a school specially organized for this purpose and transferred since to a training Center [fr

  14. Estimation of the systematic uncertainties of the measurement of the neutrino mixing angle θ{sub 13} related to the trigger system of the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Stueken, David Anselm

    2013-10-14

    The Double Chooz experiment, located in the Ardennes region next to the CHOOZ-B nuclear power plant, is a reactor antineutrino experiment to measure neutrino oscillations. It has been designed as precision experiment to measure the neutrino mixing angel θ{sub 13} with highest possible accuracy due to its small value close to zero. The electron antineutrino flux emitted by the reactor cores is measured by two identical neutrino detectors located at different distances from the reactor cores. Each detector consist of a 10.3 m{sup 3} target volume filled with liquid scintillator and surrounded by 390 photomultiplier tubes. The far detector is located 1.05 km away from the reactor cores to be most sensitive to oscillation effects. The unoscillated neutrino flux is measured by the near detector located 400 m away from the reactor cores. In order to reduce background events and other sources resulting in systematic uncertainties, special requirements have been demanded for all detector components and electronic systems. In this context, a most efficiently operating data acquisition system is essential. The subsystem responsible to start data storage for events of interest is the so called ''trigger system''. The design concept of the Double Chooz trigger system introduces two redundancy concepts in order to trigger the data acquisition in the most robust and efficient way: The trigger decision is based on a combination of an energy threshold and the number of active photomultiplier tubes (multiplicity condition). Secondly, the system is divided into two identical but independently operating subsystems for most robust operations of the full system. Additionally, the two subsystem provide the possibility to measure the efficiency of the system. Apart from generating the trigger signal for the data acquisition, the system provides an online event classification in order to adjust the amount of stored data for each event type. After one and a half year

  15. The chooz a expert survey program and its main conclusions for plant life management

    International Nuclear Information System (INIS)

    Barthelet, B.; Heuze, A.; Hennart, J.C.; Havard, P.

    2001-01-01

    Because of the importance of PWR components life management represents for Electricity Companies, significant R and D programs are dedicated to identifying and analysing mechanisms and damage rates of the different degradation modes of these components, systems and structures. To assess R and D assumptions and to validate non destructive test results through reviews, expert survey programs on in-situ equipment may enhance the knowledge about most of the various phenomena involved. In this regard, an extensive program was launched after the Chooz A NPP was decommissioned in 1991, after 24 years in operation. This program gathered EDF, IPSN, FRAMATOME, ELECTRABEL and TRACTEBEL into partnership. The expert survey program was performed in various laboratories between 1995 and 1999 and includes: - on-site non destructive testing before sampling, - and metallurgical and mechanical tests performed on samples taken from the nuclear and non nuclear part of the unit. The expert survey program performed by Utilities in various laboratories involved the following equipment: - reactor vessel and internal equipment, - reactor coolant system (dissimilar metal welds, SS welds, cast austenitic ferritic steels), - feedwater plant piping (erosion-corrosion), - electric cables susceptible of temperature and irradiation induced ageing, - anchoring in civil engineering structures, - main primary circuit concerning activation measurement. In conclusion, the extensive Chooz A expert survey program yields numerous significant results. The main outcomes will contribute to validate non destructive tests and enhance our knowledge of some degradation mechanisms of often quite similar components present in units in operation. It is worthy to note that this program is of prime importance for operation feedback; the cost of the whole study amounts to approximately 10 Million Euros. (author)

  16. Computer-assisted command and control for Chooz B

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    A new type of command and control system has been installed at the Chooz B nuclear power plant. A key feature of the system is a man/machine interface designed to provide operators with on-screen assistance more quickly than ever. (Author)

  17. Double Chooz sensitivity and backgrounds studies to search for θ13 leptonic mixing angle

    International Nuclear Information System (INIS)

    Mention, G.

    2005-06-01

    The Double Chooz experiment will study the oscillations of electron antineutrinos produced by the Chooz nuclear power station to measure θ 13 mixing angle. The current knowledge on this parameter, provided ny the Chooz experiment, can be improved by reducing statistical and systematic errors. A large data sample will be collected to improve the former one. Two identical detectors will be built to cancel most of experimental systematic uncertainties involved in production and detection processes. Special care will be dedicated to backgrounds generated by natural radioactivity and cosmic ray interactions. In the hereby thesis, we describe our simulation studies to compute θ 13 sensitivity and assess the discovery potential of the experiment. We concentrated particularly on quantifying the detector related systematic errors that would limit the θ 13 sensitivity. Background related systematic errors such as the accidental events produced by the radioactivity of the photomultiplier tubes, correlated events from neutrons as well as a hypothetical background (mimicking the oscillation pattern) were taken into account. After 3 years, Double Chooz will be able to disentangle an oscillation signal for sin 2 (2*θ 13 ) > 0.05 (at 3*σ) or, if no oscillations were observed, to put a limit of sin 2 (2θ 13 ) < 0.03 at 90% C.L. (author)

  18. Correlated background and impact on the measurement of θ13 with the Double Chooz detector

    International Nuclear Information System (INIS)

    Remoto, Alberto

    2012-01-01

    The Double Chooz experiment uses antineutrinos emitted from the Chooz nuclear power plant (France) to measure the oscillation mixing parameter θ 13 . By using two detectors at different baselines, a precise measurement of antineutrinos disappearance is anticipated. The Far detector has been taking physics data since April 2011, while the Near detector is under construction. Data from April 13, 2011 to March 30, 2012 taken with the Far detector only have been analyzed and an indication for antineutrino disappearance, consistent with the current neutrino oscillation hypothesis, has been found. The best fit value for the neutrino mixing parameter sin 2 (2θ 13 ) is 0.109 ± 0.030(stat.) ± 0.025(syst.). This thesis present an accurate description of the Double Chooz experiment, with particular emphasis on the Far detector and its acquisition system. The main focus of the thesis is the accurate study of the correlated background affecting the Double Chooz antineutrinos sample and its impact on the measurement of the mixing parameter θ 13 . A general overview of the current experimental scenario which aim to the characterization of the neutrino oscillation is also provided, focusing on the recent results obtained in this field. (author) [fr

  19. Development of a level-1 trigger and timing system for the Double Chooz neutrino experiment

    International Nuclear Information System (INIS)

    Reinhold, Bernd

    2009-01-01

    The measurement of the mixing angle θ 13 is the goal of several running and planned experiments. The experiments are either accelerator based (super)beam experiments (e.g. MINOS, T2K, Nova) or reactor anti-neutrino disappearance experiments (e.g. Daya Bay, RENO or Double Chooz). In order to measure or constrain θ 13 with the Double Chooz experiment the overall systematic errors have to be controlled at the one-percent or sub-percent level. The limitation of the systematic errors is achieved through various means and techniques. E.g. the experiment consists of two identical detectors at different baselines, which allow to make a differential anti-neutrino flux measurement, where basically only relative normalisation errors remain. The requirements on the systematic errors put also strong constraints on the quality of all components and materials used for both detectors, most prominently on the stability and radiopurity of the scintillator, the photomultiplier tubes, the vessels containing the detector liquids and the shielding against ambient radioactivity. The readout electronics, trigger and data acquisition system have to operate reliably as an integrated and highly efficient whole over several years. The trigger is provided by the Level-1 Trigger and Timing System, which is the subject of this thesis. It has to provide a highly efficient trigger (at the 0.1% level) for neutrino-induced events as well as for several types of background events. Its decision is realized in hardware and based on energy depositions in the muon veto and the target region. The Level-1 Trigger and Timing System furthermore provides a common System Clock and an absolute timestamp for each event. The Level-1 Trigger and Timing System consists of two types of VME modules, several Trigger Boards and a Trigger Master Board, which have been custom-designed and developed in the electronics workshop of our institute for this experiment and purpose, starting in 2005. In this thesis all

  20. Development of a level-1 trigger and timing system for the Double Chooz neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Reinhold, Bernd

    2009-02-25

    The measurement of the mixing angle {theta}{sub 13} is the goal of several running and planned experiments. The experiments are either accelerator based (super)beam experiments (e.g. MINOS, T2K, Nova) or reactor anti-neutrino disappearance experiments (e.g. Daya Bay, RENO or Double Chooz). In order to measure or constrain {theta}{sub 13} with the Double Chooz experiment the overall systematic errors have to be controlled at the one-percent or sub-percent level. The limitation of the systematic errors is achieved through various means and techniques. E.g. the experiment consists of two identical detectors at different baselines, which allow to make a differential anti-neutrino flux measurement, where basically only relative normalisation errors remain. The requirements on the systematic errors put also strong constraints on the quality of all components and materials used for both detectors, most prominently on the stability and radiopurity of the scintillator, the photomultiplier tubes, the vessels containing the detector liquids and the shielding against ambient radioactivity. The readout electronics, trigger and data acquisition system have to operate reliably as an integrated and highly efficient whole over several years. The trigger is provided by the Level-1 Trigger and Timing System, which is the subject of this thesis. It has to provide a highly efficient trigger (at the 0.1% level) for neutrino-induced events as well as for several types of background events. Its decision is realized in hardware and based on energy depositions in the muon veto and the target region. The Level-1 Trigger and Timing System furthermore provides a common System Clock and an absolute timestamp for each event. The Level-1 Trigger and Timing System consists of two types of VME modules, several Trigger Boards and a Trigger Master Board, which have been custom-designed and developed in the electronics workshop of our institute for this experiment and purpose, starting in 2005. In

  1. Determination of the direction to a source of antineutrinos via inverse beta decay in Double Chooz

    Science.gov (United States)

    Nikitenko, Ya.

    2016-11-01

    To determine the direction to a source of neutrinos (and antineutrinos) is an important problem for the physics of supernovae and of the Earth. The direction to a source of antineutrinos can be estimated through the reaction of inverse beta decay. We show that the reactor neutrino experiment Double Chooz has unique capabilities to study antineutrino signal from point-like sources. Contemporary experimental data on antineutrino directionality is given. A rigorous mathematical approach for neutrino direction studies has been developed. Exact expressions for the precision of the simple mean estimator of neutrinos' direction for normal and exponential distributions for a finite sample and for the limiting case of many events have been obtained.

  2. Measurement of the neutrino mixing angle θ13 with the Double Chooz experiment

    Science.gov (United States)

    Ostrovskiy, Igor

    2009-10-01

    The neutrino mixing angle θ13 is last one which value is still unknown. Measuring the θ13 is important for completing our understanding of three flavor neutrino oscillations. Moreover, leptonic CP violation could only be measured in case the value of θ13 is not zero. The current best limit (^2(2θ13)Ardennes. Described in this talk, is another experiment, Double Chooz, that is being prepared at the same site. The Double Chooz experiment offers several fundamental improvements and is aiming to surpass the current limit by an order of magnitude (^2(2θ13) < 0.03). Details of the detector design, overview of systematic errors and expected sensitivity, as well as current status of the experiment are presented.

  3. Nuclear energy. The innovations of the N4 reactor

    International Nuclear Information System (INIS)

    Anon.

    1998-01-01

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

  4. Realization of the low background neutrino detector Double Chooz. From the development of a high-purity liquid and gas handling concept to first neutrino data

    Energy Technology Data Exchange (ETDEWEB)

    Pfahler, Patrick

    2012-12-17

    Neutrino physics is one of the most vivid fields in particle physics. Within this field, neutrino oscillations are of special interest as they allow to determine driving oscillation parameters, which are collected as mixing angles in the leptonic mixing matrix. The exact knowledge of these parameters is the main key for the investigation of new physics beyond the currently known Standard Model of particle physics. The Double Chooz experiment is one of three reactor disappearance experiments currently taking data, which recently succeeded to discover a non-zero value for the last neutrino mixing angle {Theta}{sub 13}. As successor of the CHOOZ experiment, Double Chooz will use two detectors with improved design, each of them now composed of four concentrically nested detector vessels each filled with different detector liquid. The integrity of this multi-layered structure and the quality of the used detector liquids are essential for the success of the experiment. Within this frame, the here presented work describes the production of two detector liquids, the filling and handling of the Double Chooz far detector and the installation of all necessary hardware components therefore. In order to meet the strict requirements existing for the detector liquids, all components were individually selected in an extensive material selection process at TUM, which compared samples from different companies for their key properties: density, transparency, light yield and radio purity. Based on these measurements, the composition of muon veto scintillator and buffer liquid were determined. For the production of the detector liquids, a simple surface building close to the far detector site was upgraded into a large-scale storage and mixing facility, which allowed to separately, mix, handle and store 90 m{sup 3} of muon veto scintillator and 110 m{sup 3} of buffer liquid. For the muon veto scintillator, a master-solution composed of 4800 l LAB, 180 kg PPO and 1.8 kg of bis/MSB was

  5. Realization of the low background neutrino detector Double Chooz. From the development of a high-purity liquid and gas handling concept to first neutrino data

    International Nuclear Information System (INIS)

    Pfahler, Patrick

    2012-01-01

    Neutrino physics is one of the most vivid fields in particle physics. Within this field, neutrino oscillations are of special interest as they allow to determine driving oscillation parameters, which are collected as mixing angles in the leptonic mixing matrix. The exact knowledge of these parameters is the main key for the investigation of new physics beyond the currently known Standard Model of particle physics. The Double Chooz experiment is one of three reactor disappearance experiments currently taking data, which recently succeeded to discover a non-zero value for the last neutrino mixing angle Θ 13 . As successor of the CHOOZ experiment, Double Chooz will use two detectors with improved design, each of them now composed of four concentrically nested detector vessels each filled with different detector liquid. The integrity of this multi-layered structure and the quality of the used detector liquids are essential for the success of the experiment. Within this frame, the here presented work describes the production of two detector liquids, the filling and handling of the Double Chooz far detector and the installation of all necessary hardware components therefore. In order to meet the strict requirements existing for the detector liquids, all components were individually selected in an extensive material selection process at TUM, which compared samples from different companies for their key properties: density, transparency, light yield and radio purity. Based on these measurements, the composition of muon veto scintillator and buffer liquid were determined. For the production of the detector liquids, a simple surface building close to the far detector site was upgraded into a large-scale storage and mixing facility, which allowed to separately, mix, handle and store 90 m 3 of muon veto scintillator and 110 m 3 of buffer liquid. For the muon veto scintillator, a master-solution composed of 4800 l LAB, 180 kg PPO and 1.8 kg of bis/MSB was produced and

  6. Utilization of FADC for reconstruction and analysis of the background data in the Chooz neutrino experiment

    International Nuclear Information System (INIS)

    Veron, Didier

    1997-01-01

    This thesis describes a particular contribution to the Chooz experiment. The latter looks for the oscillations, over a distance of 1 km, of antineutrons emitted by two nuclear reactors. The electron-type antineutrinos are detected through their inverse beta interaction with a target's proton. The neutron is detected through its capture by a gadolinium nucleus revealed by an 8 MeV gamma emission. In the first part we describe the reconstruction of events as simulated by the GIANT software. We show that the positron's and neutron's stopping point can actually be reconstructed with an accuracy of 10 and 20 cm respectively. In the second part, we proceed to the analysis of the calibration's data as recorded with Fast Wave Form Digitizers. This confirms the reliability of the Monte-Carlo results and allows measurement of both the neutrons' capture probability and time by the target gadolinium. The last part deals with the background (reactor turned off) data analysis and the pin-pointing of its various sources. In order to reduce their contribution, we define spatial cuts. These cuts' reliability is validated by analysis of data obtained not only with a neutron source, but also with neutrons issued from cosmic rays. We end up with a background contribution of two to three events per day, about ten times less than the expected neutrino rate at full reactor power. (author)

  7. Measurement of θ_1_3 in Double Chooz using neutron captures on hydrogen with novel background rejection techniques

    International Nuclear Information System (INIS)

    Abe, Y.; Appel, S.; Abrahão, T.; Almazan, H.

    2016-01-01

    The Double Chooz collaboration presents a measurement of the neutrino mixing angle θ_1_3 using reactor ν̄_e observed via the inverse beta decay reaction in which the neutron is captured on hydrogen. This measurement is based on 462.72 live days data, approximately twice as much data as in the previous such analysis, collected with a detector positioned at an average distance of 1050 m from two reactor cores. Several novel techniques have been developed to achieve significant reductions of the backgrounds and systematic uncertainties. Accidental coincidences, the dominant background in this analysis, are suppressed by more than an order of magnitude with respect to our previous publication by a multi-variate analysis. These improvements demonstrate the capability of precise measurement of reactor ν̄_e without gadolinium loading. Spectral distortions from the ν̄_e reactor flux predictions previously reported with the neutron capture on gadolinium events are confirmed in the independent data sample presented here. A value of sin"2 2θ_1_3=0.095_−_0_._0_3_9"+"0"."0"3"8(stat+syst) is obtained from a fit to the observed event rate as a function of the reactor power, a method insensitive to the energy spectrum shape. A simultaneous fit of the hydrogen capture events and of the gadolinium capture events yields a measurement of sin"2 2θ_1_3=0.088±0.033(stat+syst).

  8. Liquid scintillators and liquefied rare gases for particle detectors. Background-determination in Double Chooz and scintillation properties of liquid argon

    International Nuclear Information System (INIS)

    Hofmann, Martin Alexander

    2012-01-01

    Evidence for physics beyond the well-established standard model of particle physics is found in the sector of neutrino physics, in particular in neutrino oscillations, and in experimental hints requiring the presence of Dark Matter. Neutrino oscillations demand the neutrinos to be massive and at least four additional parameters, three mixing angles and one phase, are introduced. A non-vanishing value for the third mixing angle, θ 13 , has only recently been found, amongst others by the reactor antineutrino disappearance experiment Double Chooz. This experiment detects anti ν e 's by means of the Inverse Beta Decay (IBD), which has a clear signature that can very effectively be discriminated from most of the background. However, some background still survives the selection cuts applied to the data, partly induced by radioactivity. In order to determine the amount of radioimpurities in the detector, germanium spectroscopy measurements and neutron activation analyses have been carried out for various parts of the Double Chooz far detector. A dedicated Monte-Carlo simulation was performed to obtain the singles event rate induced by the identified radioimpurities in the fiducial volume of Double Chooz. In the present thesis, parts from the outer detector systems, as well as components of the inner detector liquids were measured. In sum, a singles rate of less than 0.35 Hz above the antineutrino detection threshold of 0.7 MeV has been found. This is by far below the design goal of Double Chooz of ∝ 20 Hz. The analysis of bismuth-polonium (BiPo) coincidences in the first Double Chooz data allows to directly determine the number of decays from the U- and the Th-decay chain in the active detector parts. Assuming radioactive equilibrium, concentrations of (1.71±0.08).10 -14 (g)/(g) for uranium and (8.16±0.49).10 -14 (g)/(g) for thorium have been found, which are also well below the design goal of Double Chooz (2.10 -13 (g)/(g)). Both gamma spectroscopy measurements and

  9. Liquid scintillators and liquefied rare gases for particle detectors. Background-determination in Double Chooz and scintillation properties of liquid argon

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, Martin Alexander

    2012-11-27

    Evidence for physics beyond the well-established standard model of particle physics is found in the sector of neutrino physics, in particular in neutrino oscillations, and in experimental hints requiring the presence of Dark Matter. Neutrino oscillations demand the neutrinos to be massive and at least four additional parameters, three mixing angles and one phase, are introduced. A non-vanishing value for the third mixing angle, {theta}{sub 13}, has only recently been found, amongst others by the reactor antineutrino disappearance experiment Double Chooz. This experiment detects anti {nu}{sub e}'s by means of the Inverse Beta Decay (IBD), which has a clear signature that can very effectively be discriminated from most of the background. However, some background still survives the selection cuts applied to the data, partly induced by radioactivity. In order to determine the amount of radioimpurities in the detector, germanium spectroscopy measurements and neutron activation analyses have been carried out for various parts of the Double Chooz far detector. A dedicated Monte-Carlo simulation was performed to obtain the singles event rate induced by the identified radioimpurities in the fiducial volume of Double Chooz. In the present thesis, parts from the outer detector systems, as well as components of the inner detector liquids were measured. In sum, a singles rate of less than 0.35 Hz above the antineutrino detection threshold of 0.7 MeV has been found. This is by far below the design goal of Double Chooz of {proportional_to} 20 Hz. The analysis of bismuth-polonium (BiPo) coincidences in the first Double Chooz data allows to directly determine the number of decays from the U- and the Th-decay chain in the active detector parts. Assuming radioactive equilibrium, concentrations of (1.71{+-}0.08).10{sup -14}(g)/(g) for uranium and (8.16{+-}0.49).10{sup -14}(g)/(g) for thorium have been found, which are also well below the design goal of Double Chooz (2.10{sup -13

  10. Characterization, modelization and optimization of the Double Chooz acrylic vessels: physics impact

    International Nuclear Information System (INIS)

    Queval, R.

    2010-01-01

    Double Chooz is one of the new generation experiments designed to measure the last unknown leptonic mixing angle, θ 13 . Il studies the oscillations of electronic antineutrinos produced by the Chooz nuclear power plant. Our knowledge on θ 13 can be improved by reducing both statistical errors (increasing both the detector size and run time) and systematic errors (using identical near and far detectors). Also, special care is dedicated to analyze backgrounds generated by natural radioactivity and cosmic ray interactions. The work presented here mostly focuses on Target and Gamma Catcher acrylic vessels, in the core of the detector. A new material designed for the experiment was developed since no pre-existing material met the required conditions. This material was characterized optically, to maximize light transmission and reduce the induced dead zone. With this in mind, a physics study of the Target vessel design was performed to identify any spectral distortion or count rate modification it could induce. As far as backgrounds are concerned, the material is radio-pure enough so that the singles rate coming from acrylic is negligible compared to the one from the photomultiplier tubes. An unknown single source is coming from the external contamination in the detector, such as dust. This is why we defined cleanliness goals for the detector fabrication and integration. These goals were met, thanks to well-defined protocols and careful team work. After three years of data taking, Double Chooz could disentangle an oscillation signal at 3α for sin 2 2θ 13 ≥ 0.05 - 0.06. If no oscillations were observed, the experiment could give an upper limit on sin 2 2θ 13 of 0.02 - 0.03 at 90 % C.L. (author) [fr

  11. Neutrinos oscillations researches near a nuclear reactor

    International Nuclear Information System (INIS)

    Laiman, M.

    1999-01-01

    This thesis deals with the research of neutrinos oscillations near the Chooz B nuclear power plant in the Ardennes. The first part presents the framework of the researches and the chosen detector. The second part details the antineutrinos flux calculus from the reactors and the calculus of the expected events. The analysis procedure is detailed in the last part from the calibration to the events selection. (A.L.B.)

  12. Reactor anti-neutrinos: measurement of the θ13 leptonic mixing angle and search for potential sterile neutrinos

    International Nuclear Information System (INIS)

    Collin, A.

    2014-01-01

    The Double Chooz experiment aims to measure the θ 13 mixing angle through the disappearance -induced by the oscillation phenomenon - of anti-neutrinos produced by the Chooz nuclear reactors. In order to reduce systematic uncertainties, the experiment relies on the relative comparison of detected signals in two identical liquid scintillator detectors. The near one, giving the normalization of the emitted flux, is currently being built and will be delivered in spring 2014. The far detector, sensitive to θ 13 , is located at about one kilometer and is taking data since 2011. In this first phase of the experiment, the far detector data are compared to a prediction of the emitted neutrino flux to estimate θ 13 . In this thesis, the Double Chooz experiment and its analysis are presented, especially the background studies and the rejection of parasitic signals due to light emitted by photo-multipliers. Neutron fluxes between the different detector volumes impact the definition of the fiducial volume of neutrino interactions and the efficiency of detection. Detailed studies of these effects are presented. As part of the Double Chooz experiment, studies were performed to improve the prediction of neutrino flux emitted by reactors. This work revealed a deficit of observed neutrino rates in the short baseline experiments of last decades. This deficit could be explained by an oscillation to a sterile state. The Stereo project aims to observe a typical signature of oscillations: the distortion of neutrino spectra both in energy and baseline. This thesis presents the detector concept and simulations as well as sensitivity studies. Background sources and the foreseen shielding are also discussed. (author) [fr

  13. The reactor antineutrino anomalies

    Energy Technology Data Exchange (ETDEWEB)

    Haser, Julia; Buck, Christian; Lindner, Manfred [Max-Planck-Institut fuer Kernphysik, Heidelberg (Germany)

    2016-07-01

    Major discoveries were made in the past few years in the field of neutrino flavour oscillation. Nuclear reactors produce a clean and intense flux of electron antineutrinos and are thus an essential neutrino source for the determination of oscillation parameters. Most currently the reactor antineutrino experiments Double Chooz, Daya Bay and RENO have accomplished to measure θ{sub 13}, the smallest of the three-flavour mixing angles. In the course of these experiments two anomalies emerged: (1) the reanalysis of the reactor predictions revealed a deficit in experimentally observed antineutrino flux, known as the ''reactor antineutrino anomaly''. (2) The high precision of the latest generation of neutrino experiments resolved a spectral shape distortion relative to the expected energy spectra. Both puzzles are yet to be solved and triggered new experimental as well as theoretical studies, with the search for light sterile neutrinos as most popular explanation for the flux anomaly. This talk outlines the two reactor antineutrino anomalies. Discussing possible explanations for their occurrence, recent and upcoming efforts to solve the reactor puzzles are highlighted.

  14. Utilization of FADC for reconstruction and analysis of the background data in the Chooz neutrino experiment; Utilisation des FADC pour la reconstruction et l`analyse des donnees de bruit de fond dans l`experience neutrino de Chooz

    Energy Technology Data Exchange (ETDEWEB)

    Veron, Didier [Universite Claude Bernard, 69 - Lyon (France)

    1997-03-25

    This thesis describes a particular contribution to the Chooz experiment. The latter looks for the oscillations, over a distance of 1 km, of antineutrons emitted by two nuclear reactors. The electron-type antineutrinos are detected through their inverse beta interaction with a target`s proton. The neutron is detected through its capture by a gadolinium nucleus revealed by an 8 MeV gamma emission. In the first part we describe the reconstruction of events as simulated by the GIANT software. We show that the positron`s and neutron`s stopping point can actually be reconstructed with an accuracy of 10 and 20 cm respectively. In the second part, we proceed to the analysis of the calibration`s data as recorded with Fast Wave Form Digitizers. This confirms the reliability of the Monte-Carlo results and allows measurement of both the neutrons` capture probability and time by the target gadolinium. The last part deals with the background (reactor turned off) data analysis and the pin-pointing of its various sources. In order to reduce their contribution, we define spatial cuts. These cuts` reliability is validated by analysis of data obtained not only with a neutron source, but also with neutrons issued from cosmic rays. We end up with a background contribution of two to three events per day, about ten times less than the expected neutrino rate at full reactor power. (author) 81 refs., 152 figs.,43 tabs.

  15. Non-standard interaction effects at reactor neutrino experiments

    International Nuclear Information System (INIS)

    Ohlsson, Tommy; Zhang, He

    2009-01-01

    We study non-standard interactions (NSIs) at reactor neutrino experiments, and in particular, the mimicking effects on θ 13 . We present generic formulas for oscillation probabilities including NSIs from sources and detectors. Instructive mappings between the fundamental leptonic mixing parameters and the effective leptonic mixing parameters are established. In addition, NSI corrections to the mixing angles θ 13 and θ 12 are discussed in detailed. Finally, we show that, even for a vanishing θ 13 , an oscillation phenomenon may still be observed in future short baseline reactor neutrino experiments, such as Double Chooz and Daya Bay, due to the existences of NSIs

  16. Chooz nuclear facilities. 2009 annual report

    International Nuclear Information System (INIS)

    2010-01-01

    This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Chooz, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

  17. Flash-ADCs test, optimization of the detector design and development of a new concept of spatial reconstruction in the double chooz neutrino oscillation experiment

    International Nuclear Information System (INIS)

    Akiri, T.

    2010-09-01

    Double Chooz (DC) is a reactor neutrino oscillation experiment whose purpose is the measurement of the last unknown mixing angle θ 13 . It inherits from the past Chooz experiment which was limited by the statistical and systematic errors at the same extent of about 2.8%. To lower the statistical error, the DC detector target mass has been increased and a longer exposure is foreseen while the lowering of the systematic error is ensured by the use of two identical detectors. One will be located in the vicinity of the reactor cores to monitor the flux and spectrum of the ν-bar e emitted whereas the other one will be located where the effect of the oscillation is expected to be maximal. They are respectively so-called 'near' and 'far' detectors. The expected errors are 0.5% (stat.) and 0.6% (syst.) for a measurement down to sin 2 (2*θ 13 ) = 0.05 (θ 13 6.5 degrees) at three standard deviations after three years of data taking. The far detector is expected for November 2010 while the near detector will be operational in mid-2012. This thesis presents first a hardware work consisting in testing the Flash-ADCs that are the core of the main acquisition system of the experiment. Subsequently, it presents analyses performed on Monte Carlo simulations towards the optimization of the detector design. This work was composed of analyses to choose some detector components with the appropriate natural radioactivity contamination, analyses for the best achievable energy resolution and the most stable and robust way of triggering. The work on the optimization of the detector together with the acquired knowledge on the Flash-ADCs led us to envisage the possibility of a new spatial reconstruction based on the time of flight. All these contributions to the experiment are described in details throughout this manuscript. (author)

  18. 3D - Acquisition systems - test in Chooz B nuclear plant

    International Nuclear Information System (INIS)

    Brillault, B.; Thibault, G.

    1992-06-01

    EDF needs 3D-acquisition systems to get the precise geometry of critical nuclear spaces in order to prepare computer simulations of operations in these areas. The simulations must lead to an increase of the efficiency of the operation. The acquisition of the 3-D geometry can be done using 3D-acquisition systems. To answer the needs of the Construction Division, four different systems are compared by the Research Division in Chooz B nuclear plant in order to determine the right solution for each 3D-acquisition problem

  19. In-service - pressure test of the primary circuit of the Chooz nuclear power plant

    International Nuclear Information System (INIS)

    Barthelemy, F.L.; Lespiaucq, P.G.

    1977-01-01

    A brief summary is given of the regulations governing inspection pratices of operating nuclear power plants, in France. As an example, such an inspection performed in 1976 on the Westinghouse 320 MWe PWR built in Chooz (Ardennes) is described. Emphasis is put on the administrative organization, the technical solutions, the specific problems and the difficulties encountered. (author)

  20. Dosimetry and fluence calculations on french PWR vessels comparisons between experiments and calculations

    International Nuclear Information System (INIS)

    Nimal, J.C.; Bourdet, L.; Guilleret, J.C.; Hedin, F.

    1988-01-01

    Fluence and damage calculations on PWR pressure vessels and irradiation test specimens are presented for two types of reactor: the franco-belgian (reactor CHOOZ) and the french reactors (CPY program). Comparisons with measurements are given for activation foils and fission detectors; most of them are about irradiation test specimen locations; comparisons are made for the Chooz plant on vessel stainless steel samplings and in the reactor pit

  1. Study of the Light Emission Process from the Double Chooz Photomultipliers

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, E.; Cerrada, M.; Crespo, J. I.; Gil-Botella, I.; Jimenez, S.; Lopez, M.; Novella, P.; Palomares, C.; Santorelli, R.; Verdugo, A.

    2012-09-13

    In this document we present a study of the light emitted by the base of a Hamamatsu R7081MOD-ASSY photomultiplier (PMT) of the same type used in the Double Chooz experiment. Several characteristic features of the light signal have been found in terms of amplitude, length and pulse shape. Additional investigations on the properties of the epoxy used to cover the photomultiplier base have been carried out. A possible explanation of the light emission process is discussed at the end of the study. (Author) 1 ref.

  2. A unified analysis of the reactor neutrino program towards the measurement of the θ13 mixing angle

    International Nuclear Information System (INIS)

    Mention, G.; Motta, D.; Lasserre, Th.

    2007-04-01

    We present in this article a detailed quantitative discussion of the measurement of the leptonic mixing angle θ 13 through currently scheduled reactor neutrino oscillation experiments. We thus focus on Double Chooz (Phase I and II), Daya Bay (Phase I and II) and RENO experiments. We perform a unified analysis, including systematics, backgrounds and accurate experimental setup in each case. Each identified systematic error and background impact has been assessed on experimental setups following published data when available and extrapolating from Double Chooz acquired knowledge otherwise. After reviewing the experiments, we present a new analysis of their sensitivities to sin 2 (2θ 13 ) and study the impact of the different systematics based on the pulls approach. Through this generic statistical analysis we discuss the advantages and drawbacks of each experimental setup. (authors)

  3. A unified analysis of the reactor neutrino program towards the measurement of the {theta}{sub 13} mixing angle

    Energy Technology Data Exchange (ETDEWEB)

    Mention, G [DAPNIA/SPP, CEA Saclay, 91191 Gif sur Yvette (France)

    2008-05-15

    We presented a detailed quantitative discussion of the measurement of the leptonic mixing angle {theta}{sub 13} through currently scheduled reactor neutrino oscillation experiments. We focussed on Double Chooz (Phase I and II), Daya Bay (Phase I and II) and RENO experiments. We performed a unified analysis, including systematics, backgrounds and accurate experimental setup in each case. Each identified systematical uncertainty and background impact has been assessed on experimental setups following published data when available and extrapolating from Double Chooz acquired knowledge otherwise. We sum up, here, a new common analysis of their sensitivities to sin{sup 2}(2{theta}{sub 13}) and study the impact of the different systematics based on the pulls approach. Through this generic statistical analysis we discuss the advantages and drawbacks of each experimental setup.

  4. Light Production in the Double Chooz Photomultiplier Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, E.; Cerrada, M.; Crespo, J. I.; Gil-Botella, I.; Jimenez, S.; Lopez, M.; Novella, P.; Palomares, C.; Santorelli, R.; Verdugo, A.

    2012-09-13

    In this document we present a study of the phenomenon of light emission (called glowing) in the bases of the Hamamatsu R7081MOD-ASSY photomultiplier tubes (PMTs) used in the Double Chooz experiment. The tests have been carried out at the CIEMAT laboratories over a photomultiplier tube of the same model. We have studied the phenomenon making first a characterization of it, and then focusing on the dependence of the rate and the amount of emitted light versus voltage and temperature. In addition, we have looked for the possible existence of an ultraviolet component in the light which would be harmful for the experiment because it could be able to excite the scintillator liquid. Finally, we propose and test a method to reduce the light emission using a cover on the base of the photomultiplier tube.. (Author)

  5. Combined potential of future long-baseline and reactor experiments

    International Nuclear Information System (INIS)

    Huber, P.; Lindner, M.; Rolinec, M.; Schwetz, T.; Winter, W.

    2005-01-01

    We investigate the determination of neutrino oscillation parameters by experiments within the next ten years. The potential of conventional beam experiments (MINOS, ICARUS, OPERA), superbeam experiments (T2K, NOνA), and reactor experiments (D-CHOOZ) to improve the precision on the 'atmospheric' parameters Δm 31 2 , θ 23 , as well as the sensitivity to θ 13 are discussed. Further, we comment on the possibility to determine the leptonic CP-phase and the neutrino mass hierarchy if θ 13 turns out to be large

  6. A unified analysis of the reactor neutrino program towards the measurement of the {theta}{sub 13} mixing angle

    Energy Technology Data Exchange (ETDEWEB)

    Mention, G.; Motta, D. [DAPNIA/SPP, CEA Saclay, 91191 Gif sur Yvette (France); Lasserre, Th. [DAPNIA/SPP, CEA Saclay, 91191 Gif sur Yvette (France); Laboratoire Astroparticule et Cosmologie (APC), Paris (France)

    2007-04-15

    We present in this article a detailed quantitative discussion of the measurement of the leptonic mixing angle {theta}{sub 13} through currently scheduled reactor neutrino oscillation experiments. We thus focus on Double Chooz (Phase I and II), Daya Bay (Phase I and II) and RENO experiments. We perform a unified analysis, including systematics, backgrounds and accurate experimental setup in each case. Each identified systematic error and background impact has been assessed on experimental setups following published data when available and extrapolating from Double Chooz acquired knowledge otherwise. After reviewing the experiments, we present a new analysis of their sensitivities to sin{sup 2}(2{theta}{sub 13}) and study the impact of the different systematics based on the pulls approach. Through this generic statistical analysis we discuss the advantages and drawbacks of each experimental setup. (authors)

  7. Safety margins of PWR irradiated vessels - The Chooz A issue

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, F; Barthelet, B [Electricite de France (EDF), 75 - Paris (France); Guilleret, J C

    1988-12-31

    In 1986, some irradiated specimen of CHOOZ A (SENA) vessel showed a significant excess of {delta} RTNDT to former previsions. The lack of data on one of the two irradiated shells, and discrepancies between dosimeters results and available previous fluence calculations whose accuracy was questionable, cause the safety authorities to require an important complementary work program before putting again the plant on the grid after 1987 fuel reloading. These works are presented and discussed. They lead to a state that a conservative to day value of the vessel RTNDT is 64 degrees Celsius and that there is no underclad defect in the vessel wall and welds. Then the plant was allowed to restart with certitude that vessel irradiation will not impair its lifetime. (author). 4 refs.

  8. Fast reactor development program in France in 1997

    International Nuclear Information System (INIS)

    Langrand, J.C.; Hubert, G.; Marmonier, P.; Del Negro, R.; Saint-Arroman, G.; Coulon, P.; Mesnage, B.; Pillon, S.; Lefevre, J.C.

    1998-01-01

    The operating performance of the French nuclear plants in 1997 were again satisfactory, confirming the good 1996 results; 78% of the national electricity production was of nuclear origin (376/481 TWh); the nuclear safety indicators were even better with a slight decrease in the number of reported incidents; the availability factor was higher than 82,5%; the kWh production cost continue to decrease by about 1,7% per year; thanks to an ALARA approach, individual and collective doses are still decreasing; the liquid and gaseous releases are also still decreasing and the volume of nuclear wastes is equivalent to last year. The major event of 1997 was the confirmation for the start up of the most advanced nuclear plants, N4 type, with Chooz B1 reaching full power on May 9, 1997 and Chooz B2 following soon with first criticality on March 10 and full power on September 18. As a confirmation and enjoying the experience of Chooz B the first unit of Civaux was synchronised to the grid end of 1997 after fuel loading in Sept 1997. The status of fast reactors was deeply modified, with the governmental decision to abandon Superphenix, which led to decide to close also EFR studies in 1997, and the perspective of operating Phenix till 2004, mainly for studies on radioactive wastes incineration. Three meetings were held during the second half of the year by the > (GP - Permanent Committee) on Phenix; after reviewing all the safety aspects, they concluded to the possibility to resume operation according to the general following scheme: one cycle in 1998, about one year for works, then 6 cycles between 2000 and 2004. Of course, the main event concerning Superphenix is the decision of abandon announced in June 1997. The decree allowing operation had been cancelled in February. The works foreseen in the frame of the planned > shutdown had been conducted satisfactorily, during the first-half of the year. At the beginning of 1998, it was stated that the reactor would not operate again

  9. French experience in operating pressurized water reactor power stations. Ten years' operation of the Ardennes power station

    International Nuclear Information System (INIS)

    Teste du Bailler, A.; Vedrinne, J.F.

    1978-01-01

    In the paper the experience gained over ten years' operation of the Ardennes (Chooz) nuclear power station is summarized from the point of view of monitoring and control equipment. The reactor was the first pressurized water reactor to be installed in France; it is operated jointly by France and Belgium. The equipment, which in many cases consists of prototypes, was developed for industrial use and with the experience that has now been gained it is possible to evaluate its qualities and defects, the constraints which it imposes and the action that has to be taken in the future. (author)

  10. Chasing {theta}{sub 13} with new reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Lasserre, Th. [DSM/DAPNIA/SPP, CEA/Saclay, 91191 Gif-sur-Yvette (France)

    2005-12-15

    It is now widely accepted that a new middle baseline disappearance reactor neutrino experiment with multiple detectors could provide a clean measurement of the {theta}{sub 13} mixing angle, free from any parameter degeneracies and correlations induced by matter effect and the unknown leptonic Dirac CP phase. The current best constraint on the third mixing angle comes from the Chooz reactor neutrino experiment sin{sup 2}(2{theta}{sub 13})<0.2 (90 % C.L., {delta}m{sub atm}{sup 2}=2.010{sup -3} eV{sup 2}). Several projects of experiment, with different timescales, have been proposed over the last two years all around the world. Their sensitivities range from sin{sup 2}(2{theta}{sub 13})<0.01 to 0.03, having thus an excellent discovery potential of the {nu}{sub e} fraction of {nu}{sub 3}.

  11. Investigating the spectral anomaly with different reactor antineutrino experiments

    Directory of Open Access Journals (Sweden)

    C. Buck

    2017-02-01

    Full Text Available The spectral shape of reactor antineutrinos measured in recent experiments shows anomalies in comparison to neutrino reference spectra. New precision measurements of the reactor neutrino spectra as well as more complete input in nuclear data bases are needed to resolve the observed discrepancies between models and experimental results. This article proposes the combination of experiments at reactors which are highly enriched in U235 with commercial reactors with typically lower enrichment to gain new insights into the origin of the anomalous neutrino spectrum. The presented method clarifies, if the spectral anomaly is either solely or not at all related to the predicted U235 spectrum. Considering the current improvements of the energy scale uncertainty of present-day experiments, a significance of three sigma and above can be reached. As an example, we discuss the option of a direct comparison of the measured shape in the currently running Double Chooz near detector and the upcoming Stereo experiment. A quantitative feasibility study emphasizes that a precise understanding of the energy scale systematics is a crucial prerequisite in recent and next generation experiments investigating the spectral anomaly.

  12. Characterization of large-area photomultipliers under low magnetic fields: Design and performance of the magnetic shielding for the Double Chooz neutrino experiment

    International Nuclear Information System (INIS)

    Calvo, E.; Cerrada, M.; Fernandez-Bedoya, C.; Gil-Botella, I.; Palomares, C.; Rodriguez, I.; Toral, F.; Verdugo, A.

    2010-01-01

    A precise quantitative measurement of the effect of low magnetic fields in Hamamatsu R7081 photomultipliers has been performed. These large-area photomultipliers will be used in the Double Chooz neutrino experiment. A magnetic shielding has been developed for these photomultipliers. Its design and performance is also reported in this paper.

  13. First test of Lorentz violation with a reactor-based antineutrino experiment

    International Nuclear Information System (INIS)

    Abe, Y.; Ishitsuka, M.; Konno, T.; Kuze, M.; Aberle, C.; Buck, C.; Hartmann, F.X.; Haser, J.; Kaether, F.; Lindner, M.; Reinhold, B.; Schwetz, T.; Wagner, S.; Watanabe, H.; Anjos, J.C. dos; Gama, R.; Lima, H.P.-Jr.; Pepe, I.M.; Bergevin, M.; Felde, J.; Maesano, C.N.; Bernstein, A.; Bowden, N.S.; Dazeley, S.; Erickson, A.; Keefer, G.; Bezerra, T.J.C.; Furuta, H.; Suekane, F.; Bezrukhov, L.; Lubsandorzhiev, B.K.; Yanovitch, E.; Blucher, E.; Conover, E.; Crum, K.; Strait, M.; Worcester, M.; Busenitz, J.; Goon, J.TM.; Habib, S.; Ostrovskiy, I.; Reichenbacher, J.; Stancu, I.; Sun, Y.; Cabrera, A.; Franco, D.; Kryn, D.; Obolensky, M.; Roncin, R.; Tonazzo, A.; Caden, E.; Damon, E.; Lane, C.E.; Maricic, J.; Miletic, T.; Milincic, R.; Perasso, S.; Smith, E.; Camilleri, L.; Carr, R.; Franke, A.J.; Shaevitz, M.H.; Toups, M.; Cerrada, M.; Crespo-Anadon, J.I.; Gil-Botella, I.; Lopez-Castano, J.M.; Novella, P.; Palomares, C.; Santorelli, R.; Chang, P.J.; Horton-Smith, G.A.; McKee, D.; Shrestha, D.; Chimenti, P.; Classen, T.; Collin, A.P.; Cucoanes, A.; Durand, V.; Fechner, M.; Fischer, V.; Hayakawa, T.; Lasserre, T.; Letourneau, A.; Lhuillier, D.; Mention, G.; Mueller, Th.A.; Perrin, P.; Sida, J.L.; Sinev, V.; Veyssiere, C.

    2012-01-01

    We present a search for Lorentz violation with 8249 candidate electron antineutrino events taken by the Double Chooz experiment in 227.9 live days of running. This analysis, featuring a search for a sidereal time dependence of the events, is the first test of Lorentz invariance using a reactor-based antineutrino source. No sidereal variation is present in the data and the disappearance results are consistent with sidereal time independent oscillations. Under the Standard-Model Extension, we set the first limits on 14 Lorentz violating coefficients associated with transitions between electron and tau flavor, and set two competitive limits associated with transitions between electron and muon flavor. (authors)

  14. Cosmic-muon characterization and annual modulation measurement with Double Chooz detectors

    Science.gov (United States)

    Abrahão, T.; Almazan, H.; dos Anjos, J. C.; Appel, S.; Baussan, E.; Bekman, I.; Bezerra, T. J. C.; Bezrukov, L.; Blucher, E.; Brugière, T.; Buck, C.; Busenitz, J.; Cabrera, A.; Camilleri, L.; Carr, R.; Cerrada, M.; Chauveau, E.; Chimenti, P.; Corpace, O.; Crespo-Anadón, J. I.; Dawson, J. V.; Dhooghe, J.; Djurcic, Z.; Dracos, M.; Etenko, A.; Fallot, M.; Franco, D.; Franke, M.; Furuta, H.; Gil-Botella, I.; Giot, L.; Givaudan, A.; Gögger-Neff, M.; Gómez, H.; Gonzalez, L. F. G.; Goodman, M.; Hara, T.; Haser, J.; Hellwig, D.; Hourlier, A.; Ishitsuka, M.; Jochum, J.; Jollet, C.; Kale, K.; Kampmann, P.; Kaneda, M.; Kaplan, D. M.; Kawasaki, T.; Kemp, E.; de Kerret, H.; Kryn, D.; Kuze, M.; Lachenmaier, T.; Lane, C.; Laserre, T.; Lastoria, C.; Lhuillier, D.; Lima, H.; Lindner, M.; López-Castaño, J. M.; LoSecco, J. M.; Lubsandorzhiev, B.; Maeda, J.; Mariani, C.; Maricic, J.; Matsubara, T.; Mention, G.; Meregaglia, A.; Miletic, T.; Minotti, A.; Nagasaka, Y.; Navas-Nicolás, D.; Novella, P.; Oberauer, L.; Obolensky, M.; Onillon, A.; Oralbaev, A.; Palomares, C.; Pepe, I.; Pronost, G.; Reinhold, B.; Rybolt, B.; Sakamoto, Y.; Santorelli, R.; Schönert, S.; Schoppmann, S.; Sharankova, R.; Sibille, V.; Sinev, V.; Skorokhvatov, M.; Soiron, M.; Soldin, P.; Stahl, A.; Stancu, I.; Stokes, L. F. F.; Strait, M.; Suekane, F.; Sukhotin, S.; Sumiyoshi, T.; Sun, Y.; Svoboda, B.; Tonazzo, A.; Veyssiere, C.; Vivier, M.; Wagner, S.; Wiebusch, C.; Wurm, M.; Yang, G.; Yermia, F.; Zimmer, V.

    2017-02-01

    A study on cosmic muons has been performed for the two identical near and far neutrino detectors of the Double Chooz experiment, placed at ~120 and ~300 m.w.e. underground respectively, including the corresponding simulations using the MUSIC simulation package. This characterization has allowed us to measure the muon flux reaching both detectors to be (3.64 ± 0.04) × 10-4 cm-2s-1 for the near detector and (7.00 ± 0.05) × 10-5 cm-2s-1 for the far one. The seasonal modulation of the signal has also been studied observing a positive correlation with the atmospheric temperature, leading to an effective temperature coefficient of αT = 0.212 ± 0.024 and 0.355 ± 0.019 for the near and far detectors respectively. These measurements, in good agreement with expectations based on theoretical models, represent one of the first measurements of this coefficient in shallow depth installations.

  15. Cosmic-muon characterization and annual modulation measurement with Double Chooz detectors

    International Nuclear Information System (INIS)

    Abrahão, T.; Anjos, J.C. dos; Almazan, H.; Buck, C.; Appel, S.; Baussan, E.; Brugière, T.; Bekman, I.; Bezerra, T.J.C.; Bezrukov, L.; Blucher, E.; Busenitz, J.; Cabrera, A.; Camilleri, L.; Carr, R.; Cerrada, M.; Chauveau, E.; Chimenti, P.

    2017-01-01

    A study on cosmic muons has been performed for the two identical near and far neutrino detectors of the Double Chooz experiment, placed at ∼120 and ∼300 m.w.e. underground respectively, including the corresponding simulations using the MUSIC simulation package. This characterization has allowed us to measure the muon flux reaching both detectors to be (3.64 ± 0.04) × 10 −4 cm −2 s −1 for the near detector and (7.00 ± 0.05) × 10 −5 cm −2 s −1 for the far one. The seasonal modulation of the signal has also been studied observing a positive correlation with the atmospheric temperature, leading to an effective temperature coefficient of α T = 0.212 ± 0.024 and 0.355 ± 0.019 for the near and far detectors respectively. These measurements, in good agreement with expectations based on theoretical models, represent one of the first measurements of this coefficient in shallow depth installations.

  16. Investigation of the Chooz-A nuclear power plant bolts

    International Nuclear Information System (INIS)

    Monnet, I.; Decroix, G.M.; Dubuisson, P.; Reuchet, J.; Morlent, O.

    2002-01-01

    In Pressurised Water Reactor, some baffle-former bolts in austenitic stainless steel, are submitted to an important Intergranular cracking. This cracking may be attributed to the irradiation hardening during in pile service. As part of its concern of safety related to the ageing of the plant and, in particular, to the behaviour of the internals, IRSN wished to take part in the expertise Program on CHOOZ A power nuclear plant, within the framework of a convention with EDF, which provided the baffle-bolts. Examinations of these bolts, after 140 000 of in pile service, were carried out by the laboratories of the CEA. Hardness profiles were carried out on four bolts, which were irradiated at different doses, ranging from 0 to 22 dpa. The bolt of the core barrel, considered as unirradiated, shows a constant value of hardness, equal to the value of unirradiated material, all along the bolt and a slight increase in the head of the screw. The axial profile of hardness carried out on an irradiated bolt shows that there is a gradient of hardness between the most irradiated part (400 Hv) and the least irradiated part (270 Hv). The hardness of this bolt starts to evolve at 2.5 dpa and the maximum of hardening is reached for the most irradiated part (3.6 dpa). The hardness profile on the most irradiated bolt (between 10 and 22 dpa) indicate that hardness is homogeneous all along the bolt, 400 Hv. It confirms the existence of a threshold dose beyond which hardness does not vary any more, this threshold is estimated to be between 3.6 dpa and 10 dl Microstructural examinations of these bolts led us to conclude that hardening is correlated to dislocation microstructure. Indeed, the examination of the bolt of the core barrel confirms that this bolt is hardly unirradiated since the initial network of dislocations was not modified and that only some rare dislocation loops were observed. This is in agreement with hardness results showing no hardening of this bolt. On the two more

  17. Studying the muon background component in the Double Chooz experiment

    Energy Technology Data Exchange (ETDEWEB)

    Dietrich, Dennis

    2013-03-28

    The reactor anti-neutrino experiment Double Chooz (DC) will measure the third neutrino mixing angle θ{sub 13} with very high precision. This mixing angle is connected to fundamental questions in particle physics beyond the current Standard Model. In DC neutrinos are detected via the Inverse Beta Decay reaction, which provides a clean signal distinguishable from most backgrounds. However, as neutrino interactions in the detector are very rare and an interfering muon background is present, a proper understanding and reduction of this background is mandatory. This is crucial because muons create fast neutrons and βn-emitters which lead to background capable of mimicking the neutrino interaction in the detector. This thesis covers different analysis topics related to the cosmic ray muon background at the DC far site. The thesis covers the identification of muons, the applied rejection technique and the determination of the muon rate at DC far site. Utilizing the muon rejection cuts of the neutrino analysis a muon rate of 13 s{sup -1} in the Inner Detector (ID) and of 46 s{sup -1} in the Inner Muon Veto (IV) was found. The efficiency of the IV to identify and reject cosmic ray muons was measured and a value greater than 99.97% has been found. The stability of the determined muon rates was examined and a seasonal modulation was found, compatible with a variation of the temperature profile of the atmosphere over the year. The parameter describing the strength between the relationship of temperature and muon rate change, the effective temperature coefficient was obtained: αT=0.39±0.01(stat.)±0.02(syst.). This gave the opportunity to measure the atmospheric kaon to pion ratio with the DC far detector which was found to be r(K/π)=0.14±0.06. Additional variations of muon rate with surface pressure were found and the barometric coefficient describing this effect was measured as βp=-0.59±0.20(stat.)±0.10(syst.) permille /mbar. Another central theme of this work was

  18. Studying the muon background component in the Double Chooz experiment

    International Nuclear Information System (INIS)

    Dietrich, Dennis

    2013-01-01

    The reactor anti-neutrino experiment Double Chooz (DC) will measure the third neutrino mixing angle θ 13 with very high precision. This mixing angle is connected to fundamental questions in particle physics beyond the current Standard Model. In DC neutrinos are detected via the Inverse Beta Decay reaction, which provides a clean signal distinguishable from most backgrounds. However, as neutrino interactions in the detector are very rare and an interfering muon background is present, a proper understanding and reduction of this background is mandatory. This is crucial because muons create fast neutrons and βn-emitters which lead to background capable of mimicking the neutrino interaction in the detector. This thesis covers different analysis topics related to the cosmic ray muon background at the DC far site. The thesis covers the identification of muons, the applied rejection technique and the determination of the muon rate at DC far site. Utilizing the muon rejection cuts of the neutrino analysis a muon rate of 13 s -1 in the Inner Detector (ID) and of 46 s -1 in the Inner Muon Veto (IV) was found. The efficiency of the IV to identify and reject cosmic ray muons was measured and a value greater than 99.97% has been found. The stability of the determined muon rates was examined and a seasonal modulation was found, compatible with a variation of the temperature profile of the atmosphere over the year. The parameter describing the strength between the relationship of temperature and muon rate change, the effective temperature coefficient was obtained: αT=0.39±0.01(stat.)±0.02(syst.). This gave the opportunity to measure the atmospheric kaon to pion ratio with the DC far detector which was found to be r(K/π)=0.14±0.06. Additional variations of muon rate with surface pressure were found and the barometric coefficient describing this effect was measured as βp=-0.59±0.20(stat.)±0.10(syst.) permille /mbar. Another central theme of this work was the extrapolation

  19. The treatment of low level effluents by flocculation and settling at the Chooz nuclear power plant

    International Nuclear Information System (INIS)

    Petteau, J.L.; Roofthooft, R.

    1989-01-01

    At the Chooz plant, radioactive effluents were formerly treated by evaporation, but because throughput was low, another method was studied. After laboratory tests, a 500 L/h flocculation and settling pilot plant was constructed, followed later by a 5 m 3 /h installation. The main isotopes eliminated are caesium-134 and caesium-137. Flocculation with copper ferrocyanide reduces the total activity to less than 500 Bq/L. The installation described in the paper was commissioned in 1984 and has been in industrial operation since 1985, processing all types of effluent. The evaporator can be set aside for boric acid recovery. (author). 3 figs, 1 tab

  20. Cosmic-muon characterization and annual modulation measurement with Double Chooz detectors

    Energy Technology Data Exchange (ETDEWEB)

    Abrahão, T.; Anjos, J.C. dos [Centro Brasileiro de Pesquisas Físicas, Rio de Janeiro, RJ, 22290-180 (Brazil); Almazan, H.; Buck, C. [Max-Planck-Institut für Kernphysik, 69117 Heidelberg (Germany); Appel, S. [Physik Department, Technische Universität München, 85748 Garching (Germany); Baussan, E.; Brugière, T. [IPHC, Université de Strasbourg, CNRS/IN2P3, 67037 Strasbourg (France); Bekman, I. [III. Physikalisches Institut, RWTH Aachen University, 52056 Aachen (Germany); Bezerra, T.J.C. [SUBATECH, CNRS/IN2P3, Université de Nantes, Ecole des Mines de Nantes, 44307 Nantes (France); Bezrukov, L. [Institute of Nuclear Research of the Russian Academy of Sciences, Moscow 117312 (Russian Federation); Blucher, E. [The Enrico Fermi Institute, The University of Chicago, Chicago, Illinois 60637 (United States); Busenitz, J. [Department of Physics and Astronomy, University of Alabama, Tuscaloosa, Alabama 35487 (United States); Cabrera, A. [AstroParticule et Cosmologie, Université Paris Diderot, CNRS/IN2P3, CEA/IRFU, Observatoire de Paris, Sorbonne Paris Cité, 75205 Paris Cedex 13 (France); Camilleri, L.; Carr, R. [Columbia University, New York, New York 10027 (United States); Cerrada, M. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, CIEMAT, 28040, Madrid (Spain); Chauveau, E. [Research Center for Neutrino Science, Tohoku University, Sendai 980-8578 (Japan); Chimenti, P., E-mail: hgomez@apc.univ-paris7.fr [Universidade Federal do ABC, UFABC, Santo André, SP, 09210-580 (Brazil); and others

    2017-02-01

    A study on cosmic muons has been performed for the two identical near and far neutrino detectors of the Double Chooz experiment, placed at ∼120 and ∼300 m.w.e. underground respectively, including the corresponding simulations using the MUSIC simulation package. This characterization has allowed us to measure the muon flux reaching both detectors to be (3.64 ± 0.04) × 10{sup −4} cm{sup −2}s{sup −1} for the near detector and (7.00 ± 0.05) × 10{sup −5} cm{sup −2}s{sup −1} for the far one. The seasonal modulation of the signal has also been studied observing a positive correlation with the atmospheric temperature, leading to an effective temperature coefficient of α {sub T} = 0.212 ± 0.024 and 0.355 ± 0.019 for the near and far detectors respectively. These measurements, in good agreement with expectations based on theoretical models, represent one of the first measurements of this coefficient in shallow depth installations.

  1. IAEA Concludes Safety Review at Chooz Nuclear Power Plant in France

    International Nuclear Information System (INIS)

    2013-01-01

    Full text: An IAEA-led international team of nuclear safety experts noted good practices and made recommendations to reinforce safety measures during a review of operational safety at France's Chooz Nuclear Power Plant (NPP) that concluded today. The Operational Safety Review Team (OSART) was assembled at the French Government's request. The in-depth review, which began 17 June, focused on aspects essential to the safe operation of the NPP. The team comprised experts from Switzerland, Belgium, Germany, China, India, United Kingdom, Czech Republic, Canada, Hungary and the IAEA. The review covered the areas of management, organization and administration; training and qualification of personnel; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency planning and preparedness; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards. The OSART team identified good plant practices that will be shared with the rest of the nuclear industry for consideration. Examples include: The plant has a professional development programme as part of a joint employment effort shared by the plant and its contractors. This enables trainees to develop professional capability, understand practices and gain experience from other nuclear power plants in terms of work planning and coordination; The plant has built a strong relationship between the on-shift response team of the plant and the local fire brigade to improve firefighting and rescue operations; Self-assessment groups discuss and resolve specific issues within operations, empowering operations personnel to take ownership of improvement programmes; and The plant has improved warnings at entrances to all o range zones , areas of elevated dose rates to which only authorized staff have access. The team identified a number of improvements to operational safety at Chooz NPP. Examples include: The plant should review its process for the

  2. Étude des antineutrinos de réacteurs : mesure de l'angle de mélange leptonique θ₁₃ et recherche d'éventuels neutrinos stériles

    OpenAIRE

    Collin, Antoine

    2014-01-01

    The Double Chooz experiment aims to measure the θ₁₃ mixing angle through the disappearance—induced by the oscillation phenomenon—of anti-neutrinos produced by the Chooz nuclear reactors. In order to reduce systematic uncertainties, the experiment relies on the relative comparison of detected signals in two identical liquid scintillator detectors. The near one, giving the normalization of the emitted flux, is currently being built and will be delivered in spring 2014. The far detector, sensiti...

  3. Decree no. 96-927 from October 16, 1996 giving permission to Electricite de France to operate the Ardennes nuclear power plant including the basic nuclear installations no. 1 (reactor and auxiliary circuits), no. 2 (radioactive effluents processing plant) and no. 3 (fuel storage building), located on the territory of Chooz town (Ardennes)

    International Nuclear Information System (INIS)

    Borotra, F.

    1996-01-01

    This decree from the French ministry of industry and postal services gives to Electricite de France (EDF) the official permission to operate the Chooz nuclear power plant previously operated by the French-Belgium nuclear energy Society of the Ardennes. The operation will follow the conditions previously imposed to this Society. (J.S.)

  4. Some activities in France related to life management of nuclear power plants

    International Nuclear Information System (INIS)

    Petrequin, P.

    1998-01-01

    Significant events concerned with life management of NPPs in France were: replacement of steam generators of TRICASTIN 1 NPP, defect detected in the emergency cooling system of DAMPIERRE 1 reactor and expertise of decommissioned CHOOZ A reactor. Programs in progress described in this report are: irradiation of materials for internal components and fracture toughness studies of irradiated materials

  5. Muon background studies for shallow depth Double - Chooz near detector

    Energy Technology Data Exchange (ETDEWEB)

    Gómez, H. [Laboratoire Astroparticule et Cosmologie (APC) - Université Paris 7. Paris (France)

    2015-08-17

    Muon events are one of the main concerns regarding background in neutrino experiments. The placement of experimental set-ups in deep underground facilities reduce considerably their impact on the research of the expected signals. But in the cases where the detector is installed on surface or at shallow depth, muon flux remains high, being necessary their precise identification for further rejection. Total flux, mean energy or angular distributions are some of the parameters that can help to characterize the muons. Empirically, the muon rate can be measured in an experiment by a number of methods. Nevertheless, the capability to determine the muons angular distribution strongly depends on the detector features, while the measurement of the muon energy is quite difficult. Also considering that on-site measurements can not be extrapolated to other sites due to the difference on the overburden and its profile, it is necessary to find an adequate solution to perform the muon characterization. The method described in this work to obtain the main features of the muons reaching the experimental set-up, is based on the muon transport simulation by the MUSIC software, combined with a dedicated sampling algorithm for shallow depth installations based on a modified Gaisser parametrization. This method provides all the required information about the muons for any shallow depth installation if the corresponding overburden profile is implemented. In this work, the method has been applied for the recently commissioned Double - Chooz near detector, which will allow the cross-check between the simulation and the experimental data, as it has been done for the far detector.

  6. Muon background studies for shallow depth Double - Chooz near detector

    International Nuclear Information System (INIS)

    Gómez, H.

    2015-01-01

    Muon events are one of the main concerns regarding background in neutrino experiments. The placement of experimental set-ups in deep underground facilities reduce considerably their impact on the research of the expected signals. But in the cases where the detector is installed on surface or at shallow depth, muon flux remains high, being necessary their precise identification for further rejection. Total flux, mean energy or angular distributions are some of the parameters that can help to characterize the muons. Empirically, the muon rate can be measured in an experiment by a number of methods. Nevertheless, the capability to determine the muons angular distribution strongly depends on the detector features, while the measurement of the muon energy is quite difficult. Also considering that on-site measurements can not be extrapolated to other sites due to the difference on the overburden and its profile, it is necessary to find an adequate solution to perform the muon characterization. The method described in this work to obtain the main features of the muons reaching the experimental set-up, is based on the muon transport simulation by the MUSIC software, combined with a dedicated sampling algorithm for shallow depth installations based on a modified Gaisser parametrization. This method provides all the required information about the muons for any shallow depth installation if the corresponding overburden profile is implemented. In this work, the method has been applied for the recently commissioned Double - Chooz near detector, which will allow the cross-check between the simulation and the experimental data, as it has been done for the far detector

  7. Studies of neutrino properties at nuclear reactors. Present status and future

    CERN Document Server

    Mikaehlyan, L A

    2002-01-01

    The state and prospects of the experiments at nuclear reactors on the search for the neutrino mass, mixing and magnetic moments, identification whereof would prove the existence of events behind the standard model limits, are considered. The CHOOZ experiment established with complete determination, that the nu sub e -> nu sub x channel is not predominant in the atmospheric neutrino oscillations. The KamLAND may become the first among the experiments with the neutrino earth sources, wherein the oscillation effect will be determined; the contributions of the m sub 1 and m sub 2 masses to the electron neutrino is established and solution of the solar neutrino problem is found. The Kr2Det experiment with its high sensitivity to the small mixing angles may identify the contribution of the m sub 3 mass to the nu sub e or establish a new more exact limit on its value. These studies rest upon the unprecedented improvement of the reverse beta-decay registration methods

  8. Neutrino oscillations in Gallium and reactor experiments and cosmological effects of a light sterile neutrino

    International Nuclear Information System (INIS)

    Acero-Ortega, Mario Andres

    2009-01-01

    Neutrino oscillations is a very well studied phenomenon and the observations from Solar, very-long-baseline Reactor, Atmospheric and Accelerator neutrino oscillation experiments give very robust evidence of three-neutrino mixing. On the other hand, some experimental data have shown anomalies that could be interpreted as indication of exotic neutrino physics beyond three-neutrino mixing. Furthermore, from a cosmological point of view, the possibility of extra light species contributing as a subdominant hot (or warm) component of the Universe is still interesting. In the first part of this Thesis, we focused on the anomaly observed in the Gallium radioactive source experiments. These experiments were done to test the Gallium solar neutrino detectors GALLEX and SAGE, by measuring the electron neutrino flux produced by intense artificial radioactive sources placed inside the detectors. The measured number of events was smaller than the expected one. We interpreted this anomaly as a possible indication of the disappearance of electron neutrinos and, in the effective framework of two-neutrino mixing, we obtained sin 2 2θ ≥ 0.03 and Δm 2 ≥ 0.1 eV 2 . We also studied the compatibility of this result with the data of the Bugey and Chooz reactor antineutrino disappearance experiments. We found that the Bugey data present a hint of neutrino oscillations with 0.02 ≤ sin 2 2θ ≤ 0.07 and Δm 2 ≅ 1.95 eV 2 , which is compatible with the Gallium allowed region of the mixing parameters. Then, combining the data of Bugey and Chooz, the data of Gallium and Bugey, and the data of Gallium, Bugey and Chooz, we found that this hint persists, with an acceptable compatibility of the experimental data. Furthermore, we analyzed the experimental data of the I.L.L., S.R.S, and Gosgen nuclear Reactor experiments. We obtained a good fit of the I.L.L. data, showing 1 and 2σ allowed regions in the oscillation parameters space. However, the combination of I.L.L. data with the Bugey

  9. Atmospheric, Long Baseline, and Reactor Neutrino Data Constraints on θ13

    Science.gov (United States)

    Roa, J. E.; Latimer, D. C.; Ernst, D. J.

    2009-08-01

    An atmospheric neutrino oscillation tool that uses full three-neutrino oscillation probabilities and a full three-neutrino treatment of the Mikheyev-Smirnov-Wolfenstein effect, together with an analysis of the K2K, MINOS, and CHOOZ data, is used to examine the bounds on θ13. The recent, more finely binned, Super-K atmospheric data are employed. For L/Eν≳104km/GeV, we previously found significant linear in θ13 terms. This analysis finds θ13 bounded from above by the atmospheric data while bounded from below by CHOOZ. The origin of this result arises from data in the previously mentioned very long baseline region; here, matter effects conspire with terms linear in θ13 to produce asymmetric bounds on θ13. Assuming CP conservation, we find θ13=-0.07-0.11+0.18 (90% C.L.).

  10. Enhanced Constraints on theta13 from A Three-Flavor Oscillation Analysis of Reactor Antineutrinos at KamLAND

    Energy Technology Data Exchange (ETDEWEB)

    The KamLAND Collaboration; Gando, A.; Gando, Y.; Ichimura, K.; Ikeda, H.; Inoue, K.; Kibe, Y.; Kishimoto, Y.; Koga, M.; Minekawa, Y.; Mitsui, T.; Morikawa, T.; Nagai, N.; Nakajima, K.; Nakamura, K.; Narita, K.; Shimizu, I.; Shimizu, Y.; Shirai, J.; Suekane, F.; Suzuki, A.; Takahashi, H.; Takahashi, N.; Takemoto, Y.; Tamae, K.; Watanabe, H.; Xu, B. D.; Yabumoto, H.; Yoshida, H.; Yoshida, S.; Enomoto, S.; Kozlov, A.; Murayama, H.; Grant, C.; Keefer, G.; Piepke, A.; Banks, T. I.; Bloxham, T.; Detwiler, J. A.; Freedman, S. J.; Fujikawa, B. K.; Han, K.; Kadel, R.; O' Donnell, T.; Steiner, H. M.; Dwyer, D. A.; McKeown, R. D.; Zhang, C.; Berger, B. E.; Lane, C. E.; Maricic, J.; Miletic, T.; Batygov, M.; Learned, J. G.; Matsuno, S.; Sakai, M.; Horton-Smith, G. A.; Downum, K. E.; Gratta, G.; Efremenko, Y.; Perevozchikov, O.; Karwowski, H. J.; Markoff, D. M.; Tornow, W.; Heeger, K. M.; Decowski, M. P.

    2010-09-24

    We present new constraints on the neutrino oscillation parameters {Delta}m{sub 21}{sup 2}, {theta}{sub 12}, and {theta}{sub 13} from a three-flavor analysis of solar and KamLAND data. The KamLAND data set includes data acquired following a radiopurity upgrade and amounts to a total exposure of 3.49 x 10{sup 32} target-proton-year. Under the assumption of CPT invariance, a two-flavor analysis ({theta}{sub 13} = 0) of the KamLAND and solar data yields the best-fit values tan{sup 2} {theta}{sub 12} = 0.444{sub -0.030}{sup +0.036} and {Delta}m{sub 21}{sup 2} = 7.50{sub -0.20}{sup +0.19} x 10{sup -5} eV{sup 2}; a three-flavor analysis with {theta}{sub 13} as a free parameter yields the best-fit values tan{sup 2} {theta}{sub 12} = 0.452{sub -0.033}{sup +0.035}, {Delta}m{sub 21}{sup 2} = 7.50{sub -0.20}{sup +0.19} x 10{sup -5}eV{sup 2}, and sin{sup 2} {theta}{sub 13} = 0.020{sub -0.016}{sup +0.016}. This {theta}{sub 13} interval is consistent with other recent work combining the CHOOZ, atmospheric and long-baseline accelerator experiments. We also present a new global {theta}{sub 13} analysis, incorporating the CHOOZ, atmospheric and accelerator data, which indicates sin{sup 2} {theta}{sub 13} = 0.017{sub -0.009}{sup +0.010}, a nonzero value at the 93% C.L. This finding will be further tested by upcoming accelerator and reactor experiments.

  11. Atmospheric, Long Baseline, and Reactor Neutrino Data Constraints on θ13

    International Nuclear Information System (INIS)

    Roa, J. E.; Ernst, D. J.; Latimer, D. C.

    2009-01-01

    An atmospheric neutrino oscillation tool that uses full three-neutrino oscillation probabilities and a full three-neutrino treatment of the Mikheyev-Smirnov-Wolfenstein effect, together with an analysis of the K2K, MINOS, and CHOOZ data, is used to examine the bounds on θ 13 . The recent, more finely binned, Super-K atmospheric data are employed. For L/E ν > or approx. 10 4 km/GeV, we previously found significant linear in θ 13 terms. This analysis finds θ 13 bounded from above by the atmospheric data while bounded from below by CHOOZ. The origin of this result arises from data in the previously mentioned very long baseline region; here, matter effects conspire with terms linear in θ 13 to produce asymmetric bounds on θ 13 . Assuming CP conservation, we find θ 13 =-0.07 -0.11 +0.18 (90% C.L.).

  12. Atmospheric, long baseline, and reactor neutrino data constraints on theta_{13}.

    Science.gov (United States)

    Roa, J E; Latimer, D C; Ernst, D J

    2009-08-07

    An atmospheric neutrino oscillation tool that uses full three-neutrino oscillation probabilities and a full three-neutrino treatment of the Mikheyev-Smirnov-Wolfenstein effect, together with an analysis of the K2K, MINOS, and CHOOZ data, is used to examine the bounds on theta_{13}. The recent, more finely binned, Super-K atmospheric data are employed. For L/E_{nu} greater, similar 10;{4} km/GeV, we previously found significant linear in theta_{13} terms. This analysis finds theta_{13} bounded from above by the atmospheric data while bounded from below by CHOOZ. The origin of this result arises from data in the previously mentioned very long baseline region; here, matter effects conspire with terms linear in theta_{13} to produce asymmetric bounds on theta_{13}. Assuming CP conservation, we find theta_{13} = -0.07_{-0.11};{+0.18} (90% C.L.).

  13. Implementation of ALARA at the design stage of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Brissaud, A.; Ridoux, P. [Electricite de France, Villeurbanne (France)

    1995-03-01

    In the 1970s, Electricite de France (EdF) had limited knowledge and experience of pressurized water reactors (PWRs). Electricity generation by nuclear units was oriented towards gas-graphite reactors, even though EdF had a share in the PWR unit of CHOOZ A-1 (250 MWe, later upgraded to 320 MWe). Some facts about the origin of doses in that king of reactor were known to the research and development (R&D) support staff of EdF, which mainly comprises the French Atomic Commission (CEA), but only a few of EdF`s engineers were aware of these facts. One has to bear in mind that CHOOZ A-1 only went critical in April 1967 and was officially connected to the grid in May 1970 after some important problems had been solved. Meanwhile, the nuclear program was launched at full speed, beginning with the order for FESSENHEIM 1 in 1970, FESSENHEIM 2 and BUGEY 2 and 3 in 1971. TIHANGE 1, in which EdF had a share, went on-line in September 1975. Also, supposing that EdF had had such knowledge and experience, it is quite evident that it would have been very difficult to modify the lay-out inside the reactor building.

  14. Lessons learned from ten years of terrestrial radio - ecological monitoring campaigns conducted around chooz NPP between 1995 and 2005

    International Nuclear Information System (INIS)

    Burton, O.; Bourdon, M; Aubinet, M.; Calande, E. C.; Tomasevszky, D.; Sombre, L.

    2006-01-01

    Radio-ecological monitoring campaigns around the Chooz PWR nuclear power plant (Ardennes - France), located at a few kilometres of the Belgian border, have been conducted on the Belgian territory, between 1995 and 2005, with the aim of observing the concentrations of radionuclides of major environmental significance in soils and agricultural products and their time evolution. The following conclusions are drawn from these campaigns : - the time evolution of the observed radioactivity levels remains rather constant ; - except 1 4C which is of natural origin but is also released in airborne discharges of nuclear power plants, the artificial isotopes still significantly detected in the agricultural ecosystem are the 1 37Cs (generated by the Chernobyl accident and the military test explosions) and the 9 0Sr (generated by the military test explosions) ; - grass remains the best indicator of contamination of the agricultural ecosystem by 1 37Cs and 9 0Sr ; - the major contributors to the total soil and plant activities are from far radionuclides of natural origin, the major contributor being 4 0K

  15. The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings

    International Nuclear Information System (INIS)

    Rouelle, P.; Roy, F.

    1998-01-01

    The Electricite de France's N4 CHOOZ B nuclear power plant, two units of the world's largest PWR model (1450 Mwe each), has earned the Electric Power International's 1997 Powerplant Award. This lead NPP for EDF's N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building's inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter

  16. Echoes of the world; Echos du Monde

    Energy Technology Data Exchange (ETDEWEB)

    Anon

    2007-09-15

    EDF is authorized to dismantle Chooz A. The cooling towers of Calder Hall on the site of Sellafield were destroyed, within the framework of the demolition of the site.Belarus hopes to begin the building of a nuclear power plant in order to produce it own energy. Bangladesh is ready to build its first nuclear reactor. Germany cannot hold its commitment of CO{sub 2} reduction if it does not maintain the nuclear power production. Japan launches the next generation of light water advanced reactors. (N.C.)

  17. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series)

    International Nuclear Information System (INIS)

    Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

    1996-01-01

    The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors)

  18. New trends in shielding designs for PWRs in France

    International Nuclear Information System (INIS)

    Champion, G.; Forestier, J.; Arbelot, E.

    1985-01-01

    Various engineering solutions to confinement of the stray neutron fields will be incorporated into the design of the 1450 MWe Chooz B-1 (Ardennes B-1) PWR, the first unit of the new N4 program in France. The reactor is in the early stages of construction. These engineering solutions are the results of many shielding configuration studies performed prior to actual design. The solutions and the calculation methodologies are discussed

  19. Experimental modal analysis of the steam inlet pipe to the Chooz B1 high pressure turbine

    International Nuclear Information System (INIS)

    Guihot, O.; Anne, J.P.; Chartain, G.; Le Pironnec, D.

    1993-05-01

    This report presents the results of the modal analysis carried out on one of the steam inlet pipe of the high pressure turbine of the Chooz B1 power plant. This experimental analysis is made within the frame of the research and development project ''dynamical, acoustical and aerodynamical behaviour of the turbogenerator N4''. This research program provides amongst others, numerical studies with the software CIRCUS and ASTER, in order to verify the dynamical behaviour of the designed inlet pipe. The numerical models will be updated from results of the experimental modal analysis to improve the numerical representation of this pipe. All the identified modes in the frequency band [5.2000] Hz are presented in the report. The modal characteristics of the main modes are detailed. Further analysis have been made, in order ease the updating of the numerical models. They consisted in an analysis of the evolution of the dynamical behaviour due to a change of the boundary conditions of the inlet valve frame on one hand and resulting from the presence of an additional mass on the pipe, at the level of the middle flange, on the other hand. The analysis made in low frequency range shows that the pipe is thoroughly embedded in the frame of the high pressure turbine. On the other hand, the boundary conditions on the inlet valve frame are more difficult to determine, because the dynamical behaviour of the valve frame and the upper pipe can not be uncoupled from the considered pipe. The main shell modes of ranks 2, 3 and 4 have been very accurately identified. The most relevant modes to update the numerical models are given. (authors). 48 figs., 18 tabs., 4 refs

  20. EDF decommissioning programme a global commitment to safety, environment and cost efficiency of nuclear energy

    International Nuclear Information System (INIS)

    Grenouillet, J.-J.

    2002-01-01

    EDF has 9 NPPs permanently shutdown and under decommissioning. EDF considers that if the nuclear option is to remain open, it is necessary to deal with increasing public concerns for environmental and waste management issues. Therefore EDF has decided to achieve total dismantling of all shutdown reactor in the next 25 years. The Decommissioning Program has been developed including 2 stages of activities. The first stage consists of: 1) Final dismantling of Brennilis in 2015; 2) A dismantling demonstration of a PWR reactor building (Chooz A) before starting replacing the population of PWRs currently in operation; 3) Final dismantling of reactor containment of a GCR (Bugey 1) as a first of its kind. The second stage includes: 1)Dismantling of following 5 GCR (Saint Laurent A1 and A2, Chinon A1, A2 and A3); 2) Final dismantling of Chooz A and Bugey 1 in 2025. The successful implementation relies on the simplification of the regulatory process; availability of treatment, conditioning and disposal facilities and effective nuclear industry. The main issue is availability of time and waste solutions such as opening of a Very Low Waste disposal in 2003 (130 000 tons); opening of a new disposal for graphite and radiferous wastes (17 000 tons) in 2010 and opening in 2007-2008 of a centralized interim storage (BANEDA) facility for long-lived Medium Level Wastes (500 tons including filters, control rods etc)Three investigations are to be carried out for high level radioactive waste before 2006

  1. PARTICIPATION IN HIGH ENERGY PHYSICS

    Energy Technology Data Exchange (ETDEWEB)

    White, Christopher

    2012-12-20

    This grant funded experimental and theoretical activities in elementary particles physics at the Illinois Institute of Technology (IIT). The experiments in which IIT faculty collaborated included the Daya Bay Reactor Neutrino Experiment, the MINOS experiment, the Double Chooz experiment, and FNAL E871 - HyperCP experiment. Funds were used to support summer salary for faculty, salary for postdocs, and general support for graduate and undergraduate students. Funds were also used for travel expenses related to these projects and general supplies.

  2. Online Data Monitoring Framework Based on Histogram Packaging in Network Distributed Data Acquisition Systems

    International Nuclear Information System (INIS)

    Konno, T; Ishitsuka, M; Kuze, M; Cabarera, A; Sakamoto, Y

    2011-01-01

    O nline monitor frameworkis a new general software framework for online data monitoring, which provides a way to collect information from online systems, including data acquisition, and displays them to shifters far from experimental sites. 'Monitor Server', a core system in this framework gathers the monitoring information from the online subsystems and the information is handled as collections of histograms named H istogram Package . Monitor Server broadcasts the histogram packages to 'Monitor Viewers', graphical user interfaces in the framework. We developed two types of the viewers with different technologies: Java and web browser. We adapted XML based file for the configuration of GUI components on the windows and graphical objects on the canvases. Monitor Viewer creates its GUIs automatically with the configuration files.This monitoring framework has been developed for the Double Chooz reactor neutrino oscillation experiment in France, but can be extended for general application to be used in other experiments. This document reports the structure of the online monitor framework with some examples from the adaption to the Double Chooz experiment.

  3. Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor

    International Nuclear Information System (INIS)

    Magat, Ph.

    1997-04-01

    Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P N method (SP N ) for the modeling of the core. The comparison between S N calculations and SP N calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P 3 , there is no effect; -) the P 1 development is adequate; and -) the P 0 development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP 1 , SP 3 , SP 5 and SP 7 ). These calculations show that the implementation of the SP N method is feasible but the extra-costs in computation times and memory are important. We recommend: SP 5 P 1 calculations for heterogeneous 2-dimension geometry and SP 3 P 1 calculations for the homogeneous 3-dimension geometry. (A.C.)

  4. Neutrino Interactions

    Energy Technology Data Exchange (ETDEWEB)

    Kamyshkov, Yuri [Univ. of Tennesse, Knoxville, TN (United States); Handler, Thomas [Univ. of Tennesse, Knoxville, TN (United States)

    2016-10-24

    The neutrino group of the University of Tennessee, Knoxville was involved from 05/01/2013 to 04/30/2015 in the neutrino physics research funded by DOE-HEP grant DE-SC0009861. Contributions were made to the Double Chooz nuclear reactor experiment in France where second detector was commissioned during this period and final series of measurements has been started. Although Double Chooz was smaller experimental effort than competitive Daya Bay and RENO experiments, its several advantages make it valuable for understanding of systematic errors in measurements of neutrino oscillations. Double Chooz was the first experiment among competing three that produced initial result for neutrino angle θ13 measurement, giving other experiments the chance to improve measured value statistically. Graduate student Ben Rybolt defended his PhD thesis on the results of Double Chooz experiment in 2015. UT group has fulfilled all the construction and analysis commitments to Double Chooz experiment, and has withdrawn from the collaboration by the end of the mentioned period to start another experiment. Larger effort of UT neutrino group during this period was devoted to the participation in another DOE-HEP project - NOvA experiment. The 14,000-ton "FAR" neutrino detector was commissioned in northern Minnesota in 2014 together with 300-ton "NEAR" detector located at Fermilab. Following that, the physics measurement program has started when Fermilab accelerator complex produced the high-intensity neutrino beam propagating through Earth to detector in MInnessota. UT group contributed to NOvA detector construction and developments in several aspects. Our Research Associate Athanasios Hatzikoutelis was managing (Level 3 manager) the construction of the Detector Control System. This work was successfully accomplished in time with the commissioning of the detectors. Group was involved in the development of the on-line software and study of the signatures of the cosmic ray backgrounds

  5. Neutrino Interactions

    International Nuclear Information System (INIS)

    Kamyshkov, Yuri; Handler, Thomas

    2016-01-01

    The neutrino group of the University of Tennessee, Knoxville was involved from 05/01/2013 to 04/30/2015 in the neutrino physics research funded by DOE-HEP grant DE-SC0009861. Contributions were made to the Double Chooz nuclear reactor experiment in France where second detector was commissioned during this period and final series of measurements has been started. Although Double Chooz was smaller experimental effort than competitive Daya Bay and RENO experiments, its several advantages make it valuable for understanding of systematic errors in measurements of neutrino oscillations. Double Chooz was the first experiment among competing three that produced initial result for neutrino angle θ_1_3 measurement, giving other experiments the chance to improve measured value statistically. Graduate student Ben Rybolt defended his PhD thesis on the results of Double Chooz experiment in 2015. UT group has fulfilled all the construction and analysis commitments to Double Chooz experiment, and has withdrawn from the collaboration by the end of the mentioned period to start another experiment. Larger effort of UT neutrino group during this period was devoted to the participation in another DOE-HEP project - NOvA experiment. The 14,000-ton 'FAR' neutrino detector was commissioned in northern Minnesota in 2014 together with 300-ton 'NEAR' detector located at Fermilab. Following that, the physics measurement program has started when Fermilab accelerator complex produced the high-intensity neutrino beam propagating through Earth to detector in MInnessota. UT group contributed to NOvA detector construction and developments in several aspects. Our Research Associate Athanasios Hatzikoutelis was managing (Level 3 manager) the construction of the Detector Control System. This work was successfully accomplished in time with the commissioning of the detectors. Group was involved in the development of the on-line software and study of the signatures of the cosmic ray backgrounds

  6. Implications of the Super-K atmospheric, long baseline, and reactor data for the mixing angles θ13 and θ23

    Science.gov (United States)

    Escamilla-Roa, J.; Latimer, D. C.; Ernst, D. J.

    2010-01-01

    A three-neutrino analysis of oscillation data is performed using the recent, more finely binned Super-K oscillation data, together with the CHOOZ, K2K, and MINOS data. The solar parameters Δ21 and θ12 are fixed from a recent analysis and Δ32, θ13, and θ23 are varied. We utilize the full three-neutrino oscillation probability and an exact treatment of Earth’s Mikheyev-Smirnov-Wolfenstein (MSW) effect with a castle-wall density. By including terms linear in θ13 and ɛ:=θ23-π/4, we find asymmetric errors for these parameters θ13=-0.07-0.11+0.18 and ɛ=0.03-0.15+0.09. For θ13, we see that the lower bound is primarily set by the CHOOZ experiment while the upper bound is determined by the low energy e-like events in the Super-K atmospheric data. We find that the parameters θ13 and ɛ are correlated—the preferred negative value of θ13 permits the preferred value of θ23 to be in the second octant, and the true value of θ13 affects the allowed region for θ23.

  7. Constraints on θ13 from a three-flavor oscillation analysis of reactor antineutrinos at KamLAND

    International Nuclear Information System (INIS)

    Gando, A.; Gando, Y.; Ichimura, K.; Ikeda, H.; Kibe, Y.; Kishimoto, Y.; Minekawa, Y.; Mitsui, T.; Morikawa, T.; Nagai, N.; Nakajima, K.; Narita, K.; Shimizu, I.; Shimizu, Y.; Shirai, J.; Suekane, F.; Suzuki, A.; Takahashi, H.; Takahashi, N.; Takemoto, Y.

    2011-01-01

    We present new constraints on the neutrino oscillation parameters Δm 21 2 , θ 12 , and θ 13 from a three-flavor analysis of solar and KamLAND data. The KamLAND data set includes data acquired following a radiopurity upgrade and amounts to a total exposure of 3.49x10 32 target-proton-year. Under the assumption of CPT invariance, a two-flavor analysis (θ 13 =0) of the KamLAND and solar data yields the best-fit values tan 2 θ 12 =0.444 -0.030 +0.036 and Δm 21 2 =7.50 -0.20 +0.19 x10 -5 eV 2 ; a three-flavor analysis with θ 13 as a free parameter yields the best-fit values tan 2 θ 12 =0.452 -0.033 +0.035 , Δm 21 2 =7.50 -0.20 +0.19 x10 -5 eV 2 , and sin 2 θ 13 =0.020 -0.016 +0.016 . This θ 13 interval is consistent with other recent work combining the CHOOZ, atmospheric and long-baseline accelerator experiments. We also present a new global θ 13 analysis, incorporating the CHOOZ, atmospheric, and accelerator data, which indicates sin 2 θ 13 =0.009 -0.007 +0.013 . A nonzero value is suggested, but only at the 79% C.L.

  8. Contribution to the build-up of a core calculation frame: comparison between ''diffusion'' and ''SPn'' operators on various configurations of the first N4 core

    International Nuclear Information System (INIS)

    Magat, Ph.

    1997-04-01

    The aim of this study is to compare two calculation methods implemented in the neutronic code CRONOS 2: the diffusion approximation and the SP n method. The APOLLO 2 code is used to build the multiparameter cross section libraries.The comparison is based on the first core of N4 type Chooz reactor. The rod worth and the power map have been calculated. Some recommendations about the SP n development order of flux are made and the results show that the diffusion calculations over-estimate the black rod efficiency up to 10%. (A.C.)

  9. French N4 Series Design and Manufacturing

    International Nuclear Information System (INIS)

    Lebreton, Gerard

    1989-01-01

    The construction contract for the first two N4 units. on the Chooz site close to the French-Belgian border (Chooz B1 and B2). was awarded by Electricite de France (EDF) to Fumarate in May 1984. At present, project construction is approximately 50% complete for unit B1. The main civil works are practically finished and the reactor vessel and the steam generator were delivered on site by mid-1988. The connection to the grid by 1991 appears quite feasible. Continuation of the French nuclear power programme in the 1990s, specifically on the CIVEX and Penly sites, is based on this model. The N4 thus follows the four-loop, 1,300 MW class series designated P.a. whose first unit is Cattenom 1, which will serve in the following as a reference to appreciate the design evolution of the N4 NSSS. The design evolutions selected to reach these objectives are in technical continuity with the previous series, integrate the vast experience gained within the French programme and abroad, and were supported by a large R and D programme and further by industrial qualification tests

  10. Inspection of Chooz power plant after ten years operation

    International Nuclear Information System (INIS)

    Saglio, Robert.

    1978-01-01

    This report is intended to discuss the results from the complete technical audit of the vessel effected in 1976 at the Ardennes reactor (CNA 305 MWe). This audit had a special character as far as this power plant has never been inspected before: the start-up had taken place in 1967 and was then prior to the development of French regulations (and even to the first version of the ASME code, Section XI). In that time, no inspection was expected; it has yet been possible to have a complete audit in ten days. The automatic start-up of focused ultrasonic testing so appeared to have reached the required reliability and a good sensitivity [fr

  11. Status of fast reactors and ADS programmes in France in 2000

    International Nuclear Information System (INIS)

    Astegiano, J.C.; Tailland, C.; Grenet, P.; Frith, R.; Fiorini, J.L.; Louvet, J.

    2001-01-01

    Two of the 4 N4 units (Chooz B1 et 2) were declared to be in commercial operation during 2000. The EDF owns 58 PWRs, for a total capacity of 63 GWe. The availability factor was 80.8 %, the new N4s run well. No particular incident occurred, 469 incidents were declared, 134 of which were rated 'level 1' and 2 incidents, 'level 2'. These incidents were often attributed to human factors originally and lasting generic problems. France, thanks to its large infrastructure of existing nuclear power stations which provides it with cheap energy, also finds itself in a strong position in the battle against 'The Greenhouse Effect'. The General Administrator of the CEA, Mr. Pascal Colombani, has called for a general restructuration of the nuclear industry in order to meet the challenges of deregulation and globalisation. The French Government has confirmed a process due to be completed in autumn of this year. Two main conclusions emerge dealing with the Long-Term Options for a French energy policy: Existing nuclear plants should retain their cost advantage over combined-cycle gas plants - the only serious economic challenge in terms of meeting the bulk of future electricity demand; Unless gas prices remain stable over the period as a whole, nuclear energy is also likely to retain its economic edge when the time comes to build new generating capacity. Renovation and inspection works have been actively pursued on the PHENIX plant, a large part of the work has been completed, mainly reinforcement against earthquakes, inspection of the internal structures and general maintenance. Maintenance work revealed cracks in the SGU and it has been decided to replace the potentially affected zones of SGU 1 and 3. Decommissioning work is underway at SUPERPHENIX, core unloading has proceeded normally, an increasing number of systems and components have been removed from service. A new decree will have to be issued for the final disposal of both the primary and secondary sodium; it will include

  12. Inception of the light water reactor system in France and its fuel cycle

    International Nuclear Information System (INIS)

    Gaussot, D.; Elkouby, A.; Sornein, J.

    1977-01-01

    The electro-nuclear equipment program of Electricite de France (E.D.F.) currently is based on the construction or the commitment of a considerable number of units equiped with pressurized water reactors. The program was preceded by the construction jointly by EDF and Belgian electric utilities of nuclear plants at Chooz (SENA) and Tihange (SEMO). The program as such began with the construction of Units 1 and 2 at Fessenheim, (of which information is given on the start-up), then of Bugey 2, 3, 4 and 5. This was followed by the construction of a series of units of 900 MW, and since 1976, of a series of units of 1300 MW. The successful implementation of such a program is based on a number of technical and organizational guidelines, which are described. The main characteristics of the 3-loop 900 MW and 4-loop 1300 MW NSSS and their fuel are considered. Stress is laid in particular on the development of the 900 MW NSSS from Fessenheim to Bugey. These programs call for winning and processing a great deal of nuclear material right through the fuel cycle. Data are given on the quantities involved and on the production potential permitting the fulfillment of the program (nat. U, enriched UF 6 , fuel subassemblies, reprocessing). The requirements of the EDF, (the NSSS supplier and the industries), the contribution made by the French Atomic Energy Commission (CEA), and international cooperation now in progress, are described. Lastly a number of significant actions both under way and scheduled, are discussed [fr

  13. Contribution to the build-up of a core calculation frame: comparison between ``diffusion`` and ``SP{sub n}`` operators on various configurations of the first N4 core; Contribution a l`elaboration d`un schema de coeur en transport: comparaison des operateurs ``diffusion`` et ``SP{sub n}`` sur differentes configurations du premier coeur N4

    Energy Technology Data Exchange (ETDEWEB)

    Magat, Ph

    1997-04-01

    The aim of this study is to compare two calculation methods implemented in the neutronic code CRONOS 2: the diffusion approximation and the SP{sub n} method. The APOLLO 2 code is used to build the multiparameter cross section libraries.The comparison is based on the first core of N4 type Chooz reactor. The rod worth and the power map have been calculated. Some recommendations about the SP{sub n} development order of flux are made and the results show that the diffusion calculations over-estimate the black rod efficiency up to 10%. (A.C.) 12 refs.

  14. PWR internals segmentation and packaging experience in the U.S

    International Nuclear Information System (INIS)

    Kreitman, P.J.

    2008-01-01

    Seven commercial nuclear power plants of the Pressurized Water Reactor (PWR) design have been permanently shut down in the US to date. Six of these plants have been decommissioned using dismantling methods. The remaining plant, Indian Point 1, located in Buchanan, NY, has been placed in Safe Storage. Of the six dismantled plants, five underwent extensive segmentation, separation, and packaging of their internal components to allow removal and disposal of the reactor vessel assembly. Only the Trojan Plant in Portland, Oregon, was able to ship the reactor vessel assembly with its internals intact to the final disposal site near Richland, Washington. An important part of the planning for the upcoming decommissioning of similar plants in Europe such as Chooz A in France and Zorita in Spain will be to study the lessons learned from the US efforts and apply the best practices to future projects. This paper will chronicle the evolution of the reactor internals segmentation and packaging process to date, including the planning, methodology, equipment, waste management, and packaging strategy. (author)

  15. Feedback from dismantling operations (level 2) on EDF's first generation reactors

    International Nuclear Information System (INIS)

    West, J P.; Dionisio-Gomes, A.; Kus, J P.; Mervaux, P.; Bernet, P.; Dalmas, R.

    2003-01-01

    EDF's policy as regards the dismantling of the reactors that have ceased commercial operation, namely the eight power plants of the first generation and the Creys-Malville power plant, is explained. Generally speaking, prior to the year 2001, EDF had opted for the de-construction of these power plants to comply with a 'long wait' scenario, which consisted of waiting for a period of 5 to 10 years to achieve IAEA level 2 (partial release of the site), then postponing the total de-construction of the facilities for 25 to 50 years. Today, EDF has decided to undertake the total de-construction of these reactors, which have ceased commercial operation, over a period of 25 years. The purpose of this document is to present: - The reactors concerned, their background and their 'regulatory' situation, - The main operations performed and/or currently in progress, - The main elements of feedback from such operations, shedding light on the approach adopted in 2001. The installations concerned by the de-construction programme are as follows: - The 8 power plants of the first generation, which were built during the fifties and sixties and ceased commercial operation between 1973 and 1994, namely: Brennilis (industrial prototype using heavy water technology, jointly operated by EDF and CEA), the 6 power units of the NUGG type (natural uranium gas graphite) at Chinon, Saint-Laurent des Eaux and Bugey and the PWR reactor at Chooz A, - The storage silos at Saint-Laurent, where the sleeves for the fuel assemblies of reactors SLA1 and SLA2 are stored, corresponding to approximately 2000 tonnes of graphite, - The Creys-Malville reactor, FBR (fast breeder reactor) shut down in accordance with a government decision, which is currently undergoing decommissioning. At the current stage, our feedback from the dismantling operations carried out on nuclear facilities is based on (i) the work carried out or in progress that will make it possible to achieve the equivalent of IAEA level 2 in the

  16. Experience acquired by EDF in implementation of its dismantling programme

    Energy Technology Data Exchange (ETDEWEB)

    Klaeyle, S.; Dalmas, R.; Davoust, M. [EDF - Centre d' Ingenerie Deconstruction Environnement (CIDEN), 69 - Villeurbanne (France)

    2008-07-01

    EDF decided in 2001 to implement immediate dismantling of its first generation nuclear plants. Seven years after this decision, the physical progress of the programme is 24% and is due to reach 50 % by 2013. This paper presents the experiences acquired in the fields of organization, project and programme management, purchasing strategies and waste management. Until now, the principal works involve Brennilis (Heavy water), ChoozA (PWR) and Creys Malville (fast breeder reactor). The detailed pre-project concerning the first of the six gas graphite reactors is complete and the call for bids process has been launched. The organization to manage projects, established at the De-construction and Environment Engineering Center (CIDEN), is effective and productive. Estimates of costs and expenses are coherent, which makes the forecasts put together to finance the programme secure. CIDEN has carried out significant engineering work over the last six years, making it possible to apply for the administrative authorizations which have now been obtained or are in the process of being obtained. Technical specifications are prepared at an optimized level of detail according to a contractual policy adapted to the complexity of the operations and the sharing of risk with manufacturers. The ChoozA contractualization process has been launched and the first dismantling work has begun in the nuclear auxiliary part. The main Brennilis contract will be completed in mid- 2008 and dismantling works will restart after renewal of the decree which was cancelled in mid-2007. Treatment of sodium from Creys Malville is about to begin, leading to elimination of the sodium risk by 2013. The very low activity waste (TFA) and low to medium activity waste (FA-MA) removal chains are operational. The intermediate activity/long lived (MA-VL) waste will be stored in a facility which will be brought into operational service in 2012. The graphite storage center is due to open between 2017 and 2019

  17. Determining the neutrino mass hierarchy with INO, T2K, NOvA and reactor experiments

    International Nuclear Information System (INIS)

    Ghosh, Anushree; Choubey, Sandhya; Thakore, Tarak

    2013-01-01

    The relatively large measured value of θ 13 has opened up the possibility of determining the neutrino mass hierarchy through earth matter effects. Amongst the current accelerator experiments only NOvA has a long enough baseline to observe earth matter effects. However, even NOvA is plagued with uncertainty on the knowledge of the true value of Δ CP which drastically reduces its sensitivity to the neutrino mass hierarchy. Earth matter effects in atmospheric neutrinos on the other hand is almost independent of δ CP . The 50 kton magnetized Iron CALorimeter at the India-based Neutrino Observatory (ICAL at the rate lNO) will be observing atmospheric neutrinos. The charge identification capability of this detector gives it an edge over others for mass hierarchy determination through observation of earth matter effects. We study in detail the neutrino mass hierarchy sensitivity of the data from this experiment simulated using the Nuance based generator developed for ICAL at the rate lNO and folded with the detector resolution and efficiencies obtained by the INO collaboration from a full detector Geant based simulation. The data from ICAL at the rate lNO is then combined with simulated of T2K, NOvA Double Chooz, RENO and Daya Bay experiments and a combined sensitivity study to the mass hierarchy performed. With 10 years of ICAL at the rate lNO data combined with T2K, NOvA and reactor data, one could get 2.8σ - 5σ discovery for the neutrino mass hierarchy depending on the true value of (θ23, θ13 and δ CP . (author)

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

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

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

  19. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  20. Reactor core and initially loaded reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo.

    1989-01-01

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

  1. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

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

  2. Reactor core of FBR type reactor

    International Nuclear Information System (INIS)

    Hayashi, Hideyuki; Ichimiya, Masakazu.

    1994-01-01

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

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  4. Reactor core for LMFBR type reactors

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  5. Nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

  6. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  8. The CAREM reactor and present currents in reactor design

    International Nuclear Information System (INIS)

    Ordonez, J.P.

    1990-01-01

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

  9. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  10. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

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

  11. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  12. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  13. Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

    International Nuclear Information System (INIS)

    Hirayama, Shoichi; Kawabe, Takaya

    1995-01-01

    The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

  14. Reactor physics aspects of CANDU reactors

    International Nuclear Information System (INIS)

    Critoph, E.

    1980-01-01

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

  15. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  16. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  17. Nuclear reactor types

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

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

  18. Reactor containment and reactor safety in the United States

    International Nuclear Information System (INIS)

    Kouts, H.

    1986-01-01

    The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

  19. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  20. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  1. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  2. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  3. The neutrino in all its states - Seminar dedicated to Jacques Bouchez - Slides of the presentations

    International Nuclear Information System (INIS)

    Spiro, M.; Pessard, H.; Rubbia, A.; Petcov, S.; Cousins, B.; Fechner, M.; Mezetto, M.

    2011-01-01

    The present scientific seminar, organized in the memory of Jacques Bouchez is centered on neutrino physics and presents the state of the art on experiments, on future projects and on the theory of neutrinos (oscillations and MSW effect). This document is made up of the slides of 7 presentations: 1) The achievements of J.Bouchez; 2) Reactor neutrino experiments from Bugey to double-Chooz (via RENO and Daya-Bay); 3) Neutrinos and accelerators: on the way toward the third flavor (NOMA, OPERA and T2K experiments); 4) Neutrino oscillations and MSW effect; 5) Some statistical questions in neutrino physics; 6) Long baseline oscillations: towards Japan future neutrino oscillation experiments; and 7) Next generation of neutrino oscillation experiments. (A.C.)

  4. Large short-baseline νμ disappearance

    International Nuclear Information System (INIS)

    Giunti, Carlo; Laveder, Marco

    2011-01-01

    We analyze the LSND, KARMEN, and MiniBooNE data on short-baseline ν μ →ν e oscillations and the data on short-baseline ν e disappearance obtained in the Bugey-3 and CHOOZ reactor experiments in the framework of 3+1 antineutrino mixing, taking into account the MINOS observation of long-baseline ν μ disappearance and the KamLAND observation of very-long-baseline ν e disappearance. We show that the fit of the data implies that the short-baseline disappearance of ν μ is relatively large. We obtain a prediction of an effective amplitude sin 2 2θ μμ > or approx. 0.1 for short-baseline ν μ disappearance generated by 0.2 2 2 , which could be measured in future experiments.

  5. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  6. Research reactor standards and their impact on the TRIGA reactor community

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

    The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

  7. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  8. Improved nuclear reactor construction with bottom supported reactor vessel

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1987-01-01

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

  9. Reactor as furnace and reactor as lamp

    International Nuclear Information System (INIS)

    Goldanskii, V.I.

    1992-01-01

    There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

  10. Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

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

    2012-01-01

    The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

  11. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  12. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  13. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  14. The research reactors their contribution to the reactors physics

    International Nuclear Information System (INIS)

    Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

    2000-01-01

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

  15. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

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

  16. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  17. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  18. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  19. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  20. FBR type reactor

    International Nuclear Information System (INIS)

    Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

    1994-01-01

    In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

  1. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

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

  2. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  3. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

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

  4. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

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

  5. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  6. Nuclear reactor, reactor core thereof, and device for constituting the reactor

    International Nuclear Information System (INIS)

    Takiyama, Masashi.

    1994-01-01

    A reactor core is constituted by charging coolants (light water) in a reactor pressure vessel and distributing fuel assemblies, reflecting material sealing pipes, moderator (heavy water and helium gas) sealing pipes, and gas sealing pipes therein. A fuel guide tube is surrounded by a cap and the gap therebetween is made hollow and filled with coolant steams. The cap is supported by a baffle plate. The moderator sealing pipe is disposed in a flow channel of coolants in adjacent with the cap. The position of the moderator sealing tube in the reactor core is controlled by water stream from a hydraulic pump with a guide tube extending below the baffle plate being as a guide. Then, the position of the moderator sealing tube is varied to conduct power control, burnup degree compensation, and reactor shut down. With such procedures, moderator cooling facility is no more necessary to simplify the structure. Further, heat generated from the moderator is transferred to the coolants thereby improving heat efficiency of the reactor. (I.N.)

  7. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  8. Optimal reactor strategy for commercializing fast breeder reactors

    International Nuclear Information System (INIS)

    Yamaji, Kenji; Nagano, Koji

    1988-01-01

    In this paper, a fuel cycle optimization model developed for analyzing the condition of selecting fast breeder reactors in the optimal reactor strategy is described. By dividing the period of planning, 1966-2055, into nine ten-year periods, the model was formulated as a compact linear programming model. With the model, the best mix of reactor types as well as the optimal timing of reprocessing spent fuel from LWRs to minimize the total cost were found. The results of the analysis are summarized as follows. Fast breeder reactors could be introduced in the optimal strategy when they can economically compete with LWRs with 30 year storage of spent fuel. In order that fast breeder reactors monopolize the new reactor market after the achievement of their technical availability, their capital cost should be less than 0.9 times as much as that of LWRs. When a certain amount of reprocessing commitment is assumed, the condition of employing fast breeder reactors in the optimal strategy is mitigated. In the optimal strategy, reprocessing is done just to meet plutonium demand, and the storage of spent fuel is selected to adjust the mismatch of plutonium production and utilization. The price hike of uranium ore facilitates the commercial adoption of fast breeder reactors. (Kako, I.)

  9. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    Buerger, L.

    1982-01-01

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  10. Safeguarding research reactors

    International Nuclear Information System (INIS)

    Powers, J.A.

    1983-03-01

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

  11. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  12. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  13. Reactor core for FBR type reactor

    International Nuclear Information System (INIS)

    Fujita, Tomoko; Watanabe, Hisao; Kasai, Shigeo; Yokoyama, Tsugio; Matsumoto, Hiroshi.

    1996-01-01

    In a gas-sealed assembly for a FBR type reactor, two or more kinds of assemblies having different eigen frequency and a structure for suppressing oscillation of liquid surface are disposed in a reactor core. Coolant introduction channels for introducing coolants from inside and outside are disposed in the inside of structural members of an upper shielding member to form a shielding member-cooling structure in the reactor core. A structure for promoting heat conduction between a sealed gas in the assembly and coolants at the inner side or the outside of the assembly is disposed in the reactor core. A material which generates heat by neutron irradiation is disposed in the assembly to heat the sealed gases positively by radiation heat from the heat generation member also upon occurrence of power elevation-type event to cause temperature expansion. Namely, the coolants flown out from or into the gas sealed-assemblies cause differential fluctuation on the liquid surface, and the change of the capacity of a gas region is also different on every gas-sealed assemblies thereby enabling to suppress fluctuation of the reactor power. Pressure loss is increased by a baffle plate or the like to lower the liquid surface of the sodium coolants or decrease the elevating speed thereof thereby suppressing fluctuation of the reactor power. (N.H.)

  14. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  15. Nuclear Reactor RA Safety Report, Vol. 4, Reactor

    International Nuclear Information System (INIS)

    1986-11-01

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

  16. Feasible reactor power cutback logic development for an integral reactor

    International Nuclear Information System (INIS)

    Han, Soon-Kyoo; Lee, Chung-Chan; Choi, Suhn; Kang, Han-Ok

    2013-01-01

    Major features of integral reactors that have been developed around the world recently are simplified operating systems and passive safety systems. Even though highly simplified control system and very reliable components are utilized in the integral reactor, the possibility of major component malfunction cannot be ruled out. So, feasible reactor power cutback logic is required to cope with the malfunction of components without inducing reactor trip. Simplified reactor power cutback logic has been developed on the basis of the real component data and operational parameters of plant in this study. Due to the relatively high rod worth of the integral reactor the control rod assembly drop method which had been adapted for large nuclear power plants was not desirable for reactor power cutback of the integral reactor. Instead another method, the control rod assembly control logic of reactor regulating system controls the control rod assembly movements, was chosen as an alternative. Sensitivity analyses and feasibility evaluations were performed for the selected method by varying the control rod assembly driving speed. In the results, sensitivity study showed that the performance goal of reactor power cutback system could be achieved with the limited range of control rod assembly driving speed. (orig.)

  17. Nuclear power, society and environment

    International Nuclear Information System (INIS)

    Fouchet, N.

    1997-01-01

    This rubric reports on 12 short notes about scientific facts, and sociological, political and environmental aspects of nuclear power in France and other countries: a new micro-beam line for the nuclear micro-probe of Pierre Sue laboratory; the French government gives permission for the filling up of the Carnet swampy site for the possible sitting of a future nuclear power plant in the Loire river estuary; incident simulation exercise at Chooz B1 in January 1997: radioactive leak and population under shelter; about Superphenix, 'Le Monde' newspaper disseminates false information; the anti-Superphenix lobby; Georges Charpak's opinion about anti-nuclear propaganda; gamma radiation in the help of cultural heritage; a new ionizing particle detector developed by the CEA; dismantling of the FR-2 experimental reactor (Karlsruhe, Germany) and the safe confinement of the reactor vessel; the Russian specialists' proposal for the transformation of Tchernobyl's sarcophagus into a monolith of concrete; Cogema's support to scientific research devoted to environment and public health; three new member countries in the World Council of Nuclear Workers (WONUC). (J.S.)

  18. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  19. Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor; Analyse des techniques d'homogeneisation spatiale et energetique dans la resolution de l'equation du transport des neutrons dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Magat, Ph

    1997-04-01

    Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P{sub N} method (SP{sub N}) for the modeling of the core. The comparison between S{sub N} calculations and SP{sub N} calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P{sub 3}, there is no effect; -) the P{sub 1} development is adequate; and -) the P{sub 0} development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP{sub 1}, SP{sub 3}, SP{sub 5} and SP{sub 7}). These calculations show that the implementation of the SP{sub N} method is feasible but the extra-costs in computation times and memory are important. We recommend: SP{sub 5}P{sub 1} calculations for heterogeneous 2-dimension geometry and SP{sub 3}P{sub 1} calculations for the homogeneous 3-dimension geometry. (A.C.)

  20. H Reactor

    Data.gov (United States)

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

  1. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  2. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  3. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  4. Strategic planning for research reactors. Guidance for reactor managers

    International Nuclear Information System (INIS)

    2001-04-01

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

  5. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  6. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  7. Multiregion reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  8. Prediction of the brittle-ductile transition temperature shift, from irradiation experiments obtained in France

    International Nuclear Information System (INIS)

    Miannay, D.; Dussarte, D.; Soulat, P.

    1988-10-01

    In integrety - evaluation studies of the nuclear-reactor vessel, the toughness of component materials is given by a reference curve, which represents the lowest level of the toughness possible values, as a function of the temperature. Its temperature-scale position is given by the null ductility temperature. In non-irradiated conditions, the temperature of the vessel materials is experimentally obtained. To take into account the irradiation effect, a shift of this temperature, given by the correlations where chemical composition and neutronic dose are included, is considered. Up to date correlations, settled for materials brittler than those applied in the French program, are called in question again. A correlation between results obtained in France since 1973, for base metals and welded joints is given. The behavior of CHOOZ vessel, under irradiation conditions, is discussed [fr

  9. The decommissioning of nuclear power stations. France and the American experience

    International Nuclear Information System (INIS)

    Tonnac, Alain de; Perves, Jean-Pierre

    2013-01-01

    After having noticed that many electronuclear reactors have already been stopped, and that some research reactors have just been dismantled in Grenoble by the CEA, and evoked the main involved actors (ASN, IRSN, ANDRA), some characteristics of the French situation are outlined: the strategy adopted by EDF for a fast dismantling process, the long legal delays for stopping, dismantling and downgrading of sites, the absence of waste radioactivity release thresholds, and the responsibility of operators for dismantling operations. The main steps of a station dismantling are indicated and discussed: end of exploitation; steps of deconstruction (evacuation of big components, elimination of building radioactivity, building destruction, site partial or total reconversion). The French authorization process is briefly commented. Then, the American experience and situation is commented in terms of number of concerned sites, of site situation, but also in terms of organisation and selected technologies. Critical steps are outlined as well as improvements obtained in terms of exposure, of waste volumes, of effluent filtration capacities, and of cutting methods. The next part proposes an overview of the situation or steps of decommissioning in France for different sites: the Chooz PWR, the fleet of first generation reactors, graphite gas reactors, the Brennilis power station and Super Phenix in Creys Malville. Costs and delays are briefly evoked for the USA as well as for France

  10. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  12. Reactor physics challenges in GEN-IV reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, Michael K.; Hejzlar, Pavel [Massachusetts Institute of Technology, MA (United States)

    2005-02-15

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources.

  13. Reactor physics challenges in GEN-IV reactor design

    International Nuclear Information System (INIS)

    Driscoll, Michael K.; Hejzlar, Pavel

    2005-01-01

    An overview of the reactor physics aspects of GENeration Four (GEN-IV) advanced reactors is presented, emphasizing how their special requirements for enhanced sustainability, safety and economics motivates consideration of features not thoroughly analyzed in the past. The resulting concept-specific requirements for better data and methods are surveyed, and some approaches and initiatives are suggested to meet the challenges faced by the international reactor physics community. No unresolvable impediments to successful development of any of the six major types of proposed reactors are identified, given appropriate and timely devotion of resources

  14. Reactor Vessel Surveillance Program for Advanced Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

    2008-10-15

    This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

  15. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  16. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  17. Analysis of space and energy homogenization techniques for the solving of the neutron transport equation in nuclear reactor; Analyse des techniques d'homogeneisation spatiale et energetique dans la resolution de l'equation du transport des neutrons dans les reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Magat, Ph

    1997-04-01

    Today neutron transport in PWR's core is routinely computed through the transport-diffusion(2 groups) scheme. This method gives satisfactory results for reactors operating in normal conditions but the 2 group diffusion approximation is unable to take into account interface effects or anisotropy. The improvement of this scheme is logically possible through the use of a simplified P{sub N} method (SP{sub N}) for the modeling of the core. The comparison between S{sub N} calculations and SP{sub N} calculations shows an excellent agreement on eigenvalues as well as on power maps. We can notice that: -) it is no use extending the development beyond P{sub 3}, there is no effect; -) the P{sub 1} development is adequate; and -) the P{sub 0} development is totally inappropriate. Calculations performed on the N4 core of the Chooz power plant have enabled us to compare diffusion operators with transport operators (SP{sub 1}, SP{sub 3}, SP{sub 5} and SP{sub 7}). These calculations show that the implementation of the SP{sub N} method is feasible but the extra-costs in computation times and memory are important. We recommend: SP{sub 5}P{sub 1} calculations for heterogeneous 2-dimension geometry and SP{sub 3}P{sub 1} calculations for the homogeneous 3-dimension geometry. (A.C.)

  18. Tokamak reactor studies

    International Nuclear Information System (INIS)

    Baker, C.C.

    1981-01-01

    This paper presents an overview of tokamak reactor studies with particular attention to commercial reactor concepts developed within the last three years. Emphasis is placed on DT fueled reactors for electricity production. A brief history of tokamak reactor studies is presented. The STARFIRE, NUWMAK, and HFCTR studies are highlighted. Recent developments that have increased the commercial attractiveness of tokamak reactor designs are discussed. These developments include smaller plant sizes, higher first wall loadings, improved maintenance concepts, steady-state operation, non-divertor particle control, and improved reactor safety features

  19. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  1. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

  2. Reactor. Mind picture of the future Jules-Horowitz Reactor (RHJ)

    International Nuclear Information System (INIS)

    Eustache, S.

    1999-01-01

    This paper gives information about the future research reactor, named Reactor Jules-Horowitz (RJH). This irradiation reactor will be placed at industrialists disposal, for research concerning the competitiveness and the safety french electro-nuclear park. Principles and innovations are detailed. This reactor will respect the ALARA principle (as low as reasonably achievable). (A.L.B.)

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

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

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

  4. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  5. FBR type reactor

    International Nuclear Information System (INIS)

    Hayase, Tamotsu.

    1991-01-01

    The present invention concerns an FBR type reactor in which transuranium elements are eliminated by nuclear conversion. There are loaded reactor core fuels being charged with mixed oxides of plutonium and uranium, and blanket fuels mainly comprising depleted uranium. Further, liquid sodium is used as coolants. As transuranium elements, isotope elements of neptunium, americium and curium contained in wastes taken out from light water reactors or the composition thereof are used. The reactor core comprises a region with a greater mixing ratio and a region with a less mixing ratio of the transuranium elements. The mixing ratio of the transuranium elements is made greater for the fuels in the reactor core region at the boundary with the blanket of great neutron leakage. With such a constitution, since the positive reactivity value at the reactor core central portion is small in the Na void reactivity distribution in the reactor core, the positive reactivity is small upon Na boiling in the reactor core central region upon occurrence of imaginable accident, to attain reactor safety. (I.N.)

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

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

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

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

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

  10. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-01

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

  12. Nuclear reactor instrumentation at research reactor renewal

    International Nuclear Information System (INIS)

    Baers, B.; Pellionisz, P.

    1981-10-01

    The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

  13. 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

  14. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1983-01-01

    Purpose : To flatten the radial power distribution in the reactor core thereby improve the thermal performance of the reactor core by making the moderator-fuel ratio of fuel assemblies different depending on their position in the reactor core. Constitution : The volume of fuels disposed in the peripheral area of the reactor core is decreased by the increase of the volume of moderators in fuel assemblies disposed in the peripheral area of the reactor core to thereby make the moderator-fuel volume greater in the peripheral area than that in the central area. The moderator-fuel ratio adjustment is attained by making the number of water rods greater, decreasing the diameter of fuel pellets or decreasing the number of fuel pins in fuel assemblies disposed at the peripheral area of the reactor core as compared with fuel assemblies disposed at the central area of the reactor core. In this way, the infinite multiplication factors of fuels can be increased to thereby improve the reactor core performance. (Aizawa, K.)

  15. Prevention device for rapid reactor core shutdown in BWR type reactors

    International Nuclear Information System (INIS)

    Koshi, Yuji; Karatsu, Hiroyuki.

    1986-01-01

    Purpose: To surely prevent rapid shutdown of a nuclear reactor upon partial load interruption due to rapid increase in the system frequency. Constitution: If a partial load interruption greater than the sum of the turbine by-pass valve capacity and the load setting bias portion is applied in a BWR type power plant, the amount of main steams issued from the reactor is decreased, the thermal input/output balance of the reactor is lost, the reactor pressure is increased, the void is collapsed, the neutron fluxes are increased and the reactor power rises to generate rapid reactor shutdown. In view of the above, the turbine speed signal is compared with a speed setting value in a recycling flowrate control device and the recycling pump is controlled to decrease the recycling flowrate in order to compensate the increase in the neutron fluxes accompanying the reactor power up. In this way, transient changes in the reactor core pressure and the neutron fluxes are kept within a setting point for the rapid reactor shutdown operation thereby enabling to continue the plant operation. (Horiuchi, T.)

  16. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  17. High-temperature and breeder reactors - economic nuclear reactors of the future

    International Nuclear Information System (INIS)

    Djalilzadeh, A.M.

    1977-01-01

    The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

  18. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  19. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  20. Fast breeder reactors

    International Nuclear Information System (INIS)

    Heinzel, V.

    1975-01-01

    The author gives a survey of 'fast breeder reactors'. In detail the process of breeding, the reasons for the development of fast breeders, the possible breeder reactors, the design criteria, fuels, cladding, coolant, and safety aspects are reported on. Design data of some experimental reactors already in operation are summarized in stabular form. 300 MWe Prototype-Reactors SNR-300 and PFR are explained in detail and data of KWU helium-cooled fast breeder reactors are given. (HR) [de

  1. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  2. 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

  3. Fusion-fission hybrid reactors

    International Nuclear Information System (INIS)

    Greenspan, E.

    1984-01-01

    This chapter discusses the range of characteristics attainable from hybrid reactor blankets; blanket design considerations; hybrid reactor designs; alternative fuel hybrid reactors; multi-purpose hybrid reactors; and hybrid reactors and the energy economy. Hybrid reactors are driven by a fusion neutron source and include fertile and/or fissile material. The fusion component provides a copious source of fusion neutrons which interact with a subcritical fission component located adjacent to the plasma or pellet chamber. Fissile fuel and/or energy are the main products of hybrid reactors. Topics include high F/M blankets, the fissile (and tritium) breeding ratio, effects of composition on blanket properties, geometrical considerations, power density and first wall loading, variations of blanket properties with irradiation, thermal-hydraulic and mechanical design considerations, safety considerations, tokamak hybrid reactors, tandem-mirror hybrid reactors, inertial confinement hybrid reactors, fusion neutron sources, fissile-fuel and energy production ability, simultaneous production of combustible and fissile fuels, fusion reactors for waste transmutation and fissile breeding, nuclear pumped laser hybrid reactors, Hybrid Fuel Factories (HFFs), and scenarios for hybrid contribution. The appendix offers hybrid reactor fundamentals. Numerous references are provided

  4. 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)

  5. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  6. A multi-purpose reactor

    International Nuclear Information System (INIS)

    Changwen Ma

    2000-01-01

    An integrated natural circulation self pressurized reactor can be used for sea water desalination, electrogeneration, ship propulsion and district or process heating. The reactor can be used for ship propulsion because it has following advantages: it is a integrated reactor. Whole primary loop is included in a size limited pressure vessel. For a 200 MW reactor the diameter of the pressure vessel is about 5 m. It is convenient to arranged on a ship. Hydraulic driving facility of control rods is used on the reactor. It notably decreases the height of the reactor. For ship propulsion, smaller diameter and smaller height are important. Besides these, the operation reliability of the reactor is high enough, because there is no rotational machine (for example, circulating pump) in safety systems. Reactor systems are simple. There are no emergency water injection system and boron concentration regulating system. These features for ship propulsion reactor are valuable. Design of the reactor is based on existing demonstration district heating reactor design. The mechanic design principles are the same. But boiling is introduced in the reactor core. Several variants to use the reactor as a movable seawater desalination plant are presented in the paper. When the sea water desalination plant is working to produce fresh water, the reactor can supply electricity at the same time to the local electricity network. Some analyses for comprehensive application of the reactor have been done. Main features and parameters of the small (Thermopower 200 MW) reactor are given in the paper. (author)

  7. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  8. Supercritical Water Reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  9. Tank type reactor

    International Nuclear Information System (INIS)

    Otsuka, Fumio.

    1989-01-01

    The present invention concerns a tank type reactor capable of securing reactor core integrity by preventing incorporation of gases to an intermediate heat exchanger, thgereby improving the reliability. In a conventional tank type reactor, since vortex flows are easily caused near the inlet of an intermediate heat exchanger, there is a fear that cover gases are involved into the coolant main streams to induce fetal accidents. In the present invention, a reactor core is suspended by way of a suspending body to the inside of a reactor vessel and an intermediate heat exchanger and a pump are disposed between the suspending body and the reactor vessel, in which a vortex current preventive plate is attached at the outside near the coolant inlet on the primary circuit of the intermediate heat exchanger. In this way vortex or turbulence near the inlet of the intermediate heata exchanger or near the surface of coolants can be prevented. Accordingly, the cover gases are no more involved, to insure the reactor core integrity and obtain a tank type nuclear reactor of high reliability. (I.S.)

  10. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  11. Evaluation of Pressure Changes in HANARO Reactor Hall after a Reactor Shutdown

    International Nuclear Information System (INIS)

    Han, Geeyang; Han, Jaesam; Ahn, Gukhoon; Jung, Hoansung

    2013-01-01

    The major objective of this work is intended to evaluate the characteristics of the thermal behavior regarding how the decay heat will be affected by the reactor hall pressure change and the increase of pool water temperature induced in the primary coolant after a reactor shutdown. The particular reactor pool water temperature at the surface where it is evaporated owing to the decay heat resulting in the local heat transfer rate is related to the pressure change response in the reactor hall associated with the primary cooling system because of the reduction of the heat exchanger to remove the heat. The increase in the pool water temperature is proportional to the heat transfer rate in the reactor pool. Consequently, any limit on the reactor pool water temperature imposes a corresponding limit on the reactor hall pressure. At HANARO, the decay heat after a reactor shutdown is mainly removed by the natural circulation cooling in the reactor pool. This paper is written for the safety feature of the pressure change related leakage rate from the reactor hall. The calculation results show that the increase of pressure in the reactor hall will not cause any serious problems to the safety limits although the reactor hall pressure is slightly increased. Therefore, it was concluded that the pool water temperature increase is not so rapid as to cause the pressure to vary significantly in the reactor hall. Furthermore, the mathematical model developed in this work can be a useful analytical tool for scoping and parametric studies in the area of thermal transient analysis, with its proper representation of the interaction between the temperature and pressure in the reactor hall

  12. Design guide for Category III reactors: pool type reactors

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  13. 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

  14. 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

  15. Reactor safety method

    International Nuclear Information System (INIS)

    Vachon, L.J.

    1980-01-01

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature

  16. Small propulsion reactor design based on particle bed reactor concept

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  17. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    Baeten, Peter

    2006-01-01

    This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

  18. Generation III+ Reactor Portfolio

    International Nuclear Information System (INIS)

    2010-03-01

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

  19. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

    The IAEA's major first scientific publication is the Directory of Power Reactors now in operation or under construction in various parts of the world. The purpose of the directory is to present important details of various power projects in such a way as to provide a source of easy reference for anyone interested in the development of the peaceful uses of atomic energy, either at the technical or management level. Six pages have been devoted to each reactor the first of which contains general information, reactor physics data and information about the core. The second and third contain sketches of the fuel element or of the fuel element assembly, and of the horizontal and vertical sections of the reactor. On the fourth page information is grouped under the following heads: fuel element, core heat transfer, control, reactor vessel and over-all dimensions, and fluid flow. The fifth page shows a simplified flow diagram, while the sixth provides information on reflector and shielding, containment and turbo generator. Some information has also been given, when available, on cost estimates and operating staff requirements. Remarks and a bibliography constitute the last part of the description of each reactor. Reactor projects included in this directory are pressurized light water cooled power reactors. Boiling light water cooled power reactors, heavy water cooled power reactors, gas cooled power reactors, organic cooled power reactors liquid metal cooled power reactors and liquid metal cooled power reactors

  20. Analysis of dynamic stability and safety of reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

  1. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  2. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  3. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  4. Argentinean integrated small reactor design and scale economy analysis of integrated reactor

    International Nuclear Information System (INIS)

    Florido, P. C.; Bergallo, J. E.; Ishida, M. V.

    2000-01-01

    This paper describes the design of CAREM, which is Argentinean integrated small reactor project and the scale economy analysis results of integrated reactor. CAREM project consists on the development, design and construction of a small nuclear power plant. CAREM is an advanced reactor conceived with new generation design solutions and standing on the large experience accumulated in the safe operation of Light Water Reactors. The CAREM is an indirect cycle reactor with some distinctive and characteristic features that greatly simplify the reactor and also contribute to a highly level of safety: integrated primary cooling system, self pressurized, primary cooling by natural circulation and safety system relying on passive features. For a fully doupled economic evaluation of integrated reactors done by IREP (Integrated Reactor Evaluation Program) code transferred to IAEA, CAREM have been used as a reference point. The results shows that integrated reactors become competitive with power larger than 200MWe with Argentinean cheapest electricity option. Due to reactor pressure vessel construction limit, low pressure drop steam generator are used to reach power output of 200MWe for natural circulation. For forced circulation, 300MWe can be achieved. (author)

  5. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

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

  6. Reactor power measuring device

    International Nuclear Information System (INIS)

    Izumi, Mikio; Sano, Yuji; Seki, Eiji; Yoshida, Toshifumi; Ito, Toshiaki.

    1993-01-01

    The present invention provides a self-powered long detector having a sensitivity over the entire length of a reactor core as an entire control rod withdrawal range of a BWR type reactor, and a reactor power measuring device using a gamma ray thermometer which scarcely causes sensitivity degradation. That is, a hollow protection pipe is disposed passing through the reactor core from the outside of a reactor pressure vessel. The self-powered long detectors and the gamma ray thermometers are inserted and installed in the protection pipe. An average reactor power in an axial direction of the reactor relative to a certain position in the horizontal cross section of the reactor core is determined based on the power of the self-powered long detector over the entire length of the reactor core. Since the response of the self-powered detector relative to a local power change is rapid, the output is used as an input signal to a safety protection device of the reactor core. Further, a gamma ray thermometer secured in the reactor and having scarce sensitivity degradation is used instead of an incore travelling neutron monitor used for relative calibration of an existent neutron monitor secured in the reactor. (I.S.)

  7. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  9. Nuclear power reactors: reactor safety and military and civil defence

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

    The formation of fission products and plutonium in reactors is briefly described, followed by a short general discussion of reactor safety. The interaction of reactor safety and radioactive release considerations with military and civil defence is thereafter discussed. Reactors and other nuclear plants are factors which must be taken into account in the defence of the district around the site, and as potential targets of both conventional and guerilla attacks and sabotage, requiring special defence. The radiological hazards arising from serious damage to a power reactor by conventional weapons are briefly discussed, and the benefits of underground siting evaluated. Finally the author discusses the significance of the IAEA safeguards work as a preventive factor. (JIW)

  10. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

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

  11. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  12. The CEA research reactors

    International Nuclear Information System (INIS)

    Schwartz, J.P.

    1993-01-01

    Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

  13. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  14. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

    2008-01-01

    As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

  15. Reactor physics and reactor strategy investigations into the fissionable material economy of the thorium and uranium cycle in fast breeder reactors and high temperature reactors

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    In this work the properties governing the fissionable material economy of the uranium and thorium cycles are investigated for the advanced reactor types currently under development - the fast breeder reactor (FBR) and the high temperature reactor (HTR) - from the point of view of the optimum utilization of the available nuclear fuel reserves and the continuance of supply of these reserves. For this purpose, the two reactor types are first of all considered individually and are subsequently discussed as a complementary overall system

  16. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  18. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  19. Identification of nuclear reactor characteristics by the reactor noise analysis

    International Nuclear Information System (INIS)

    Yashima, Hideyuki

    1980-01-01

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

  20. Research on the reactor physics and reactor safety of VVER reactors. AER Symposium 2016

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S.

    2017-09-15

    The selected paperscan be attributed to the following main subjects: Reactor start-up tests and use of corresponding data for code validation, code development and application, approaches for safety analyses, closure of nuclear fuel cycle, prospective reactor concepts.

  1. The fast breeder reactor

    International Nuclear Information System (INIS)

    Collier, J.

    1990-01-01

    The arguments for and against the fast breeder reactor are debated. The case for the fast reactor is that the world energy demand will increase due to increasing population over the next forty years and that the damage to the global environment from burning fossil fuels which contribute to the greenhouse effect. Nuclear fission is the only large scale energy source which can achieve a cut in the use of carbon based fuels although energy conservation and renewable sources will also be important. Fast reactors produce more energy from uranium than other types of (thermal) reactors such as AGRs and PWRs. Fast reactors would be important from about 2020 onwards especially as by then many thermal reactors will need to be replaced. Fast reactors are also safer than normal reactors. The arguments against fast reactors are largely economic. The cost, especially the capital cost is very high. The viability of the technology is also questioned. (UK)

  2. 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.)

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

    International Nuclear Information System (INIS)

    Bussac, J.; Reuss, P.

    1985-01-01

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

  4. Ulysse, mentor reactor

    International Nuclear Information System (INIS)

    Bouquin, B.; Rio, I.; Safieh, J.

    1997-01-01

    On July 23, 1961, the ULYSSE reactor began its first power rise. Designed at that time to train nuclear engineering students and reactor operators, this reactor still remains an indispensable tool for nuclear teaching and a choice instrument for scientists. (author)

  5. Siting of research reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors

  6. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenaut, L.J.

    1972-01-01

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  7. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  8. Best practices for preparing vessel internals segmentation projects

    International Nuclear Information System (INIS)

    Boucau, Joseph; Segerud, Per; Sanchez, Moises

    2016-01-01

    Westinghouse has been involved in reactor internals segmentation activities in the U.S. and Europe for 30 years. Westinghouse completed in 2015 the segmentation of the reactor vessel and reactor vessel internals at the Jose Cabrera nuclear power plant in Spain and a similar project is on-going at Chooz A in France. For all reactor dismantling projects, it is essential that all activities are thoroughly planned and discussed up-front together with the customer. Detailed planning is crucial for achieving a successful project. One key activity in the preparation phase is the 'Segmentation and Packaging Plan' that documents the sequential steps required to segment, separate, and package each individual component, based on an activation analysis and component characterization study. Detailed procedures and specialized rigging equipment have to be developed to provide safeguards for preventing certain identified risks. The preparatory work can include some plant civil structure modifications for making the segmentation work easier and safer. Some original plant equipment is sometimes not suitable enough and need to be replaced. Before going to the site, testing and qualification are performed on full scale mock-ups in a specially designed pool for segmentation purposes. The mockup testing is an important step in order to verify the function of the equipment and minimize risk on site. This paper is describing the typical activities needed for preparing the reactor internals segmentation activities using under water mechanical cutting techniques. It provides experiences and lessons learned that Westinghouse has collected from its recent projects and that will be applied for the new awarded projects. (authors)

  9. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  10. Training reactor deployment. Advanced experimental course on designing new reactor cores

    International Nuclear Information System (INIS)

    Skoda, Radek

    2009-01-01

    Czech Technical University in Prague (CTU) operating its training nuclear reactor VR1, in cooperation with the North West University of South Africa (NWU), is applying for accreditation of the experimental training course ''Advanced experimental course on designing the new reactor core'' that will guide the students, young nuclear engineering professionals, through designing, calculating, approval, and assembling a new nuclear reactor core. Students, young professionals from the South African nuclear industry, face the situation when a new nuclear reactor core is to be build from scratch. Several reactor core design options are pre-calculated. The selected design is re-calculated by the students, the result is then scrutinized by the regulator and, once all the analysis is approved, physical dismantling of the current core and assembling of the new core is done by the students, under a close supervision of the CTU staff. Finally the reactor is made critical with the new core. The presentation focuses on practical issues of such a course, desired reactor features and namely pedagogical and safety aspects. (orig.)

  11. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).

  12. Three-dimensional reactor dynamics code for VVER type nuclear reactors

    International Nuclear Information System (INIS)

    Kyrki-Rajamaeki, R.

    1995-10-01

    A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)

  13. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  14. The fast reactor

    International Nuclear Information System (INIS)

    1980-02-01

    The subject is discussed as follows: brief description of fast reactors; advantage in conserving uranium resources; experience, in UK and elsewhere, in fast reactor design, construction and operation; safety; production of plutonium, security aspects; consideration of future UK fast reactor programme. (U.K.)

  15. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

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

  16. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  17. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  18. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  19. Request for Naval Reactors Comment on Proposed PROMETHEUS Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to Jet Propulsion Laboratory

    International Nuclear Information System (INIS)

    D. Kokkinos

    2005-01-01

    The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory

  20. Multi-purpose reactor

    International Nuclear Information System (INIS)

    1991-05-01

    The Multi-Purpose-Reactor (MPR), is a pool-type reactor with an open water surface and variable core arrangement. Its main feature is plant safety and reliability. Its power is 22MW t h, cooled by light water and moderated by beryllium. It has platetype fuel elements (MTR type, approx. 20%. enriched uranium) clad in aluminium. Its cobalt (Co 60 ) production capacity is 50000 Ci/yr, 200 Ci/gr. The distribution of the reactor core and associated control and safety systems is essentially based on the following design criteria: - upwards cooling flow, to waive the need for cooling flow inversion in case the reactor is cooled by natural convection if confronted with a loss of pumping power, and in order to establish a superior heat transfer potential (a higher coolant saturation temperature); - easy access to the reactor core from top of pool level with the reactor operating at full power, in order to facilitate actual implementation of experiments. Consequently, mechanisms associated to control and safety rods s,re located underneath the reactor tank; - free access of reactor personnel to top of pool level with the reactor operating at full power. This aids in the training of personnel and the actual carrying out of experiments, hence: - a vast water column was placed over the core to act as radiation shielding; - the core's external area is cooled by a downwards flow which leads to a decay tank beyond the pool (for N 16 to decay); - a small downwards flow was directed to stream downwards from above the reactor core in order to drag along any possibly active element; and - a stagnant hot layer system was placed at top of pool level so as to minimize the upwards coolant flow rising towards pool level

  1. Mirror reactor studies

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  2. Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code

    International Nuclear Information System (INIS)

    Shiba, T.; Fallot, M.

    2015-01-01

    To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)

  3. Reactor Sharing Program

    International Nuclear Information System (INIS)

    Tehan, Terry

    2002-01-01

    Support utilization of the RINSC reactor for student and faculty instructions and research. The Department of Energy award has provided financial assistance during the period 9/29/1995 to 5/31/2001 to support the utilization of the Rhode Island Nuclear Science Center (RINSC) reactor for student and faculty instruction and research by non-reactor owning educational institutions within approximately 300 miles of Narragansett, Rhode Island. Through the reactor sharing program, the RINSC (including the reactor and analytical laboratories) provided reactor services and laboratory space that were not available to the other universities and colleges in the region. As an example of services provided to the users: Counting equipment, laboratory space, pneumatic and in-pool irradiations, demonstrations of sample counting and analysis, reactor tours and lectures. Funding from the Reactor Sharing Program has provided the RINSC to expand student tours and demonstration programs that emphasized our long history of providing these types of services to the universities and colleges in the area. The funding have also helped defray the cost of the technical assistance that the staff has routinely provided to schools, individuals and researchers who have called on the RINSC for resolution of problems relating to nuclear science. The reactor has been featured in a Public Broadcasting System documentary on Pollution in the Arctic and how a University of Rhode Island Professor used Neutron Activation Analysis conducted at the RINSC to discover the sources of the ''Arctic Haze''. The RINSC was also featured by local television on Earth Day for its role in environmental monitoring

  4. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  5. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  6. Reactor control device

    International Nuclear Information System (INIS)

    Fukami, Haruo; Morimoto, Yoshinori.

    1981-01-01

    Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

  7. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

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

  8. 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)

  9. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-10-01

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

  11. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

    Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1980-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  13. Reactor scram device for FBR type reactor

    International Nuclear Information System (INIS)

    Kumasaka, Katsuyuki; Arashida, Genji; Itooka, Satoshi.

    1991-01-01

    In a control rod attaching structure in a reactor scram device of an FBR type reactor, an anti-rising mechanism proposed so far against external upward force upon occurrence of earthquakes relies on the engagement of a mechanical structure but temperature condition is not taken into consideration. Then, in the present invention, a material having curie temperature characteristics and which exhibits ferromagnetism only under low temperature condition and a magnet device are disposed to one of a movable control rod and a portion secured to the reactor. Alternatively, a bimetal member or a shape memory alloy which actuates to fix to the mating member only under low temperature condition is secured. The fixing device is adapted to operate so as to secure the control rods when the low temperature state is caused depending on the temperature condition. With such a constitution, when the control rods are separated from a driving device, they are prevented from rising even if they undergo external upward force due to earthquakes and so on, which can improve the reactor safety. (N.H.)

  14. 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

  15. Refuelling nuclear reactors

    International Nuclear Information System (INIS)

    Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

    1981-01-01

    An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

  16. Compilation of reactor physics data of the year 1984, AVR reactor

    International Nuclear Information System (INIS)

    Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

    1985-12-01

    The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

  17. Compilation of reactor physics data of the year 1983, AVR reactor

    International Nuclear Information System (INIS)

    Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

    1985-06-01

    The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

  18. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  19. Multipurpose research reactors

    International Nuclear Information System (INIS)

    1988-01-01

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

  20. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1978-10-01

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  1. An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

    NARCIS (Netherlands)

    Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

    2009-01-01

    In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

  2. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

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

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  3. Advances in reactor physics education: Visualization of reactor parameters

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  4. 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.)

  5. New reactor concepts

    International Nuclear Information System (INIS)

    Meskens, G.; Govaerts, P.; Baugnet, J.-M.; Delbrassine, A.

    1998-11-01

    The document gives a summary of new nuclear reactor concepts from a technological point of view. Belgium supports the development of the European Pressurized-Water Reactor, which is an evolutionary concept based on the European experience in Pressurized-Water Reactors. A reorientation of the Belgian choice for this evolutionary concept may be required in case that a decision is taken to burn plutonium, when the need for flexible nuclear power plants arises or when new reactor concepts can demonstrate proved benefits in terms of safety and cost

  6. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  7. Fission reactors and materials

    International Nuclear Information System (INIS)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions

  8. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  9. Method of operating a reactor

    International Nuclear Information System (INIS)

    Oosumi, Katsumi; Yamamoto, Michiyoshi.

    1980-01-01

    Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

  10. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  11. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  12. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  13. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

    1977-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

  14. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1984-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  15. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1976-09-01

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  16. Department of reactor technology

    International Nuclear Information System (INIS)

    1980-01-01

    The activities of the Department of Reactor Technology at Risoe during 1979 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  17. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  18. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  19. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  20. The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

    2001-07-01

    Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

  1. Canada-India Reactor (CIR)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1960-12-15

    Design information on the Canada-India Reactor is presented. Data are given on reactor physics, the core, fuel elements, core heat transfer, control, reactor vessel, fluid flow, reflector and shielding, containment, cost estimates, and research facilities. Drawings of vertical and horizontal sections of the reactor and fluid flow are included. (M.C.G.)

  2. 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

  3. One piece reactor removal

    International Nuclear Information System (INIS)

    Chia, Wei-Min; Wang, Song-Feng

    1993-01-01

    The strategy of Taiwan Research Reactor Renewal plan is to remove the old reactor block with One Piece Reactor Removal (OPRR) method for installing a new research reactor in original building. In this paper, the engineering design of each transportation works including the work method, the major equipments, the design policy and design criteria is described and discussed. In addition, to ensure the reactor block is safety transported for storage and to guarantee the integrity of reactor base mat is maintained for new reactor, operation safety is drawn special attention, particularly under seismic condition, to warrant safe operation of OPRR. ALARA principle and Below Regulatory Concern (BRC) practice were also incorporated in the planning to minimize the collective dose and the total amount of radioactive wastes. All these activities are introduced in this paper. (J.P.N.)

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  5. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

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

    International Nuclear Information System (INIS)

    1989-07-01

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

  7. Compact torsatron reactors

    International Nuclear Information System (INIS)

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

    1988-05-01

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

  8. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    2000-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  9. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1998-01-01

    Full text: In 1998 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  10. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

    Cupac, S.; Vukadin, Z.

    1996-01-01

    Full text: In 2000 Ra reactor was not operated. New instrumentation is not complete, without it, it is not possible to think about reactor start-up. Since 1985, when reactor operation was forbidden, there are 480 fuel elements left in 48 fuel channels in the reactor core. Heavy water was removed from the reactor core because of the repair of the heavy water pumps in 1986. The old instrumentation was removed. Eleven years after being left to its own destiny, it would be difficult to imagine that anybody would think of reactor restart without examining the state of reactor vessel and other vital reactor components. Maintaining the reactor under existing conditions without final decision about restart or permanent shutdown is destructive for this nuclear facility. The existing state that pertains for more than 10 years would have only one result, destruction of the RA reactor [sr

  11. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  12. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  13. Reactor operation method

    International Nuclear Information System (INIS)

    Osumi, Katsumi; Miki, Minoru.

    1979-01-01

    Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

  14. Nuclear reactor buildings

    International Nuclear Information System (INIS)

    Nagashima, Shoji; Kato, Ryoichi.

    1985-01-01

    Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

  15. Advanced reactor development

    International Nuclear Information System (INIS)

    Till, C.E.

    1989-01-01

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

  16. The reactor Cabri

    International Nuclear Information System (INIS)

    Ailloud, J.; Millot, J.P.

    1964-01-01

    It has become necessary to construct in France a reactor which would permit the investigation of the conditions of functioning of future installations, the choice, the testing and the development of safety devices to be adopted. A water reactor of a type corresponding to the latest CEA constructions in the field of laboratory or university reactors was decided upon: it appeared important to be able to evaluate the risks entailed and to study the possibilities of increasing the power, always demanded by the users; on the other hand, it is particularly interesting to clarify the phenomena of power oscillation and the risks of burn out. The work programme for CABRI will be associated with the work carried out on the American Sperts of the same type, during its construction, very useful contacts were made with the American specialists who designed the se reactors. A brief description of the reactor is given in the communication as well as the work programme for the first years with respect to the objectives up to now envisaged. Rough description of the reactor. CABRI is an open core swimming-pool reactor without any lateral protection, housed in a reinforced building with controlled leakage, in the Centre d'Etudes Nucleaires de Cadarache. It lies alone in the middle of an area whose radius is 300 meters long. Control and measurements equipment stand out on the edge of that zone. It consumes MTR fuel elements. The control-safety rods are propelled by compressed air. The maximum flow rate of cooling circuit is 1500 m 3 /h. Transient measurements are recorded in a RW330 unit. Aims and work programme. CABRI is meant for: - studies on the safety of water reactors - for the definition of the safety margins under working conditions: research of maximum power at which a swimming-pool reactor may operate with respect to a cooling accident, of local boiling effect on the nuclear behaviour of the reactor, performances of the control and safety instruments under exceptional

  17. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J.

    1982-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core where there is established a reactor coolant temperature set point at which it is desired to operate the reactor and first reactor coolant temperature band limits within which the set point is characterized. The reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in the core as the reactor coolant temperature approaches the first band limits to maintain the reactor coolant temperature near the set point and within the first band limits. The reactivity charges associated with movement of respective coolant displacer element clusters is calculated and compared with a calculated derived reactivity charge in order to select the cluster to be moved. (author)

  18. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  19. Innovative hybrid biological reactors using membranes; Reactores biologico hibrido innovadores utilizando membranas

    Energy Technology Data Exchange (ETDEWEB)

    Diez, R.; Esteban-Garcia, A. L.; Florio, L. de; Rodriguez-Hernandez, L.; Tejero, I.

    2011-07-01

    In this paper we present two lines of research on hybrid reactors including the use of membranes, although with different functions: RBPM, biofilm reactors and membranes filtration RBSOM, supported biofilm reactors and oxygen membranes. (Author) 14 refs.

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

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

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

  2. The Maple reactor project

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Labrie, J.-P.

    2003-01-01

    MDS Nordion supplies the majority of the world's reactor-produced medical isotopes. These isotopes are currently produced in the NRU reactor at AECL's Chalk River Laboratories (CRL). Medical isotopes and related technology are relied upon around the world to prevent, diagnose and treat disease. The NRU reactor, which has played a key role in supplying medical isotopes to date, has been in operation for over 40 years. Replacing this aging reactor has been a priority for MDS Nordion to assure the global nuclear medicine community that Canada will continue to be a dependable supplier of medical isotopes. MDS Nordion contracted AECL to construct two MAPLE reactors dedicated to the production of medical isotopes. The MDS Nordion Medical Isotope Reactor (MMIR) project started in September 1996. This paper describes the MAPLE reactors that AECL has built at its CRL site, and will operate for MDS Nordion. (author)

  3. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-15

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

  4. Methanogenesis in Thermophilic Biogas Reactors

    DEFF Research Database (Denmark)

    Ahring, Birgitte Kiær

    1995-01-01

    Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process as ....... Experiments using biogas reactors fed with cow manure showed that the same biogas yield found at 550 C could be obtained at 610 C after a long adaptation period. However, propionate degradation was inhibited by increasing the temperature.......Methanogenesis in thermophilic biogas reactors fed with different wastes is examined. The specific methanogenic activity with acetate or hydrogen as substrate reflected the organic loading of the specific reactor examined. Increasing the loading of thermophilic reactors stabilized the process...... as indicated by a lower concentration of volatile fatty acids in the effluent from the reactors. The specific methanogenic activity in a thermophilic pilot-plant biogas reactor fed with a mixture of cow and pig manure reflected the stability of the reactor. The numbers of methanogens counted by the most...

  5. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

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

    International Nuclear Information System (INIS)

    Ozawa, T.

    1974-01-01

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

  7. Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

    International Nuclear Information System (INIS)

    Bacher, P.

    1997-01-01

    A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

  8. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  9. Research reactor support

    International Nuclear Information System (INIS)

    2005-01-01

    Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its

  10. The prototype fast reactor

    International Nuclear Information System (INIS)

    Broomfield, A.M.

    1985-01-01

    The paper concerns the Prototype Fast Reactor (PFR), which is a liquid metal cooled fast reactor power station, situated at Dounreay, Scotland. The principal design features of a Fast Reactor and the PFR are given, along with key points of operating history, and health and safety features. The role of the PFR in the development programme for commercial reactors is discussed. (U.K.)

  11. NCSU Reactor Sharing Program

    International Nuclear Information System (INIS)

    Perez, P.B.

    1993-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities

  12. Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

    International Nuclear Information System (INIS)

    1975-06-01

    The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)

  13. 2012 review of French research reactors

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2013-01-01

    Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)

  14. Upgradation of Apsara reactor

    International Nuclear Information System (INIS)

    Mammen, S.; Mukherjee, P.; Bhatnagar, A.; Sasidharan, K.; Raina, V.K.

    2009-01-01

    Apsara is a 1 MW swimming pool type research reactor using high enriched uranium as fuel with light water as coolant and moderator. The reactor is in operation for more than five decades and has been extensively used for basic research, radioisotope production, neutron radiography, detector testing, shielding experiments etc. In view of its long service period, it is planned to carry out refurbishment of the reactor to extend its useful life. During refurbishment, it is also planned to upgrade the reactor to a 2 MW reactor to improve its utilization and to upgrade the structure, system and components in line with the current safety standards. This paper gives a brief account of the design features and safety aspects of the upgraded Apsara reactor. (author)

  15. Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

  16. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  17. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

  18. Materials for passively safe reactors

    International Nuclear Information System (INIS)

    Simnad, T.

    1993-01-01

    Future nuclear power capacity will be based on reactor designs that include passive safety features if recent progress in advanced nuclear power developments is realized. There is a high potential for nuclear systems that are smaller and easier to operate than the current generation of reactors, especially when passive or intrinsic characteristics are applied to provide inherent stability of the chain reaction and to minimize the burden on equipment and operating personnel. Taylor, has listed the following common generic technical features as the most important goals for the principal reactor development systems: passive stability, simplification, ruggedness, case of operation, and modularity. Economic competitiveness also depends on standardization and assurance of licensing. The performance of passively safe reactors will be greatly influenced by the successful development of advanced fuels and materials that will provide lower fuel-cycle costs. A dozen new designs of advanced power reactors have been described recently, covering a wide spectrum of reactor types, including pressurized water reactors, boiling water reactors, heavy-water reactors, modular high-temperature gas-cooled reactors (MHTGRs), and fast breeder reactors. These new designs address the need for passive safety features as well as the requirement of economic competitiveness

  19. Reactor power control device

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  20. The fast breeder reactor

    International Nuclear Information System (INIS)

    Davis, D.A.; Baker, M.A.W.; Hall, R.S.

    1990-01-01

    Following submission of written evidence, the Energy Committee members asked questions of three witnesses from the Central Electricity Generating Board and Nuclear Electric (which will be the government owned company running nuclear power stations after privatisation). Both questions and answers are reported verbatim. The points raised include where the responsibility for the future fast reactor programme should lie, with government only or with private enterprise or both and the viability of fast breeder reactors in the future. The case for the fast reactor was stated as essentially strategic not economic. This raised the issue of nuclear cost which has both a construction and a decommissioning element. There was considerable discussion as to the cost of building a European Fast reactor and the cost of the electricity it would generate compared with PWR type reactors. The likely demand for fast reactors will not arrive for 20-30 years and the need to build a fast reactor now is questioned. (UK)

  1. Method of reactor operation

    International Nuclear Information System (INIS)

    Maeda, Katsuji.

    1982-01-01

    Purpose: To prevent stress corrosion cracks in stainless steels caused from hydrogen peroxide in reactor operation in which the density of hydrogen peroxide in the reactor water is controlled upon reactor start-up. Method: A heat exchanger equipped with a heat source for applying external heat is disposed into the recycling system for reactor coolants. Upon reactor start-up, the coolants are heated by the heat exchanger till arriving at a temperature at which the dissolving rate is faster than the forming rate of hydrogen peroxide in the coolants, and nuclear heating is started after reaching the above temperature. The temperature of the reactor water is increased in such a manner and, when it arrives at 140 0 C, extraction of control elements is started and the heat source for the heat exchanger is interrupted simultaneously. In this way spikes in the density of hydrogen peroxide are suppressed upon reactor start-up to thereby decrease the stress corrosion cracks in stainless steels. (Horiuchi, T.)

  2. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  3. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

    Purpose: To significantly improve the working efficiency of a nuclear reactor by reflecting the control rod history effect on thermal variants required for the monitoring of the reactor operation. Constitution: An incore power distribution calculation section reads the incore neutron fluxes detected by neutron detectors disposed in the reactor to calculate the incore power distribution. A burnup degree distribution calculation section calculates the burnup degree distribution in the reactor based on the thus calculated incore power distribution. A control rod history date store device supplied with the burnup degree distribution renews the stored control rod history data based on the present control rod pattern and the burnup degree distribution. Then, thermal variants of the nuclear reactor are calculated based on the thus renewed control rod history data. Since the control rod history effect is reflected on the thermal variants required for the monitoring of the reactor operation, the working efficiency of the nuclear reactor can be improved significantly. (Seki, T.)

  4. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1975-11-01

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  5. Reactor materials research as an effective instrument of nuclear reactor perfection

    International Nuclear Information System (INIS)

    Baryshnikov, M.

    2006-01-01

    The work is devoted to reactor materiology, as to the practical tool of nuclear reactor development. The work is illustrated with concrete examples from activity experience of the appropriate division of the Russian Research Centre Kurchatov Institute - Institute of Reactor Materials Research and Radiation Nanotechnologies. Besides the description of some modern potentials of the mentioned institute is given. (author)

  6. Molten salt reactor concept

    International Nuclear Information System (INIS)

    Sood, D.D.

    1980-01-01

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

  7. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    International Nuclear Information System (INIS)

    Bonin, H.W.; Hilborn, J.W.; Carlin, G.E.; Gagnon, R.; Busatta, P.

    2014-01-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as 99 Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as 99 Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO 2 SO 4 ) with 994.2 g of 235 U (enrichment at 20%) providing an excess reactivity at operating temperature (40 o C) of 3.8 mk for a molality determined as 1.46 mol kg -1 for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 o C. Peak temperature and power were determined as 83 o C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the temperature and void coefficients are

  8. Homogeneous SLOWPOKE reactors for replacing SLOWPOKE-2 research reactors and the production of radioisotopes

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada); Hilborn, J.W. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Carlin, G.E. [Ontario Power Generation, Toronto, Ontario (Canada); Gagnon, R.; Busatta, P. [Canadian Forces (Canada)

    2014-07-01

    Inspired from the inherently safe SLOWPOKE-2 research reactor, the Homogeneous SLOWPOKE reactor was conceived with a double goal: replacing the heterogeneous SLOWPOKE-2 reactors when they reach end-of-core life to continue their missions of neutron activation analysis and neutron radiography at universities, and to produce radioisotopes such as {sup 99}Mo for medical applications. A homogeneous reactor core allows a much simpler extraction of radioisotopes (such as {sup 99}Mo) for applications in industry and nuclear medicine. The 20 kW Homogeneous SLOWPOKE reactor was modelled using both the deterministic WIMS-AECL and the probabilistic MCNP 5 reactor simulation codes. The homogeneous fuel mixture was a dilute aqueous solution of Uranyl Sulfate (UO{sub 2}SO{sub 4}) with 994.2 g of {sup 235}U (enrichment at 20%) providing an excess reactivity at operating temperature (40 {sup o}C) of 3.8 mk for a molality determined as 1.46 mol kg{sup -1} for a Zircaloy-2 reactor vessel. Because this reactor is intended to replace the core of SLOWPOKE-2 reactors, the Homogeneous SLOWPOKE reactor core had a height about twice its diameter. The reactor could be controlled by mechanical absorber rods in the beryllium reflector, chemical control in the core, or a combination of both. The safety of the Homogeneous SLOWPOKE reactor was analysed for both normal operation and transient conditions. Thermal-hydraulics calculations used COMSOL Multiphysics and the results showed that natural convection was sufficient to ensure adequate reactor cooling in all situations. The most severe transient simulated resulted from a 5.87 mk step positive reactivity insertion to the reactor in operation at critical and at steady state at 20 {sup o}C. Peak temperature and power were determined as 83 {sup o}C and 546 kW, respectively, reached 5.1 s after the reactivity insertion. However, the power fell rapidly to values below 20 kW some 35 s after the peak and remained below that value thereafter. Both the

  9. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Matsuura, S.; Nakahara, Y.; Takano, H.

    1983-09-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  10. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  11. Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

    International Nuclear Information System (INIS)

    Wittenbrock, N.G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

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

    International Nuclear Information System (INIS)

    Till, C.E.

    1990-01-01

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

  13. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  14. RB research reactor safety report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This new version of the safety report is a revision of the safety report written in 1962 when the RB reactor started operation after reconstruction. The new safety report was needed because reactor systems and components have been improved and the administrative procedures were changed. the most important improvements and changes were concerned with the use of highly enriched fuel (80% enriched), construction of reactor converter outside the reactor vessel, improved control system by two measuring start-up channels, construction of system for heavy water leak detection, new inter phone connection between control room and other reactor rooms. This report includes description of reactor building with installations, rector vessel, reactor core, heavy water system, control system, safety system, dosimetry and alarm systems, experimental channels, neutron converter, reactor operation. Safety aspects contain analyses of accident reasons, method for preventing reactivity insertions, analyses of maximum hypothetical accidents for cores with natural uranium, 2% enriched and 80% enriched fuel elements. Influence of seismic events on the reactor safety and well as coupling between reactor and the converter are parts of this document

  15. Digital control of research reactors

    International Nuclear Information System (INIS)

    Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

  16. LMFBR type reactor

    International Nuclear Information System (INIS)

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

    1988-01-01

    Purpose: To flatten the power distribution while maintaining the flattening in the axial power distribution in LMFBR type reactors. Constitution: Main system control rods are divided into control rods used for the operation and starting rods used for the starting of the reactor, and the starting rods are disposed in the radial periphery of the reactor core, while the control rods are disposed to the inside of the starting rods. With such a constitution, adjusting rods can be disposed in the region where the radial power peaking is generated to facilitate the flattening of the power distribution even in such a design that the ratio of the number of control rods to that of fuel assemblies is relatively large. That is, in this reactor, the radial power peaking is reduced by about 10% as compared with the conventional reactor core. As a result, the maximum linear power density during operation is reduced by about 10% to increase the thermal margin of the reactor core. If the maximum linear power density is set identical, the number of the fuel assemblies can be decreased by about 10%, to thereby reduce the fuel production cost. (K.M.)

  17. Fast breeder reactors

    International Nuclear Information System (INIS)

    Waltar, A.E.; Reynolds, A.B.

    1981-01-01

    This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of reactor theory, heat transfer, and fluid mechanics. The book is expected to be used most widely for a one-semester general course on fast breeder reactors, with the extent of material covered to vary according to the interest of the instructor. The book could also be used effectively for a two-quarter or a two-semester course. In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety. Methodology in fast reactor design and analysis, together with physical descriptions of systems, is emphasized in this text more than numerical results. Analytical and design results continue to change with the ongoing evolution of FBR design whereas many design methods have remained fundamentally unchanged for a considerable time

  18. BWR type reactor

    International Nuclear Information System (INIS)

    Watanabe, Shoichi

    1988-01-01

    Purpose: To inhibit the lowering of the neutron moderation effect due to voids in the upper portion of the reactor core, thereby flatten the axial power distribution. Constitution: Although it has been proposed to enlarge the diameter at the upper portion of a water rod thereby increasing the moderator in the upper portion, since the water rod situates within the channel box, the increase in the capacity thereof is has certain limit. In the present invention, it is designed such that the volume of the region at the outside of the channel box for the fuel assembly to which non-boiling water in the non-boiling water region can enter is made greater in the upper portion than in the lower portion of the reactor core. Thus, if the moderator density in the upper portion of the reactor core should be decreased due to the generation of the voids, the neutron moderating effect in the upper portion of the reactor core is not lowered as compared with the lower portion of the reactor core and, accordingly, the axial power distribution can be flattening more as compared with that in the conventional nuclear reactors. (Takahashi, M.)

  19. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

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

  20. Molten salt reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. Oklo natural reactor

    International Nuclear Information System (INIS)

    Fujii, Isao

    1985-01-01

    In 1954, Professor Kazuo, Kuroda of Arkansas University in USA published the possibility that spontaneously generated natural nuclear reactors existed in prehistoric age. In 1972, 18 years after that, Commissariat a l'Energie Atomique published that in the Oklo uranium deposit in Gabon, Africa, a natural nuclear reactor was found. This fact was immediately informed to the whole world, but in Japan, its details have not necessarily been well known. The chance of investigating into this fact and visiting the Oklo deposit by the favor of COMUF, the owner of the Oklo deposit, was given, therefore, the state of the natural reactors, which has been known so far, is reported. At present, 12 natural reactors have been found in the vicinity of the Oklo deposit. The natural reactors were generated spontaneously in uranium deposits about 1.7 billion years ago when the isotopic abundance of U-235 was 3 %, and the chain reaction started naturally. When the concentration of U-235 lowered, the reaction stopped naturally. The abnormality in the U-235 abundance in natural uranium was found, and the cause was pursued. The evidence of the existence of natural reactors was shown. (Kako, I.)

  2. Cross-flow electrochemical reactor cells, cross-flow reactors, and use of cross-flow reactors for oxidation reactions

    Science.gov (United States)

    Balachandran, Uthamalingam; Poeppel, Roger B.; Kleefisch, Mark S.; Kobylinski, Thaddeus P.; Udovich, Carl A.

    1994-01-01

    This invention discloses cross-flow electrochemical reactor cells containing oxygen permeable materials which have both electron conductivity and oxygen ion conductivity, cross-flow reactors, and electrochemical processes using cross-flow reactor cells having oxygen permeable monolithic cores to control and facilitate transport of oxygen from an oxygen-containing gas stream to oxidation reactions of organic compounds in another gas stream. These cross-flow electrochemical reactors comprise a hollow ceramic blade positioned across a gas stream flow or a stack of crossed hollow ceramic blades containing a channel or channels for flow of gas streams. Each channel has at least one channel wall disposed between a channel and a portion of an outer surface of the ceramic blade, or a common wall with adjacent blades in a stack comprising a gas-impervious mixed metal oxide material of a perovskite structure having electron conductivity and oxygen ion conductivity. The invention includes reactors comprising first and second zones seprated by gas-impervious mixed metal oxide material material having electron conductivity and oxygen ion conductivity. Prefered gas-impervious materials comprise at least one mixed metal oxide having a perovskite structure or perovskite-like structure. The invention includes, also, oxidation processes controlled by using these electrochemical reactors, and these reactions do not require an external source of electrical potential or any external electric circuit for oxidation to proceed.

  3. 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)

  4. Reactor calculations for improving utilization of TRIGA reactor

    International Nuclear Information System (INIS)

    Ravnik, M.

    1986-01-01

    A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

  5. Fusion Reactor Materials

    International Nuclear Information System (INIS)

    Decreton, M.

    2002-01-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

  6. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Carlson, G.A.; Moir, R.W.

    1978-01-01

    We have carried out conceptual design studies of fusion reactors based on the three current mirror confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fission fuel for fission reactors. We have designed a large commercial hybrid based on standard mirror confinement, and also a small pilot plant hybrid. Tandem mirror designs include a commercial 1000 MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single cell pilot plant

  7. Mirror fusion reactors

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    Conceptual design studies were made of fusion reactors based on the three current mirror-confinement concepts: the standard mirror, the tandem mirror, and the field-reversed mirror. Recent studies of the standard mirror have emphasized its potential as a fusion-fission hybrid reactor, designed to produce fuel for fission reactors. We have designed a large commercial hybrid and a small pilot-plant hybrid based on standard mirror confinement. Tandem mirror designs include a commercial 1000-MWe fusion power plant and a nearer term tandem mirror hybrid. Field-reversed mirror designs include a multicell commercial reactor producing 75 MWe and a single-cell pilot plant

  8. Advanced CANDU reactors

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  9. Experimental particle physics

    International Nuclear Information System (INIS)

    Steinberg, R.I.; Lane, C.E.

    1992-09-01

    The goals of this research are the experimental testing of fundamental theories of physics beyond the standard model and the exploration of cosmic phenomena through the techniques of particle physics. We are working on the MACRO experiment, which employs a large-area underground detector to search fore grand unification magnetic monopoles and dark matter candidates and to study cosmic ray muons as well as low- and high-energy neutrinos; the Chooz experiment to search for reactor neutrino oscillations at a distance of 1 km from the source; a new proposal (the Perry experiment) to construct a one-kiloton liquid scintillator in the Fairport, Ohio underground facility IMB to study neutrino oscillations with a 13 km baseline; and development of technology for improved liquid scintillators and for very-low-background materials in support of the MACRO and Perry experiments and for new solar neutrino experiments

  10. Lifetime management of the nuclear units in France

    International Nuclear Information System (INIS)

    Combes, J-P.; Godin, R.

    1994-01-01

    A systematic design study entitled 'Lifetime Project' has been initiated at Electricite de France, to estimate, plan, and maximize the life span of the French PWR plants. It is estimated that the present units will have a lifetime of 30 to 50 years. The life of a unit will be determined by that of its components, by economic considerations (whether it is cheaper to repair or replace the unit), and by safety considerations, which may be affected by changing safety standards. A 'periodic safety reassessment' takes place about every ten years. A list of 18 critical components can be summarized by saying that the main concerns are: radiation embrittlement of and within the reactor vessel, the steam generators, the concrete containment (which can not be replaced), instrumentation and control. Examination of samples from decommissioned plants, such as Chooz A, will provide valuable evidence of mechanisms of degradation due to aging

  11. Improvement of pressure-vessel surveillance of a PWR-power plant of the Societe d'Energie Nucleaire Franco-Belge des Ardennes (S.E.N.A.)

    International Nuclear Information System (INIS)

    Bevilacqua, A.; Lloret, R.; Riehl, R.

    1984-01-01

    This paper describes a new dosimetry, installed inside and outside the Pressure Vessel of CHOOZ Nuclear Power Plant of the Societe d'Energie Nucleaire Franco-Belge des Ardennes (S.E.N.A.), during its 1982-83 operation cycle. The inner dosimetry deals with a simulated capsule located under the reactor plate, and includes copper, nickel, iron, niobium, copper-cobalt, neptunium and uranium dosimeters. Its aim is to qualify the information given by the existing copper dosimetry. The spectrum used with these measurements is obtained by the 1 D ANISN Code and BIP-N 2 library. The outer dosimety is the fluence determination along the outer wall of the vessel. Two tubes, equiped by neutron dosimeters, seven meters long, were fixed along the vessel. On the median plane, the results are compared to a 2 D DOT transport calculation. Preliminary results are given which improve the vessel and specimens neutronic characterisation. (Auth.)

  12. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    International Nuclear Information System (INIS)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed

  13. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    Energy Technology Data Exchange (ETDEWEB)

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  14. Conceptual design of reactor assembly of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Selvaraj, A.; Balasubramaniyan, V.; Raghupathy, S.; Elango, D.; Sodhi, B.S.; Chetal, S.C.; Bhoje, S.B.

    1996-01-01

    The conceptual design of Reactor Assembly of 500 MWe Prototype Fast Breeder Reactor (as selected in 1985) was reviewed with the aim of 'simplification of design', 'Compactness of the reactor assembly' and 'ease in construction'. The reduction in size has been possible by incorporating concentric core arrangement, adoption of elastomer seals for Rotatable plugs, fuel handling with one transfer arm type mechanism, incorporation of mechanical sealing arrangement for IHX at the penetration in Inner vessel redan and reduction in number of components. The erection of the components has been made easier by adopting 'hanging' support for roof slab with associated changes in the safety vessel design. This paper presents the conceptual design of the reactor assembly components. (author). 8 figs, 2 tabs

  15. Fast breeder reactor research

    International Nuclear Information System (INIS)

    1975-01-01

    Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

  16. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  17. Reactor coolant pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  18. United Kingdom and USSR reactor types

    International Nuclear Information System (INIS)

    Lewins, Jeffery

    1988-01-01

    The features of the RBMK reactor operated at Chernobyl are compared with reactor types pertinent to the UK. The UK reactors covered are in three classes: the commercial reactors now built and operated or in commission (Magnox and Advanced Gas-cooled Reactor (AGR)); the prototype Steam Generating Heavy Water Reactor (SGHWR) and Prototype Fast Reactor (PFR) that have comparable performance to commercial reactors; and the proposed Pressurised Water Reactor (PWR) or Sizewell 'B' design which, it will be recollected, is different in detail from PWRs built elsewhere. We do not include research and test reactors nor the Royal Navy PWRs. The appendices explain resonances, Doppler and Xenon effects, the reactor physics of Chernobyl and positive void coefficients all of which are relevant to the comparisons. (author)

  19. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    International Nuclear Information System (INIS)

    Allen, Francis; Bonin, Hugues

    2008-01-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  20. Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

    2008-07-01

    The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

  1. Reactor-core-reactivity control device

    International Nuclear Information System (INIS)

    Miura, Teruo; Sakuranaga, Tomonobu.

    1983-01-01

    Purpose: To improve the reactor safety upon failures of control rod drives by adapting a control rod not to drop out accidentally from the reactor core but be inserted into the reactor core. Constitution: The control rod is entered or extracted as usual from the bottom of the pressure vessel. A space is provided above the reactor core within the pressure vessel, in which the moving scope of the control rod is set between the space above the reactor core and the reactor core. That is, the control rod is situated above the reactor core upon extraction thereof and, if an accident occurs to the control rod drive mechanisms to detach the control rod and the driving rod, the control rod falls gravitationally into the reactor core to improve the reactor safety. In addition, since the speed limiter is no more required to the control rod, the driving force can be decreased to reduce the size of the rod drive mechanisms. (Ikeda, J.)

  2. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  3. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-08-15

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

  4. 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

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  7. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

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

  8. Reactor water quality degradation suppressing method upon reactor start up

    International Nuclear Information System (INIS)

    Maeda, Katsuharu.

    1993-01-01

    Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

  9. Prometheus Project Reactor Module Final Report, For Naval Reactors Information

    International Nuclear Information System (INIS)

    MJ Wollman; MJ Zika

    2006-01-01

    The Naval Reactors Prime Contractor Team (NRPCT) led the development of a power plant for a civilian nuclear electric propulsion (NEP) system concept as part of the Prometheus Project. This report provides a summary of the facts, technical insights, and programmatic perspectives gained from this two-year program. The Prometheus Project experience has been extensively documented to better position the US for future space reactor development. Major Technological and engineering challenges exist to develop a system that provides useful electric power from a nuclear fission heat source operating in deep space. General issues include meeting mission requirements in a system that has a mass low enough to launch from earth while assuring public safety and remaining safely shutdown during credible launch accidents. These challenges may be overcome in the future if there is a space mission with a compelling need for nuclear power to drive development. Past experience and notional mission requirements indicate that any useful space reactor system will be unlike past space reactors and existing terrestrial reactors

  10. Repairing liner of the reactor; Reparacion del liner del reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-15

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

  11. Reactor of the XXI century

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.; Solov'ev, Yu.A.

    2001-01-01

    The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

  12. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  13. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  14. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  15. Trench reactor: an overview

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.; Sankoorikal, J.T.; Schmidt, R.S.; Lofshult, J.; Ramin, T.; Sokmen, N.; Lin, L.C.

    1988-01-01

    Recent fast, sodium-cooled reactor designs reflect new conditions. In nuclear energy these conditions are (a) emphasis on maintainability and operability, (b) design for more transparent safety, and (c) a surplus of uranium and enrichment availability that eases concerns about light water reactor fueling costs. In utility practice the demand is for less capital exposure, short construction time, smaller new unit sizes, and low capital cost. The PRISM, SAFR, and integral fast reactor (IFR) concepts are responses to these conditions. Fast reactors will not soon be deployed commercially, so more radical designs can be considered. The trench reactor is the product of such thinking. Its concepts are intended as contributions to the literature, which may be picked up by one of the existing programs or used in a new experimental project. The trench reactor is a thin-slab, pool-type reactor operated at very low power density and- for sodium-modest temperature. The thin slab is repeated in the sodium tank and the reactor core. The low power density permits a longer than conventional core height and a large-diameter fuel pin. Control is by borated steel slabs that can be lowered between the core and lateral sodium reflector. Shutdown is by semaphore slabs that can be swung into place just outside the control slabs. The paper presents major characteristics of the trench reactor that have been changed since the last report

  16. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.

  17. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-01

    This single page document is the October 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  18. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-10-15

    This single page document is the October 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  19. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-09-15

    This single page document is the September 15, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production Reactor.

  20. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-15

    This single page document is the December 16, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  1. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-12-01

    This single page document is the December 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the Production reactor.

  2. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  3. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

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

  4. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  5. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  6. Reactor FaceMap Tool: A modern graphics tool for displaying reactor data

    International Nuclear Information System (INIS)

    Roberts, J.C.

    1991-01-01

    A prominent graphical user interface in reactor physics applications at the Savannah River Site is the reactor facemap display. This is a two dimensional view of a cross section of a reactor. In the past each application which needed a facemap implemented its own version. Thus, none of the code was reused, the facemap implementation was hardware dependent and the user interface was different for each facemap. The Reactor FaceMap Tool was built to solve these problems. Through the use of modern computing technologies such as X Windows, object-oriented programming and client/server technology the Reactor FaceMap Tool has the flexibility to work in many diverse applications and the portability to run on numerous types of hardware

  7. continuous stirred tank reactor (CSTR)

    African Journals Online (AJOL)

    AFRICAN JOURNALS ONLINE (AJOL) · Journals · Advanced Search ... stirred tank reactor (CSTR) and the small and large intestines as plug flow reactor (PFR) ... from the two equations are used for the reactor sizing of the modeled reactors.

  8. Calibration of RB reactor power

    International Nuclear Information System (INIS)

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

    1976-09-01

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

  9. Reactor protecting device

    International Nuclear Information System (INIS)

    Ono, Hiroshi; Kasuga, Hajime; Kasuga, Hiroshi.

    1984-01-01

    Purpose: To reduce the recycling flowrate thereby decrease the neutron flux level before the reactor shutdown upon generation of abnormality such as increase in the neutron flux, by setting the safety level lower than the value for generating the reaction scram signal. Constitution: A netron flux safety level setter and an instruction signal generator are disposed between a neutron flux detector and a recycling flowrate control device. A neutron flux safety level lower than the level for generating a reactor scram signal and higher that the level for the ordinary operation is set and, if the detection level for the neutron flux in the reactor core arrives at the safety level, a neutron flux decreasing instruction signal is outputted from the instruction signal generator to the recycling flowrate control device to thereby decrease the recycling flowrate and decrease the neutron flux without reaching the reactor shutdown, whereby the thermal safety of the fuel rod can be maintained and the reactor operation performance can be improved. (Moriyama, K.)

  10. Reactor feedwater device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To suppress soluble radioactive corrosion products in a feedwater device. Method: In a light water cooled nuclear reactor, an iron injection system is connected to feedwater pipeways and the iron concentration in the feedwater or reactor coolant is adjusted between twice and ten times of the nickel concentration. When the nickel/iron ratio in the reactor coolant or feedwater goes nearer to 1/2, iron ions are injected together with iron particles to the reactor coolant to suppress the leaching of stainless steels, decrease the nickel in water and increase the iron concentration. As a result, it is possible to suppress the intrusion of nickel as one of parent nuclide of radioactive nuclides. Further, since the iron particles intruded into the reactor constitute nuclei for capturing the radioactive nuclides to reduce the soluble radioactive corrosion products, the radioactive nuclides deposited uniformly to the inside of the pipeways in each of the coolant circuits can be reduced. (Kawakami, Y.)

  11. Reactor physics computations

    International Nuclear Information System (INIS)

    Shapiro, A.

    1977-01-01

    Those reactor-core calculations which provide the effective multiplication factor (or eigenvalue) and the stationary (or fundamental mode) neutron-flux distribution at selected times during the lifetime of the core are considered. The multiplication factor is required to establish the nuclear composition and configuration which satisfy criticality and control requirements. The steady-state flux distribution must be known to calculate reaction rates and power distributions which are needed for the thermal, mechanical and shielding design of the reactor, as well as for evaluating refueling requirements. The calculational methods and techniques used for evaluating the nuclear design information vary with the type of reactor and with the preferences and prejudices of the reactor-physics group responsible for the calculation. Additionally, new methods and techniques are continually being developed and made operational. This results in a rather large conglomeration of methods and computer codes which are available for reactor analysis. The author provides the basic calculational framework and discusses the more prominent techniques which have evolved. (Auth.)

  12. Mirror reactor studies

    International Nuclear Information System (INIS)

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

    1976-01-01

    Design studies of a fusion mirror reactor, a fusion-fission mirror reactor, and two small mirror reactors are summarized. The fusion reactor uses 150-keV neutral-beam injectors based on the acceleration of negative ions. The injectors provide over 1 GW of continuous power at an efficiency greater than 80 percent. The fusion reactor has three-stage, modularized, Venetian blind, plasma direct converter with a predicted efficiency of 59 percent and a new concept for removal of the lune-shaped blanket: a crane is brought between the two halves of the Yin-Yang magnet, which are separated by a float. The design has desirable features such as steady-state operation, minimal impurity problems, and low first-wall thermal stress. The major disadvantage is low Q resulting in high recirculating power and hence high cost of electrical power. However, the direct capital cost per unit of gross electrical power is reasonable [$1000/kW(e)

  13. Reactor containing facility

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1992-01-01

    A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

  14. Test reactors in the world

    International Nuclear Information System (INIS)

    Corella, M.R.; Gomez Alonso, M.

    1983-01-01

    INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

  15. Research and materials irradiation reactors

    International Nuclear Information System (INIS)

    Ballagny, A.; Guigon, B.

    2004-01-01

    Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

  16. RB research reactor Safety Report

    International Nuclear Information System (INIS)

    Sotic, O.; Pesic, M.; Vranic, S.

    1979-04-01

    This RB reactor safety report is a revised and improved version of the Safety report written in 1962. It contains descriptions of: reactor building, reactor hall, control room, laboratories, reactor components, reactor control system, heavy water loop, neutron source, safety system, dosimetry system, alarm system, neutron converter, experimental channels. Safety aspects of the reactor operation include analyses of accident causes, errors during operation, measures for preventing uncontrolled activity changes, analysis of the maximum possible accident in case of different core configurations with natural uranium, slightly and highly enriched fuel; influence of possible seismic events

  17. Integral fast reactor

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1989-01-01

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

  18. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D; Takac, S; Markovic, H; Raisic, N; Zdravkovic, Z; Radanovic, Lj [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    In 1958 the zero energy reactor RB was built with the purpose of enabling critical experiments with various reactor systems to be carried out. The first core assembly built in this reactor consists of heavy water as moderator and natural uranium metal as fuel. In order to be able to obtain very accurate results when measuring the main characteristics of the assembly the reactor was built as a completely bare system. (author)

  19. Reactor utilization, Annex A

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1984-01-01

    Reactor was operated until August 1984 due to prohibition issued by the Ministry since the reactor does not have the emergency cooling system nor special filters in the ventilation system yet. This means that the operation plan was fulfilled by 69%. This annex includes detailed tables containing data about utilization of reactor experimental channels, irradiated samples, as well as interruptions of operation. Detailed data about reactor power during this period are shown as well

  20. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  1. Method of safely operating nuclear reactor

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro.

    1976-01-01

    Purpose: To provide a method of safely operating an nuclear reactor, comprising supporting a load applied to a reactor container partly with secondary container facilities thereby reducing the load borne by the reactor container when water is injected into the core to submerge the core in an emergency. Method: In a reactor emergency, water is injected into the reactor core thereby to submerge the core. Further, water is injected into a gap between the reactor container and the secondary container facilities. By the injection of water into the gap between the reactor container and the secondary container facilities a large apparent mass is applied to the reactor container, as a result of which the reactor container undergoes the same vibration as that of the secondary container facilities. Therefore, the load borne by the reactor container itself is reduced and stress at the bottom part of the reactor container is released. This permits the reactor to be operated more safely. (Moriyama, K.)

  2. Old and new ways in reactor technology. Reactor concepts and reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Schulten, R

    1989-01-01

    Compared to developments of other technical-scale systems, the period between the recognition of the underlying physics of nuclear fission and the development of a functioning nuclear reactor and its further development to the present level of maturity has been relatively short. The whole development is based on the chain reaction and is rendered safe by the possible auto-stabilization of this reaction. Consequently, the safety of nuclear reactors properly designed is based on automatic mechanisms, which prevent spreads of radioactivity even in major accidents. Controversial opinions about nuclear power uses are mostly based on wrong perceptions both of reactor safety and of radioactive waste, unless they are characterized by sheer ideology. The use of nuclear power worldwide has assumed an important, growing role in the combined uses of a variety energy sources in a surprisingly short period of time and will continue to make a safe, economic, and thus responsible contribution in the long run.

  3. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    1993-11-01

    The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

  4. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    International Nuclear Information System (INIS)

    Morrell, Douglas

    2011-01-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  5. Fusion Reactor Materials

    Energy Technology Data Exchange (ETDEWEB)

    Decreton, M

    2002-04-01

    The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  7. Test reactor: basic to U.S. breeder reactor development

    International Nuclear Information System (INIS)

    Miller, B.J.; Harness, A.J.

    1975-01-01

    Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

  8. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Furtek, A.

    2008-01-01

    Concerns over energy resources availability, climate changes and energy supply security suggest an important role for nuclear energy in future energy supplies. So far nuclear energy evolved through three generations and is still evolving into new generation that is now being extensively studied. Nuclear Power Plants are producing 16% of the world's electricity. Today the world is moving towards hydrogen economy. Nuclear technologies can provide energy to dissociate water into oxygen and hydrogen and to production of synthetic fuel from coal gasification. The introduction of breeder reactors would turn nuclear energy from depletable energy supply into an unlimited supply. From the early beginnings of nuclear energy in the 1940s to the present, three generations of nuclear power reactors have been developed: First generation reactors: introduced during the period 1950-1970. Second generation: includes commercial power reactors built during 1970-1990 (PWR, BWR, Candu, Russian RBMK and VVER). Third generation: started being deployed in the 1990s and is composed of Advanced LWR (ALWR), Advanced BWR (ABWR) and Passive AP600 to be deployed in 2010-2030. Future advances of the nuclear technology designs can broaden opportunities for use of nuclear energy. The fourth generation reactors are expected to be deployed by 2030 in time to replace ageing reactors built in the 1970s and 1980s. The new reactors are to be designed with a view of the following objectives: economic competitiveness, enhanced safety, minimal radioactive waste production, proliferation resistance. The Generation IV International Forum (GIF) was established in January 2000 to investigate innovative nuclear energy system concepts. GIF members include Argentina, Brazil, Canada, Euratom, France Japan, South Africa, South Korea, Switzerland, United Kingdom and United States with the IAEA and OECD's NEA as permanent observers. China and Russia are expected to join the GIF initiative. The following six systems

  9. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  10. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  11. 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)

  12. Research reactors in Austria - Present situation

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  13. 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)

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

    International Nuclear Information System (INIS)

    Chenais, Jacques

    2016-01-01

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

  15. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1975-02-01

    This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

  16. Nuclear reactors: physics and materials

    Energy Technology Data Exchange (ETDEWEB)

    Yadigaroglu, G

    2005-07-01

    In the form of a tutorial addressed to non-specialists, the article provides an introduction to nuclear reactor technology and more specifically to Light Water Reactors (LWR); it also shows where materials and chemistry problems are encountered in reactor technology. The basics of reactor physics are reviewed, as well as the various strategies in reactor design and the corresponding choices of materials (fuel, coolant, structural materials, etc.). A brief description of the various types of commercial power reactors follows. The design of LWRs is discussed in greater detail; the properties of light water as coolant and moderator are put in perspective. The physicochemical and metallurgical properties of the materials impose thermal limits that determine the performance and the maximum power a reactor can deliver. (author)

  17. CER. Research reactors in France

    International Nuclear Information System (INIS)

    Estrade, Jerome

    2012-01-01

    Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)

  18. Development task of compact reactor

    International Nuclear Information System (INIS)

    Kurushima, Morihiro

    1982-01-01

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

  19. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    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 design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

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

    International Nuclear Information System (INIS)

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

    1978-05-01

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

  1. 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.)

  2. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M. [Nuclear Science Program, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  3. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

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

  4. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  5. Fundamentals of reactor chemistry

    International Nuclear Information System (INIS)

    Akatsu, Eiko

    1981-12-01

    In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around the nuclear reactors. The curricula of the courses contain also the subject material of chemistry. With reference to the foreign curricula, a plan of educational subject material of chemistry in the Nuclear Engineering School of JAERI was considered, and the fundamental part of reactor chemistry was reviewed in this report. Since the students of the Nuclear Engineering School are not chemists, the knowledge necessary in and around the nuclear reactors was emphasized in order to familiarize the students with the reactor chemistry. The teaching experience of the fundamentals of reactor chemistry is also given. (author)

  6. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  7. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  8. Research reactor`s role in Korea

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

  9. Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-10-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

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

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

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

  11. Directory of Nuclear Research Reactors 1994

    International Nuclear Information System (INIS)

    1995-08-01

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

  12. Licensing of nuclear reactor operators

    International Nuclear Information System (INIS)

    1979-09-01

    Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

  13. Operational safety and reactor life improvements of Kyoto University Reactor

    International Nuclear Information System (INIS)

    Utsuro, M.; Fujita, Y.; Nishihara, H.

    1990-01-01

    Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

  14. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  15. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  16. Reactor container cooling device

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1995-11-10

    The device of the present invention efficiently lowers pressure and temperature in a reactor container upon occurrence of a severe accident in a BWR-type reactor and can cool the inside of the container for a long period of time. That is, (1) pipelines on the side of an exhaustion tower of a filter portion in a filter bent device of the reactor container are in communication with pipelines on the side of a steam inlet of a static container cooling device by way of horizontal pipelines, (2) a back flow check valve is disposed to horizontal pipelines, (3) a steam discharge valve for a pressure vessel is disposed closer to the reactor container than the joint portion between the pipelines on the side of the steam inlet and the horizontal pipelines. Upon occurrence of a severe accident, when the pressure vessel should be ruptured and steams containing aerosol in the reactor core should be filled in the reactor container, the inlet valve of the static container cooling device is closed. Steams are flown into the filter bent device of the reactor container, where the aerosols can be removed. (I.S.).

  17. The integral fast reactor

    International Nuclear Information System (INIS)

    Till, C.E.

    1987-01-01

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

  18. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  19. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

    1983-01-01

    Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

  20. RA Reactor applications, Annex A

    International Nuclear Information System (INIS)

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

    1990-01-01

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