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)

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.

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.

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.

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)

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.)

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.)

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.

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.)

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}.

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.

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.)

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

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)

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.

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

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

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

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

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.

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.

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

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

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

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

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.

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

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

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).

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.

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.

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...

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.)

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)

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

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

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.

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)

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)

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

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)

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

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

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).

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

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

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

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

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.

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)

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-01-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testings of CENDL-2 and ENDF/B-6. (4 tabs., 2 figs.)

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

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

Homogeneous fast reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1995-11-01

How to choose correct weighting spectrum has been studied to produce multigroup constants for fast reactor benchmark calculations. A correct weighting option makes us obtain satisfying results of K eff and central reaction rate ratios for nine fast reactor benchmark testing of CENDL-2 and ENDF/B-6. (author). 8 refs, 2 figs, 4 tabs

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.

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)

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.)

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

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.

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)

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

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}.

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)

The performance of ENDF/B-V.2 nuclear data for fast reactor calculations

International Nuclear Information System (INIS)

Atkinson, C.A.; Collins, P.J.

1987-01-01

Calculations with ENDF/B-V.2 data have been made for twenty-five fast-spectrum integral assemblies covering a wide range of sizes and compositions. Analysis was done by transport codes with refined cross section processing methods and detailed reactor modelling. The predictions of fission rate distributions and control rod worths were emphasized for the more prototypic benchmark cores. The results show considerable improvements in agreement with experiment compared with analysis using ENDF/B-IV data, but it is apparent that significant errors remain for fast reactor design calculations

MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

Energy Technology Data Exchange (ETDEWEB)

GERBER MS

2009-04-28

The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

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

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.

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

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.)

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

Hanford B Reactor Building Hazard Assessment Report

International Nuclear Information System (INIS)

Griffin, P. W.

1999-01-01

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

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)

Reactor pressure vessel steels ASTM A533B and A508 Cl.2

International Nuclear Information System (INIS)

Pelli, R.; Kemppainen, M.; Toerroenen, K.

1979-11-01

This report presents the tensile test results of steels ASTM A533B and A508 Cl.2 obtained in connection with a programme initiated to gather and create information needed for the assessment of the structural integrity of the reactor pressure vessels. The tensile properties were studied between -196 and 300 degC varying austenitizing and tempering temperatures and having two different carbon contents for the heats of A533B. (author)

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

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.

105-B Reactor museum feasibility assessment (Phase 2) project

International Nuclear Information System (INIS)

Heckel, R. P.

2000-01-01

This 105-B Reactor Museum feasibility assessment project report documents project activities that have been performed, including a review and assessment of previously existing information, a walk-through of the facility, an assessment of potential hazards, and selection of mitigative measures deemed to be appropriate to allow unescorted access by members of the public to a specified primary tour route

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.

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.

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)

PARR-2: reactor description and experiments

International Nuclear Information System (INIS)

Wyne, M.F.; Meghji, J.H.

1990-12-01

PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

Validation study of the reactor physics lattice transport code WIMSD-5B by TRX and BAPL critical experiments of light water reactors

International Nuclear Information System (INIS)

Khan, M.J.H.; Alam, A.B.M.K.; Ahsan, M.H.; Mamun, K.A.A.; Islam, S.M.A.

2015-01-01

Highlights: • To validate the reactor physics lattice code WIMSD-5B by this analysis. • To model TRX and BAPL critical experiments using WIMSD-5B. • To compare the calculated results with experiment and MCNP results. • To rely on WIMSD-5B code for TRIGA calculations. - Abstract: The aim of this analysis is to validate the reactor physics lattice transport code WIMSD-5B by TRX (thermal reactor-one region lattice) and BAPL (Bettis Atomic Power Laboratory-one region lattice) critical experiments of light water reactors for neutronics analysis of 3 MW TRIGA Mark-II research reactor at AERE, Dhaka, Bangladesh. This analysis is achieved through the analysis of integral parameters of five light water reactor critical experiments TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 based on evaluated nuclear data libraries JEFF-3.1 and ENDF/B-VII.1. In integral measurements, these experiments are considered as standard benchmark lattices for validating the reactor physics lattice transport code WIMSD-5B as well as evaluated nuclear data libraries. The integral parameters of the said critical experiments are calculated using the reactor physics lattice transport code WIMSD-5B. The calculated integral parameters are compared to the measured values as well as the earlier published MCNP results based on the Chinese evaluated nuclear data library CENDL-3.0 for assessment of deterministic calculation. It was found that the calculated integral parameters give mostly reasonable and globally consistent results with the experiment and the MCNP results. Besides, the group constants in WIMS format for the isotopes U-235 and U-238 between two data files have been compared using WIMS library utility code WILLIE and it was found that the group constants are well consistent with each other. Therefore, this analysis reveals the validation study of the reactor physics lattice transport code WIMSD-5B based on JEFF-3.1 and ENDF/B-VII.1 libraries and can also be essential to

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.

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.

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)

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)

Theoretical study for ICRF sustained LHD type p-11B reactor

International Nuclear Information System (INIS)

Watanabe, Tsuguhiro

2003-04-01

This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p- 11 B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D- 3 He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p- 11 B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D- 3 He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor (γ HH ). It is shown that high average beta plasma confinement, a large confinement factor (γ HH > 3) and the hot ion mode (T i /T e > 1.4) are necessary to achieve the ignition of the D- 3 He helical reactor. Characteristics of ICRF sustained p- 11 B reactor are analyzed in section 4. The nuclear fusion reaction rate is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p- 11 B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

Core monitoring at the WNP-2 reactor

International Nuclear Information System (INIS)

Skeen, D.R.; Torres, R.H.; Burke, W.J.; Jenkins, I.; Jones, S.W.

1992-01-01

The WNP-2 reactor is a 3,323-MW(thermal) boiling water reactor (BWR) that is operated by the Washington Public Power Supply System. The WNP-2 reactor began commercial operation in 1984 and is currently in its eighth cycle. The core monitoring system used for the first cycle of operation was supplied by the reactor vendor. Cycles 2 through 6 were monitored with the POWERPLEX Core Monitoring Software System (CMSS) using the XTGBWR simulation code. In 1991, the supply system upgraded the core monitoring system by installing the POWERPLEX 2 CMSS prior to the seventh cycle of operation for WNP-2. The POWERPLEX 2 CMSS was developed by Siemens Power Corporation (SPC) and is based on SPC's advanced state-of-the-art reactor simulator code MICROBURN-B. The improvements in the POWERPLEX 2 system are possible as a result of advances in minicomputer hardware

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

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.)

G 2 reactor project; Projet de pile a double fin: G 2

Energy Technology Data Exchange (ETDEWEB)

Ailleret, [Electricite de France (EDF), Dir. General des Etudes de Recherches, 75 - Paris (France); Taranger, P; Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA actually constructs the G-2 reactor core working with natural uranium, which will use graphite as moderator, and gas under pressure as cooling fluid. This report presents the specificity of the new reactor: - the different elements of the reactor core, - the control and the security of the reactor, - the renewal of the fuel, - the biologic surrounding wall, - and the cooling circuit. (M.B.) [French] le Commissariat a l'Energie Atomique construit actuellement la pile G-2 a Uranium naturel, qui utilisera le graphite comme moderateur, et le gaz sous pression comme fluide de refroidissement. Ce rapport presente les specificite du nouveau reacteur: - les differents elements de la pile, - le controle et la securite du reacteur, - le renouvellement du combustible, - l'enceinte biologique, - et le circuit de refroidissement. (M.B.)

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

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

Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Tsuguhiro (ed.)

2003-04-01

This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

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.

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.

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

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

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

Active species in a large volume N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Kutasi, K; Pintassilgo, C D; Loureiro, J; Coelho, P J

2007-01-01

A large volume post-discharge reactor placed downstream from a flowing N 2 -O 2 microwave discharge is modelled using a three-dimensional hydrodynamic model. The density distributions of the most populated active species present in the reactor-O( 3 P), O 2 (a 1 Δ g ), O 2 (b 1 Σ g + ), NO(X 2 Π), NO(A 2 Σ + ), NO(B 2 Π), NO 2 (X), O 3 , O 2 (X 3 Σ g - ) and N( 4 S)-are calculated and the main source and loss processes for each species are identified for two discharge conditions: (i) p = 2 Torr, f = 2450 MHz, and (ii) p = 8 Torr, f = 915 MHz; in the case of a N 2 -2%O 2 mixture composition and gas flow rate of 2 x 10 3 sccm. The modification of the species relative densities by changing the oxygen percentage in the initial gas mixture composition, in the 0.2%-5% range, are presented. The possible tuning of the species concentrations in the reactor by changing the size of the connecting afterglow tube between the active discharge and the large post-discharge reactor is investigated as well

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)

Plans for use of ENDF/B in reactor research in Indonesia

International Nuclear Information System (INIS)

Santoso, B.; Syaukat, A.; Subki, I.; Ganesan, S.

1989-07-01

Nuclear data are numerical constants of nature which quantify the nuclear behaviour of all elements and isotopes which make up the reactor medium and its environment, and which are needed as input for performing design calculations for safe and reliable operation of nuclear reactors. The nuclear data are available in the form of recommended values in specially formatted computerized files such as the Evaluated Nuclear Data File-B, known as ENDF/B. The development of base technology in the scheme of original reactor design calculations involves the mastering of the art of ENDF/B data processing. This paper briefly discusses the current status of this activity in Jakarta and gives an account of the future plans, with emphasis on the role of ENDF/B in reactor calculations. (author). 15 refs, 9 figs

Study of parameters affecting the conversion in a plug flow reactor for reactions of the type 2A→B

Science.gov (United States)

Beltran-Prieto, Juan Carlos; Long, Nguyen Huynh Bach Son

2018-04-01

Modeling of chemical reactors is an important tool to quantify reagent conversion, product yield and selectivity towards a specific compound and to describe the behavior of the system. Proposal of differential equations describing the mass and energy balance are among the most important steps required during the modeling process as they play a special role in the design and operation of the reactor. Parameters governing transfer of heat and mass have a strong relevance in the rate of the reaction. Understanding this information is important for the selection of reactor and operating regime. In this paper we studied the irreversible gas-phase reaction 2A→B. We model the conversion that can be achieved as function of the reactor volume and feeding temperature. Additionally, we discuss the effect of activation energy and the heat of reaction on the conversion achieved in the tubular reactor. Furthermore, we considered that dimerization occurs instantaneously in the catalytic surface to develop equations for the determination of rate of reaction per unit area of three different catalytic surface shapes. This data can be combined with information about the global rate of conversion in the reactor to improve regent conversion and yield of product.

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.

Benchmark testing of Canadol-2.1 for heavy water reactor

International Nuclear Information System (INIS)

Liu Ping

1999-01-01

The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

IRPHE/B and W-SS-LATTICE, Spectral Shift Reactor Lattice Experiments

International Nuclear Information System (INIS)

2003-01-01

Description: B and W has performed and analysed a series of physics experiments basically concerned with the technology of heterogeneous reactors moderated and cooled by a variable mixture of heavy and light water. A reactor so moderated is termed Spectral Shift Control Reactor (S SCR). In the practical application of this concept, the moderator mixture is rich in heavy water at the beginning of core life, so a relatively large fraction of the neutrons are epithermal and are absorbed in the fertile material. As fuel is consumed, the moderator is diluted with light water. In this way the neutron spectrum is shifted, thereby increasing the proportion of thermal neutrons and the reactivity of the system. The general objective of the S SCR Basic Physics Program was to study the nuclear properties of rod lattices moderated by D 2 O-H 2 O mixtures. The volume ratio of moderator to non-moderator in all lattices was approximately 1.0, and the fuel was either 4%-enriched UO 2 clad in stainless steel or 93%-enriched UO 2 -ThO 2 (Nth/N 15) pellets clad in aluminum. The D 2 O concentration in the moderator ranged from zero to about 90 mole %. The experimental program includes critical experiments with both types of fuel, exponential experiments at room temperature with both types of fuel, exponential experiments at elevated temperatures with the 4%-enriched UO 2 fuel, and neutron age measurements in ThO 2 lattices. The theoretical program included the development of calculation methods applicable to these systems, and the analysis and correlation of the experimental data. A first report provides the results of critical experiments performed under the Spectral Shift Control Reactor Basic Physics Program. A second report documents experimental results and theoretical interpretation of a series of twenty uniform lattice critical experiments in which the neutron spectrum is varied over a fairly broad range. A third report addresses issues that bear on the problems associated with

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

IAEA Concludes Safety Review at Chooz Nuclear Power Plant in France

International Nuclear Information System (INIS)

2013-01-01

management of corrective maintenance and leak repair to minimize backlogs and maintain plant safety; The plant should enhance the chemistry control program to cover all chemistry aspects of plant systems; The plant should enhance the process of root cause analysis of safety significant events to improve the depth of analysis of such events; and The plant should consider improving systems to allow clear identification of deficiencies in the field once they have been recognized. Chooz management expressed a determination to address all the areas identified for improvement and requested the IAEA to schedule a follow-up mission in approximately 18 months. The team delivered a draft of its recommendations, suggestions and good practices to the plant management in the form of Technical Notes for factual comments. The technical notes and comments from the plant and the French Safety Authority will be reviewed at IAEA headquarters. The final report will be submitted to the Government of France within three months. This was the 175th mission of the OSART programme, which began in 1982, and the 25th mission in France. The next OSART mission in France is scheduled at Flamanville Units 1 and 2 in 2014. Background: General information about OSART missions can be found on the IAEA Website. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the plant's overall safety status. Experts participating in the IAEA's June 2010 International Conference on Operational Safety of Nuclear Power Plants (NPP) reviewed the experience of the OSART programme and concluded: In OSART missions NPPs are assessed against IAEA Safety Standards which reflect the current international consensus on what constitutes a high level of safety; and OSART recommendations and suggestions are of utmost importance for operational safety improvement of NPPs. The IAEA

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

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

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

Testing ENDF/B-V data for thermal reactors

International Nuclear Information System (INIS)

Craig, D.S.

1982-10-01

Lattice parameters have been calculated for some thermal reactor benchmark lattices using ENDF/B-V data. These lattices were TRX-1, -2; BAPL-UO 2 -1,-2,-3; BNL-ThO 2 - 233 UO 2 -H 2 0-1,-2,-3; MIT-4,-5,-6; and PNL-31,-33,-35 (infinite lattices). In addition, parameters were calculated for 3 ZEEP lattices, 3 High-Conversion U0 2 -H 2 0 lattices, and 7 BNL-Th0 2 - 233 U0 2 -D 2 0 lattices. These calculations were made using the integral transport cell code RAHAB with the resonance reaction rates obtained using the OZMA code operating in the discrete ordinate mode. This code calculates the resonance rates allowing for the interaction of all resonances. Four group reaction rates for use in method comparisons are given for several lattices. The author discusses the use of the OZMA code for these calculations, including the choice of options and the orders of the angular quadratures, and compares results obtained using the CRNL thermal scattering data with those obtained using ENDF/B data

Depleted Reactor Analysis With MCNP-4B

International Nuclear Information System (INIS)

Caner, M.; Silverman, L.; Bettan, M.

2004-01-01

Monte Carlo neutronics calculations are mostly done for fresh reactor cores. There is today an ongoing activity in the development of Monte Carlo plus burnup code systems made possible by the fast gains in computer processor speeds. In this work we investigate the use of MCNP-4B for the calculation of a depleted core of the Soreq reactor (IRR-1). The number densities as function of burnup were taken from the WIMS-D/4 cell code calculations. This particular code coupling has been implemented before. The Monte Carlo code MCNP-4B calculates the coupled transport of neutrons and photons for complicated geometries. We have done neutronics calculations of the IRR-1 core with the WIMS and CITATION codes in the past Also, we have developed an MCNP model of the IRR-1 standard fuel for a criticality safety calculation of a spent fuel storage pool

Reactor BR2

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2000-07-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported.

Reactor BR2

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. Various aspects concerning the operation of the BR2 Reactor, the utilisation of the CALLISTO loop and the irradiation programme, the BR2 R and D programme and the production of isotopes and of NTD-silicon are discussed. Progress and achievements in 1999 are reported

Proposed nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel: Appendix B, foreign research reactor spent nuclear fuel characteristics and transportation casks. Volume 2

International Nuclear Information System (INIS)

1995-03-01

This is Appendix B of a draft Environmental Impact Statement (EIS) on a Proposed Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. It discusses relevant characterization and other information of foreign research reactor spent nuclear fuel that could be managed under the proposed action. It also discusses regulations for the transport of radioactive materials and the design of spent fuel casks

Set of rules SOR 2 reactor site criteria

International Nuclear Information System (INIS)

1976-06-01

The purpose of this set of rules is to describe criteria which guide the Director in his evaluation of the suitability of proposed sites for stationary power and testing reactors subject to SOR 2. (B.G.)

Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-08-15

The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

Doping-Induced Isotopic Mg11B2 Bulk Superconductor for Fusion Application

Directory of Open Access Journals (Sweden)

Qi Cai

2017-03-01

Full Text Available Superconducting wires are widely used for fabricating magnetic coils in fusion reactors. Superconducting magnet system represents a key determinant of the thermal efficiency and the construction/operating costs of such a reactor. In consideration of the stability of 11B against fast neutron irradiation and its lower induced radioactivation properties, MgB2 superconductor with 11B serving as the boron source is an alternative candidate for use in fusion reactors with a severe high neutron flux environment. In the present work, the glycine-doped Mg11B2 bulk superconductor was synthesized from isotopic 11B powder to enhance the high field properties. The critical current density was enhanced (103 A·cm−2 at 20 K and 5 T over the entire field in contrast with the sample prepared from natural boron.

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.).

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

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.

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...

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)

Fast-reactor-data testing of ENDF/B-V at ORNL

International Nuclear Information System (INIS)

Wright, R.Q.; Ford, W.E. III; Lucius, J.L.; Webster, C.C.; Marable, J.H.

1982-01-01

The Cross Section Evaluation Working Group (CSEWG) is coordinating a program to assess the adequacy of ENDF/B-V cross sections for both fast- and thermal-reactor design applications. A secondary goal is to evaluate cross-section processing codes, cross-section libraries, and radiation-transport codes. Fast reactor data testing (FRDT) goals are accomplished, in part, by comparison of calculated results with documented performance parameters of CSEWG fast reactor benchmarks and with results obtained by other data testers. The purpose of this paper is to describe the results of FRDT at Oak Ridge National Laboratory

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.).

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.).

Combustion of Na2B4O7 + Mg + C to synthesis B4C powders

International Nuclear Information System (INIS)

Jiang Guojian; Xu Jiayue; Zhuang Hanrui; Li Wenlan

2009-01-01

Boron carbide powder was fabricated by combustion synthesis (CS) method directly from mixed powders of borax (Na 2 B 4 O 7 ), magnesium (Mg) and carbon. The adiabatic temperature of the combustion reaction of Na 2 B 4 O 7 + 6 Mg + C was calculated. The control of the reactions was achieved by selecting reactant composition, relative density of powder compact and gas pressure in CS reactor. The effects of these different influential factors on the composition and morphologies of combustion products were investigated. The results show that, it is advantageous for more Mg/Na 2 B 4 O 7 than stoichiometric ratio in Na 2 B 4 O 7 + Mg + C system and high atmosphere pressure in the CS reactor to increase the conversion degree of reactants to end product. The final product with the minimal impurities' content could be fabricated at appropriate relative density of powder compact. At last, boron carbide without impurities could be obtained after the acid enrichment and distilled water washing.

Preliminary Hazard Classification for the 105-B Reactor

International Nuclear Information System (INIS)

Kerr, N.R.

1997-08-01

This document summarizes the inventories of radioactive and hazardous materials present within the 105-B Reactor and uses the inventory information to determine the preliminary hazard classification for the surveillance and maintenance activities of the facility. The result of this effort was the preliminary hazard classification for the 105-B Building surveillance and maintenance activities. The preliminary hazard classification was determined to be Nuclear Category 3. Additional hazard and accident analysis will be documented in a separate report to define the hazard controls and final hazard classification

Remaining Sites Verification Package for the 126-B-2, 183-B Clearwells

International Nuclear Information System (INIS)

Dittmer, L.M.

2007-01-01

The 126-B-2, 183-B Clearwells were built as part of the 183-B Water Treatment Facility and are composed of 2 covered concrete reservoirs. The bulk of the water stored in the clearwells was used as process water to cool the 105-B Reactor and as a source of potable water. Residual conditions were determined to meet the remedial action objectives specified in the Remaining Sites ROD through an evaluation of the available process knowledge. The results of the evaluation do not preclude any future uses and allow for unrestricted use of shallow zone soils. The results also indicate that residual concentrations are protective of groundwater and the Columbia River.

Comparative studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 data libraries by monte carlo modeling of high temperature reactors on plutonium based fuel cycles

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw; Cetnar, Jerzy

2004-01-01

We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library. (author)

Exxon nuclear neutronics design methods for pressurized water reactors. Supplement 2

International Nuclear Information System (INIS)

Skogen, F.B.; Stout, R.B.

1977-01-01

Modifications to the Exxon Nuclear PWR neutronic design calculational methods are presented as well as the results obtained when these improved methods are compared to reactor measurements. The basic PWR design tools remain unchanged; i.e., the XPOSE code is used for generating the basic nuclear parameters, the PDQ-7 code is used for calculating reactivity and x-y power distributions, and the XTG code is used for three-dimensional analysis. The recent start-up experiences at D. C. Cook Unit 1 and H. B. Robinson Unit 2 have provided a significant increase in the data base supporting the current ENC PWR neutronic methods. The verification comparisons contained in the supplement include reactor measurements from D. C. Cook Unit 1, Cycle 2; H. B. Robinson Unit 2, Cycles 4 and 5; Palisades Cycle 2, and R. E. Ginna, Cycle 7

Calculation and Analysis of B/T (Burning and/or Transmutation Rate of Minor Actinides and Plutonium Performed by Fast B/T Reactor

Directory of Open Access Journals (Sweden)

Marsodi

2006-01-01

Full Text Available Calculation and analysis of B/T (Burning and/or Transmutation rate of MA (minor actinides and Pu (Plutonium has been performed in fast B/T reactor. The study was based on the assumption that the spectrum shift of neutron flux to higher side of neutron energy had a potential significance for designing the fast B/T reactor and a remarkable effect for increasing the B/T rate of MA and/or Pu. The spectrum shifts of neutron have been performed by change MOX to metallic fuel. Blending fraction of MA and or Pu in B/T fuel and the volume ratio of fuel to coolant in the reactor core were also considered. Here, the performance of fast B/T reactor was evaluated theoretically based on the calculation results of the neutronics and burn-up analysis. In this study, the B/T rate of MA and/or Pu increased by increasing the blending fraction of MA and or Pu and by changing the F/C ratio. According to the results, the total B/T rate, i.e. [B/T rate]MA + [B/T rate]Pu, could be kept nearly constant under the critical condition, if the sum of the MA and Pu inventory in the core is nearly constant. The effect of loading structure was examined for inner or outer loading of concentric geometry and for homogeneous loading. Homogeneous loading of B/T fuel was the good structure for obtaining the higher B/T rate, rather than inner or outer loading

Status of data testing of ENDF/B-V reactor dosimetry file

International Nuclear Information System (INIS)

Magurno, B.A.

1979-01-01

The ENDF/B-V Reactor Dosimetry File was released August 1979, and Phase II data testing started. The results presented here are from Brookhaven National Laboratory only, and are considered preliminary. The tests include calculated spectrum-averaged cross sections using 235 U fission spectrum (Watt), 252 Cf spontaneous fission spectrum (Watt and Maxwellian), and the Coupled Fast Reactor Measurement Facility (CFRMF) spectrum. 6 tables

Optimization of a heterogeneous catalytic hydrodynamic cavitation reactor performance in decolorization of Rhodamine B: application of scrap iron sheets.

Science.gov (United States)

Basiri Parsa, Jalal; Ebrahimzadeh Zonouzian, Seyyed Alireza

2013-11-01

A low pressure pilot scale hydrodynamic cavitation (HC) reactor with 30 L volume, using fixed scrap iron sheets, as the heterogeneous catalyst, with no external source of H2O2 was devised to investigate the effects of operating parameters of the HC reactor performance. In situ generation of Fenton reagents suggested an induced advanced Fenton process (IAFP) to explain the enhancing effect of the used catalyst in the HC process. The reactor optimization was done based upon the extent of decolorization (ED) of aqueous solution of Rhodamine B (RhB). To have a perfect study on the pertinent parameters of the heterogeneous catalyzed HC reactor, the following cases as, the effects of scrap iron sheets, inlet pressure (2.4-5.8 bar), the distance between orifice plates and catalyst sheets (submerged and inline located orifice plates), back-pressure (2-6 bar), orifice plates type (4 various orifice plates), pH (2-10) and initial RhB concentration (2-14 mg L(-1)) have been investigated. The results showed that the highest cavitational yield can be obtained at pH 3 and initial dye concentration of 10 mg L(-1). Also, an increase in the inlet pressure would lead to an increase in the ED. In addition, it was found that using the deeper holes (thicker orifice plates) would lead to lower ED, and holes with larger diameter would lead to the higher ED in the same cross-sectional area, but in the same holes' diameters, higher cross-sectional area leads to the lower ED. The submerged operation mode showed a greater cavitational effects rather than the inline mode. Also, for the inline mode, the optimum value of 3 bar was obtained for the back-pressure condition in the system. Moreover, according to the analysis of changes in the UV-Vis spectra of RhB, both degradation of RhB chromophore structure and N-deethylation were occurred during the catalyzed HC process. Copyright © 2013 Elsevier B.V. All rights reserved.

Stage 2: dismantling of reactor case of the experimental F.B.R. Rapsodie

International Nuclear Information System (INIS)

Roger, J.

1994-01-01

This document defines the main objectives of stage 2 dismantling of the Rapsodie experimental fast neutron reactor and specifies its time schedule. The work already in progress consists in containing the reactor vessel and its internal equipment, as well as the neutron protection concrete, inside the two leak-tight barriers, and in dismantling all the systems and equipment systems contaminated by sodium. This work, which includes the destruction of 37 metric tons of contaminated sodium from the primary system, was begun in 1987 and will be completed in 1994. The duration of the waiting period for complete dismantling (stage 3) has not been defined. However, the containment and monitoring means implemented should allow a safe waiting period of several decades. (author). 4 figs

Emergency response guide-B ECCS guideline evaluation analyses for N reactor

International Nuclear Information System (INIS)

Chapman, J.C.; Callow, R.A.

1989-07-01

INEL conducted two ECCS analyses for Westinghouse Hanford. Both analyses will assist in the evaluation of proposed changes to the N Reactor Emergency Response Guide-B (ERG-B) Emergency Core System (ECCS) guideline. The analyses were a sensitivity study for reduced-ECCS flow rates and a mechanistically determined confinement steam source for a delayed-ECCS LOCA sequence. The reduced-ECCS sensitivity study established the maximum allowable reduction in ECCS flow as a function of time after core refill for a large break loss-of-coolant accident (LOCA) sequence in the N Reactor. The maximum allowable ECCS flow reduction is defined as the maximum flow reduction for which ECCS continues to provide adequate core cooling. The delayed-ECCS analysis established the liquid and steam break flows and enthalpies during the reflood of a hot core following a delayed ECCS injection LOCA sequence. A simulation of a large, hot leg manifold break with a seven-minute ECCS injection delay was used as a representative LOCA sequence. Both analyses were perform using the RELAP5/MOD2.5 transient computer code. 13 refs., 17 figs., 3 tabs

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)

Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

Energy Technology Data Exchange (ETDEWEB)

Dougherty, D.; Adams, J. W.

1983-08-01

Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation.

Evaluation of the Three Mile Island Unit 2 reactor building decontamination process

International Nuclear Information System (INIS)

Dougherty, D.; Adams, J.W.

1983-08-01

Decontamination activities from the cleanup of the Three Mile Island Unit 2 Reactor Building are generating a variety of waste streams. Solid wastes being disposed of in commercial shallow land burial include trash and rubbish, ion-exchange resins (Epicor-II) and strippable coatings. The radwaste streams arising from cleanup activities currently under way are characterized and classified under the waste classification scheme of 10 CFR Part 61. It appears that much of the Epicor-II ion-exchange resin being disposed of in commerical land burial will be Class B and require stabilization if current radionuclide loading practices continue to be followed. Some of the trash and rubbish from the cleanup of the reactor building so far would be Class B. Strippable coatings being used at TMI-2 were tested for leachability of radionuclides and chelating agents, thermal stability, radiation stability, stability under immersion and biodegradability. Actual coating samples from reactor building decontamination testing were evaluated for radionuclide leaching and biodegradation

Reactor BR2: Introduction

International Nuclear Information System (INIS)

Gubel, P.

2000-01-01

The BR2 reactor is still SCK-CEN's most important nuclear facility. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. A safety audit was conduced by the IAEA, the conclusions of which demonstrated the excellent performance of the plant in terms of operational safety. In 1999, the CALLISTO facility was extensively used for various programmes involving LWR pressure vessel materials, IASCC of LWR structural materials, fusion reactor materials and martensic steels for use in ADS systems. In 1999, BR2's commercial programmes were further developed

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

EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

International Nuclear Information System (INIS)

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2001-01-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2001-04-01

The BR2 is a materials testing reactor and is still one of SCK-CEN's important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. During the last three years, the availability of the installation was maintained at an average level of 97.6 percent. In the year 2000, the reactor was operated for a total of 104 days at a mean power of 56 MW. In 2000, most irradiation experiments were performed in the CALLISTO PWR loop. The report describes irradiations achieved or under preparation in 2000, including the development of advanced facilities and concept studies for new programmes. An overview of the scientific irradiation programmes as well as of the R and D programme of the BR2 reactor in 2000 is given.

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.)

Safety features of TR-2 reactor

International Nuclear Information System (INIS)

Tuerker, T.

2001-01-01

TR-2 is a swimming pool type research reactor with 5 MW thermal power and uses standard MTR plate type fuel elements. Each standard fuel element consist of 23 fuel plates with a meat + cladding thickness of 0.127 cm, coolant channel clearance is 0.21 cm. Originally TR-2 is designed for %93 enriched U-Al. Alloy fuel meat.This work is based on the preparation of the Final Safety Analyses Report (FSAR) of the TR-2 reactor. The main aspect is to investigate the behaviour of TR-2 reactor under the accident and abnormal operating conditions, which cowers the accident spectrum unique for the TR-2 reactor. This presentation covers some selected transient analyses which are important for the safety aspects of the TR-2 reactor like reactivity induced startup accidents, pump coast down (Loss of Flow Accident, LOFA) and other accidents which are charecteristic to the TR-2

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

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.

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

Analysis of the ZPPR-15 Critical Experiments with ENDF/B-V.2 and ENDF/B-VII.0 Data

International Nuclear Information System (INIS)

Kim, Sang Ji; Yang, Won Sik; Lee, Changho

2008-01-01

This paper presents the analysis results for the ZPPR-15 critical experiments. Using the ENDF/B-V.2 and ENDF/B-VII.0 data, three loading configurations of the ZPPR-15 Phase A experiments were analyzed with the ANL code suite for a fast reactor neutronics analysis, including the recently updated MC 2 -2 code. For the VIM Monte Carlo analyses with 3-D as-built models, the ENDF/B-VII.0 data improved the core multiplication factors by 0.21 to 0.37 %Δk, relative to the ENDF/B-V.2 data. With the plate heterogeneity effects taken into account by the SDX 1-D unit cell calculations, the DIF3D nodal transport solutions with the ENDF/B-V.2 data showed a good agreement for the core multiplication factors with the VIM Monte Carlo results to within 0.12 %Δk, but those with the ENDF/B-VII.0 data showed relatively larger deviations. Sensitivity studies based on the RZ models with homogenized cells showed excellent agreement for the core multiplication factors between the deterministic and Monte Carlo calculations to within 0.1 %Δk for both ENDF/B data. These results indicate that the MC 2 -2 methods are adequate for generating the multigroup cross sections for a fast reactor analysis, but the SDX process to account for the heterogeneity effect needs to be improved for the ENDF/B-VII.0 data. (authors)

Reactor BR2. Introduction

International Nuclear Information System (INIS)

Gubel, P.

2002-01-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system

Reactor BR2. Introduction

Energy Technology Data Exchange (ETDEWEB)

Gubel, P

2002-04-01

The BR2 materials testing reactor is one of SCK-CEN's most important nuclear facilities. After an extensive refurbishment to compensate for the ageing of the installation, the reactor was restarted in April 1997. In 2001, the reactor was operated for a total of 123 days at a mean power of 59 MW in order to satisfy the irradiation conditions of the internal and external programmes using mainly the CALLISTO PWR loop. The mean consumption of fresh fuel elements was 5.26 per 1000 MWd. Main achievements in 2001 included the development of a three-dimensional full-scale model of the BR2 reactor for simulation and prediction of irradiation conditions for various experiments; the construction of the FUTURE-MT device designed for the irradiation of fuel plates under representative conditions of geometry, neutron spectrum, heat flux and thermal-hydraulic conditions and the development of in-pile instrumentation and a data acquisition system.

Reactor benchmarks and integral data testing and feedback into ENDF/B-VI

International Nuclear Information System (INIS)

McKnight, R.D.; Williams, M.L.

1992-01-01

The role of integral data testing and its feedback into the ENDF/B evaluated nuclear data files are reviewed. The use of the CSEWG reactor benchmarks in the data testing process is discussed and selected results based on ENDF/B Version VI data are presented. Finally, recommendations are given to improve the implementation in future integral data testing of ENDF/B

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

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

Set of rules SOR 2 licensing of nuclear reactors

International Nuclear Information System (INIS)

1976-05-01

This is the set of rules promulgated by the Israel Atomic Energy Commission pursuant to the Supervision of Supplies and Services Law 5718-1957, Order regarding Supervision of Nuclear Reactors (1974) Chapter 3: Permits, to provide for the Licensing of Nuclear Reactors. (B.G.)

THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

International Nuclear Information System (INIS)

Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

1988-04-01

THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

ETOA, ABBN Multigroup Constants from ENDF/B for Fast Reactors

International Nuclear Information System (INIS)

Nishimura, Hideo

1977-01-01

1 - Nature of physical problem solved: Production of ABBN type group constants up to 70 groups for fast reactor calculations, reading ENDF/B library as input. 2 - Method of solution: The multigroup method of Bondarenko et al. is used for processing basic nuclear data. Calculational algorithms for an unresolved resonance region are the same as those in the MC 2 code. For a resolved resonance region, an ultrafine energy structure dependent on a level scheme is adopted. 3 - Restrictions on the complexity of the problem: Maximum number of: energy groups: 70; sigma 0 values: 6; temperatures: 5. Self-shielding factors for an unrealistically low value of sigma 0 are not guaranteed because of the approximations used in the unresolved resonance region

Core design calculations of IRIS reactor using modified CORD-2 code package

International Nuclear Information System (INIS)

Pevec, D.; Grgic, D.; Jecmenica, R.; Petrovic, B.

2002-01-01

Core design calculations, with thermal-hydraulic feedback, for the first cycle of the IRIS reactor were performed using the modified CORD-2 code package. WIMSD-5B code is applied for cell and cluster calculations with two different 69-group data libraries (ENDF/BVI rev. 5 and JEF-2.2), while the nodal code GNOMER is used for diffusion calculations. The objective of the calculation was to address basic core design problems for innovative reactors with long fuel cycle. The results were compared to our results obtained with CORD-2 before the modification and to preliminary results obtained with CASMO code for a similar problem without thermal-hydraulic feedback.(author)

Application of MCNPX 2.7.D for reactor core management at the research reactor BR2

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, Edgar

2011-01-01

The paper discusses application of the Monte Carlo burn up code MCNPX 2.7.D for whole core criticality and depletion analysis of the Material Testing Research Reactor BR2 at SCK-CEN in Mol, Belgium. Two different approaches in the use of MCNPX 2.7.D are presented. The first methodology couples the evolution of fuel depletion, evaluated by MCNPX 2.7.D in an infinite lattice with a steady-state 3-D power distribution in the full core model. The second method represents fully automatic whole core depletion and criticality calculations in the detailed 3-D heterogeneous geometry model of the BR2 reactor. The accuracy of the method and computational time as function of the number of used unique burn up materials in the model are being studied. The depletion capabilities of MCNPX 2.7.D are compared vs. the developed at the BR2 reactor department MCNPX & ORIGEN-S combined method. Testing of MCNPX 2.7.D on the criticality measurements at the BR2 reactor is presented. (author)

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

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

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

A Pebble-Bed Breed-and-Burn Reactor

International Nuclear Information System (INIS)

Greenspan, Ehud

2016-01-01

The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

A Pebble-Bed Breed-and-Burn Reactor

Energy Technology Data Exchange (ETDEWEB)

Greenspan, Ehud [Univ. of California, Berkeley, CA (United States)

2016-03-31

The primary objective of this project is to use three-dimensional fuel shuffling in order to reduce the minimum peak radiation damage of ~550 dpa present Breed-and-Burn (B&B) fast nuclear reactor cores designs (they feature 2-D fuel shuffling) call for to as close as possible to the presently accepted value of 200 dpa thereby enabling earlier commercialization of B&B reactors which could make substantial contribution to energy sustainability and economic stability without need for fuel recycling. Another objective is increasing the average discharge burnup for the same peak discharge burnup thereby (1) increasing the fuel utilization of 2-D shuffled B&B reactors and (2) reducing the reprocessing capacity required to support a given capacity of FRs that are to recycle fuel.

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

Energy Technology Data Exchange (ETDEWEB)

M. Chen; CM Regan; D. Noe

2006-01-09

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

TA-2 Water Boiler Reactor Decommissioning Project

International Nuclear Information System (INIS)

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

1991-06-01

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

German Phase B [risk study] highlights the role of [reactor] accident management

International Nuclear Information System (INIS)

Anon.

1989-01-01

Phase B of the German probabilistic risk assessment study, now scheduled for publication this month, suggests that reactor accident management measures can prevent or mitigate about 90 per cent of event sequences. (author)

Preparation of high quality superconducting thin MgB2 films for electronics

International Nuclear Information System (INIS)

Surdu, Andrei; Zdravkov, Vladimir; Sidorenko, Anatolie; Rossolenko, Anna; Ryazanov, Valerii; Bdikin, Igor; Kroemer, Oliver; Nold, Eberhard; Koch, Thomas; Schimmel, Thomas

2007-01-01

In this work we report the growth of high-Tc MgB 2 smooth films which are prepared in a two-step process: 1) deposition of the precursor films and 2) their annealing in Mg vapor with a specially designed, reusable reactor. Our method opens perspectives for the use of MgB 2 films in microelectronics, especially for high-frequency applications. (authors)

Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

International Nuclear Information System (INIS)

Bissani, M; O'Kelly, D S

2006-01-01

A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

Energy Technology Data Exchange (ETDEWEB)

Bissani, M; O' Kelly, D S

2006-05-08

A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

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

Thermal reactor benchmark testing of CENDL-2 and ENDF/B-6

International Nuclear Information System (INIS)

Liu Guisheng

1996-01-01

In order to test CENDL-2, ten homogeneous and eight heterogeneous thermal assemblies were used. Both of 123 group cross section libraries based on CENDL-2 and ENDF/B-6 were generated by a nuclear data processing system NSLINK, respectively. The calculations of resonance self-shielding, cell spectra, cell reaction rate ratios and effective multiplication factors (K eff ) of these assemblies have been performed by the modified PASC-1 code system. The calculated results using CENDL-2 show an excellent agreement with corresponding experimental values. However, for some assemblies the K eff values calculated by ENDF/B-6 data are underestimated. (7 tabs.)

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

International Nuclear Information System (INIS)

Ellis, R.J.

1990-01-01

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

BR2 Reactor: Irradiation of fuels

International Nuclear Information System (INIS)

Verwimp, A.

2005-01-01

Safe, reliable and economical operation of reactor fuels, both UO 2 and MOX types, requires in-pile testing and qualification up to high target burn-up levels. In-pile testing of advanced fuels for improved performance is also mandatory. The objectives of research performed at SCK-CEN are to perform Neutron irradiation of LWR (Light Water Reactor) fuels in the BR2 reactor under relevant operating and monitoring conditions, as specified by the experimenter's requirements and to improve the on-line measurements on the fuel rods themselves

Thermal reactor benchmark testing of CENDL-2 and ENDF/B-6

Energy Technology Data Exchange (ETDEWEB)

Guisheng, Liu [Chinese Nuclear Data Center, Beijing, BJ (China)

1996-06-01

In order to test CENDL-2, ten homogeneous and eight heterogeneous thermal assemblies were used. Both of 123 group cross section libraries based on CENDL-2 and ENDF/B-6 were generated by a nuclear data processing system NSLINK, respectively. The calculations of resonance self-shielding, cell spectra, cell reaction rate ratios and effective multiplication factors (K{sub eff}) of these assemblies have been performed by the modified PASC-1 code system. The calculated results using CENDL-2 show an excellent agreement with corresponding experimental values. However, for some assemblies the K{sub eff} values calculated by ENDF/B-6 data are underestimated. (7 tabs.).

Experience with the RE fuel transition at the Studsvik R2 reactor

International Nuclear Information System (INIS)

Pazsit, I.; Saltvedt, K.

1991-01-01

Irradiation of 7 LEU fuel elements is underway in the Studsvik R2 reactor. Four of these have 490 g U-235, and three 320 g U-235 loading, and the enrichment is 19.7% for all of them. The irradiation of LEU fuel started in 1987. The heavier elements have burnup figures 67% (CERCA), 50% (B and W), 47% (NUKEM) and 19% (B and W). One of the lighter elements has reached a burnup of 65%. To support the whole-core conversion process, reactor physical calculations were performed to see if a one-step conversion is possible with a suitable fuel management strategy such that all HEU fuel is burned up. The calculations show that it is possible to perform such a conversion with fuel elements containing 400 g U-235. (orig.)

Structural integrity of water reactor pressure boundary components

International Nuclear Information System (INIS)

Loss, F.J.

1977-01-01

The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

Ageing management of the BR2 research reactor

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

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

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.

Investigation of sensors and instrument components in boiling water reactors. Results from Oskarshamn 2, Barsebaeck 2 in Sweden and Kernkraftwerk Muehleberg in Switzerland

International Nuclear Information System (INIS)

Bergdahl, B.G.

1998-05-01

The reactor monitoring instruments are important for the operation and safety of the plants. Static properties of the instruments are controlled annually, but the dynamic properties are rarely, if ever, examined. This study is the result of a project initiated by the Swedish Nuclear Power Inspectorate. The examinations are based on signal analysis and simultaneous measurement of multiple signals. Results from Oskarshamn 2 (O2), Barsebaeck 2 (B2) and Kernkraftwerk Muehleberg (KKM) are discussed in this report. The presentation is focused on reactor pressure and reactor level signals. the analysis of O2 revealed that the dynamics for 3 out of 14 sensors was 'filtered', meaning that a rapid level displacement is registered with delay. Inspection showed that a 1 sec filter was installed instead of 1.2 sec. The study also showed that old pressure-sensors in use both at O2 and B2 could not cope with high frequencies, and that some level-sensors were disturbed by mechanical oscillations at Bw. At KKM, a 2 Hz resonance was observed with 12 pressure and level sensors. The oscillation was created by an old pressure sensor and influenced the other sensors through the common impulse network

Preparation of covariance data for the fast reactor. 2

International Nuclear Information System (INIS)

Shibata, Keiichi; Hasagawa, Akira

1998-03-01

For some isotopes important for core analysis of the fast reactor, covariance data of neutron nuclear data in the evaluated nuclear data library (JENDL-3.2) were presumed to file. Objected isotopes were 10-B, 11-B, 55-Mn, 240-Pu and 241-Pu. Physical amounts presumed on covariance were cross section, isolated and unisolated resonance parameters and first order Legendre coefficient of elastic scattering angle distribution. Presumption of the covariance was conducted in accordance with the data estimation method of JENDL-3.2 as possible. In other ward, when the estimated value was based on the experimental one, error of the experimental value was calculated, and when based on the calculated value, error of the calculated one was obtained. Their estimated results were prepared with ENDF-6 format. (G.K.)

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

Science.gov (United States)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

CO2 Energy Reactor - Integrated Mineral Carbonation: Perspectives on Lab-Scale Investigation and Products Valorization

OpenAIRE

Rafael M Santos; Pol CM Knops; Keesjan L Rijnsburger; Yi Wai eChiang

2016-01-01

To overcome the challenges of mineral CO2 sequestration, Innovation Concepts B.V. is developing a unique proprietary gravity pressure vessel (GPV) reactor technology and has focussed on generating reaction products of high economic value. The GPV provides intense process conditions through hydrostatic pressurization and heat exchange integration that harvests exothermic reaction energy, thereby reducing energy demand of conventional reactor designs, in addition to offering other benefits. In ...

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

BR2 reactor neutron beams

International Nuclear Information System (INIS)

Neve de Mevergnies, M.

1977-01-01

The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

Dose estimation in B16 tumour bearing mice for future irradiation in the thermal column of the TRIGA reactor after B/Gd/LDL adduct infusion

Energy Technology Data Exchange (ETDEWEB)

Protti, N., E-mail: nicoletta.protti@pv.infn.it [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy); Ballarini, F.; Bortolussi, S. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy); Bruschi, P. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy); Stella, S. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy); Geninatti, S.; Alberti, D.; Aime, S. [University of Torino, Chemistry Department, via Nizza 52, 10126 Torino (Italy); Altieri, S. [University of Pavia, Department of Nuclear and Theoretical Physics, via Bassi 6, 27100 Pavia (Italy)] [National Institute of Nuclear Physics (INFN) Section of Pavia, via Bassi 6, 27100 Pavia (Italy)

2011-12-15

To test the efficacy of a new {sup 10}B-vector compound, the B/Gd/LDL adduct synthesised at Torino University, in vivo irradiations of murine tumours are in progress at the TRIGA Mark II reactor of the Pavia University. A localised B16 melanoma tumour is generated in C57BL/6 mice and subsequently infused with the adduct. During the irradiation, the mouse will be put in a shield to protect the whole body except the tumour in the back-neck area. To optimise the treatment set-up, MCNP simulations were performed. A very simplified mouse model was built using MCNP geometry capabilities, as well as the geometry of the shield made of 99% {sup 10}B enriched boric acid. A hole in the shield is foreseen in correspondence of the back-neck region. Many configurations of the shield were tested in terms of neutron flux, dose distribution and mean induced activity in the tumour region and in the radiosensitive organs of the mouse. In the final set-up, up to five mice can be treated simultaneously in the reactor thermal column and the neutron fluence in the tumour region for 10 min of irradiation is of about 5 Multiplication-Sign 10{sup 12} cm{sup -2}.

Efficient H2O2/CH3COOH oxidative desulfurization/denitrification of liquid fuels in sonochemical flow-reactors.

Science.gov (United States)

Calcio Gaudino, Emanuela; Carnaroglio, Diego; Boffa, Luisa; Cravotto, Giancarlo; Moreira, Elizabeth M; Nunes, Matheus A G; Dressler, Valderi L; Flores, Erico M M

2014-01-01

The oxidative desulfurization/denitrification of liquid fuels has been widely investigated as an alternative or complement to common catalytic hydrorefining. In this process, all oxidation reactions occur in the heterogeneous phase (the oil and the polar phase containing the oxidant) and therefore the optimization of mass and heat transfer is of crucial importance to enhancing the oxidation rate. This goal can be achieved by performing the reaction in suitable ultrasound (US) reactors. In fact, flow and loop US reactors stand out above classic batch US reactors thanks to their greater efficiency and flexibility as well as lower energy consumption. This paper describes an efficient sonochemical oxidation with H2O2/CH3COOH at flow rates ranging from 60 to 800 ml/min of both a model compound, dibenzotiophene (DBT), and of a mild hydro-treated diesel feedstock. Four different commercially available US loop reactors (single and multi-probe) were tested, two of which were developed in the authors' laboratory. Full DBT oxidation and efficient diesel feedstock desulfurization/denitrification were observed after the separation of the polar oxidized S/N-containing compounds (S≤5 ppmw, N≤1 ppmw). Our studies confirm that high-throughput US applications benefit greatly from flow-reactors. Copyright © 2013 Elsevier B.V. All rights reserved.

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.)

Fission product release from SLOWPOKE-2 reactors

Energy Technology Data Exchange (ETDEWEB)

Harnden-Gillis, A M.C. [Queen` s Univ., Kingston, ON (Canada). Dept. of Physics

1994-12-31

Increasing radiation fields at several SLOWPOKE-2 reactors fuelled with highly enriched uranium aluminum alloy fuel have begun to interfere with the daily operation of these reactors. To investigate this phenomenon, samples of reactor container water and gas from the headspace were obtained at four SLOWPOKE-2 reactor facilities and examined by gamma ray spectroscopy methods. These radiation fields are due to the circulation of fission products within the reactor container vessel. The most likely source of the fission product release is an area of uranium-bearing material exposed to the coolant at the end weld line which originated at the time of fuel fabrication. The results of this study are compared with observations from an underwater visual examination of one core and the metallographic examination of archived fuel elements. 19 refs., 4 tabs., 8 figs.

Assessment of liquid hydrogen cooled MgB2 conductors for magnetically confined fusion

International Nuclear Information System (INIS)

Glowacki, B A; Nuttall, W J

2008-01-01

Importantly environmental factors are not the only policy-driver for the hydrogen economy. Over the timescale of the development of fusion energy systems, energy security issues are likely to motivate a shift towards both hydrogen production and fusion as an energy source. These technologies combine local control of the system with the collaborative research interests of the major energy users in the global economy. A concept Fusion Island Reactor that might be used to generate H 2 (rather than electricity) is presented. Exploitation of produced hydrogen as a coolant and as a fuel is proposed in conjunction with MgB 2 conductors for the tokomak magnets windings, and electrotechnical devices for Fusion Island's infrastructure. The benefits of using MgB 2 over the Nb-based conductors during construction, operation and decommissioning of the Fusion Island Reactor are presented. The comparison of Nb 3 Sn strands for ITER fusion magnet with newly developed high field composite MgB 2 PIT conductors has shown that at 14 Tesla MgB 2 possesses better properties than any of the Nb 3 Sn conductors produced. In this paper the potential of MgB 2 conductors is examined for tokamaks of both the conventional ITER type and a Spherical Tokamak geometry. In each case MgB 2 is considered as a conductor for a range of field coil applications and the potential for operation at both liquid helium and liquid hydrogen temperatures is considered. Further research plans concerning the application of MgB 2 conductors for Fusion Island are also considered

Pressurized water reactor simulator. Workshop material. 2. ed

International Nuclear Information System (INIS)

2005-01-01

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

Reactivity determination of the Al2O3-B4C burnable poison as a function of its concentration in the IPEN/MB-01 reactor

International Nuclear Information System (INIS)

Giada, Marino Reis

2005-01-01

Burnable poison rods made of Al 2 O 3 -B 4 C pellets with different concentrations of 10 B have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. The experiments evaluated the reactivity of the burnable poison rods as a function of the 10 B concentration, and the shadowing effect on the control rod reactivity worth as a function of the distance between the burnable position rods and the control rod. The results showed that the burnable poison rods have a non-linear behavior as function of the 10 B concentration, starting to reach an asymptotic value for concentrations higher than 7 g/cm 3 of 10 B. The shadowing effect on the control rods was substantial. When the burnable poison rods were beside the control rod, its reactivity worth decreased as much as 30 %, and when they were 10,5 cm distant, the control rod worth decreased by 7 %. The MCNP results for the burnable poison reactivity effects agreed within experimental errors with the measured values. (author)

The research reactor as a tool in the master in nuclear reactors in Argentina

International Nuclear Information System (INIS)

Notari, Carla

2003-01-01

complete the Master with a seminar: Nuclear Power Plants, and a Thesis. In the frame of the academic plan, multiple activities are organized related to research reactors and also to nuclear power plants. Since the very beginning the performance of selected experiments in a nuclear reactor was recognized as an extraordinary tool to give the students an insight in the principal phenomena associated with the chain reaction and the related engineering problems. This experiments have an intrinsic elevated cost, associated with the relevance of the installation and with the specialized personnel involved. CNEA provides the career with this educational instrument through the Ra-1 and RA-3 reactors located at Constituyentes and Ezeiza Atomic Centers respectively. Various activities are under way but the most established, in the Reactor Physics Course, is the estimation of kinetic parameters in RA-1 reactor. The practice includes three different experiments: Approach to critical and calibration of control rods by the compensation method: Starting in a subcritical state with source the calibration of control rod B1 vs B2 is done by introduction of the first and withdrawal of the second. The methods used are based on the Point Kinetic Model; Measurement of control rods effectivity by the rod-drop method: Separate Rod Drop of rods B1 B2 B3 of the overall ensemble B1 B2 B3 B4 and total scram starting with three withdrawn and one partially inserted, is the procedure followed to estimate the reactivity worth of B1 B2 B3 and scram. The Point Kinetic Model and the Modal Kinetic Model are used; Reactor noise technique for the estimation of reactor parameters: α and Λ. The kinetic parameters are estimated assuring that the Point Kinetic Model is valid (detection chambers near to the core), that the fluctuation of the fission density is the dominant source of the correlated part of neutron noise (measurement at low power, <10kw), the dominance of the fundamental armonic (simultaneous use of

Tribology study on TiB2+WSi2 composite against WC

Science.gov (United States)

Murthy, T. S. R. Ch.; Basha, M. M.; Sonber, J. K.; Singh, K.; Raju, K.; Sairam, K.; Nagaraj, A.; Majumdar, S.; Rao, G. V. S. Nageswara; Kain, Vivekanand

2018-04-01

Titanium diboride (TiB2) is one of the potential material for green energy applications such as neutron absorber in high temperature/advanced nuclear reactors, receiver materials for second generation concentrated solar power. We developed the process flow sheet for synthesis and consolidation of various series of TiB2 based materials in our laboratory. Amongst these, TiB2+WSi2 exhibited better sinterability and oxidation resistance properties. In the present work, tribology properties of TiB2+2.5%WSi2 composite was studied against WC-Co ball using different normal loads (5, 10 and 20 N) and frequencies (10, 15 Hz) under dry condition. Coefficient of friction (COF) and wear rate was measured at all test conditions. Wear mechanism was analyzed by microstructural characterization. It was found that COF is decreased from 0.46 to 0.36 with increasing load (5 to 20 N) at 15 Hz frequency; whereas at 10 Hz frequency COF is measured a constant average value of 0.49. The specific wear rate measured was minimum at 5 N load and 15 Hz frequency combination and was found to be 2.84×10-6 mm3/N m. The wear mechanisms identified during reciprocative sliding wear of composite were abrasion and surface tribo-oxidative reactions with delamination from tribo-zone.

Development of 'low activation superconducting wire' for an advanced fusion reactor

International Nuclear Information System (INIS)

Hishinuma, Y.; Yamada, S.; Sagara, A.; Kikuchi, A.; Takeuchi, T.; Matsuda, K.; Taniguchi, H.

2011-01-01

In the D-T burning plasma reactor beyond ITER such as DEMO and fusion power plants assuming the steady-state and long time operation, it will be necessary to consider carefully induced radioactivity and neutron irradiation properties on the all components for fusion reactors. The decay time of the induced radioactivity can control the schedule and scenarios of the maintenance and shutdown on the fusion reactor. V 3 Ga and MgB 2 compound have shorter decay time within 1 years and they will be desirable as a candidate material to realize 'low activation and high magnetic field superconducting magnet' for advanced fusion reactor. However, it is well known that J c -B properties of V 3 Ga and MgB 2 wires are lower than that of the Nb-based A15 compound wires, so the J c -B enhancements on the V 3 Ga and MgB 2 wires are required in order to apply for an advanced fusion reactor. We approached and succeeded to developing the new process in order to improve J c properties of V 3 Ga and MgB 2 wires. In this paper, the recent activities for the J c improvements and detailed new process in V 3 Ga and MgB 2 wires are investigated. (author)

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.)

DELIGHT-B/REDEL, point reactivity burnup code for high-temperature gas-cooled reactor cells

International Nuclear Information System (INIS)

Shindo, Ryuiti; Watanabe, Takashi.

1977-03-01

Code DELIGHT-2 was previously developed to analyze cell burnup characteristics and to produce few-group constants for core burnup calculation in high-temperature gas-cooled reactors. In the code, burnup dependency of the burnable poison, boron-10, is considered with the homogeneous model of space. In actuality, however, the burnable poison is used as homogeneous rods or uniform rods of small granular poison and graphite, to control the reactivity and power distribution. Precise analysis of the burnup characteristics is thus difficult because of the heterogeneity due to the configuration of poison rods. In cell burnup calculation, the DELIGHT-B, which is a modification of DELIGHT-2, takes into consideration this heterogeneous effect. The auxiliary code REDEL, a reduction of DELIGHT-B, used in combination with 3 dimensional diffusion code CITATION, is for core burnup calculation with the macro-scopic cross section model. (auth.)

Compact stellarators as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Valanju, P.; Zarnstorff, M.C.; Hirshman, S.; Spong, D.A.; Strickler, D.; Williamson, D.E.; Ware, A.

2001-01-01

Two types of compact stellarators are examined as reactors: two- and three-field-period (M=2 and 3) quasi-axisymmetric devices with volume-average =4-5% and M=2 and 3 quasi-poloidal devices with =10-15%. These low-aspect-ratio stellarator-tokamak hybrids differ from conventional stellarators in their use of the plasma-generated bootstrap current to supplement the poloidal field from external coils. Using the ARIES-AT model with B max =12T on the coils gives Compact Stellarator reactors with R=7.3-8.2m, a factor of 2-3 smaller R than other stellarator reactors for the same assumptions, and neutron wall loadings up to 3.7MWm -2 . (author)

Advances in Reactor Physics, Mathematics and Computation. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, Volume 2, are divided into 7 sessions bearing on: - session 7: Deterministic transport methods 1 (7 conferences), - session 8: Interpretation and analysis of reactor instrumentation (6 conferences), - session 9: High speed computing applied to reactor operations (5 conferences), - session 10: Diffusion theory and kinetics (7 conferences), - session 11: Fast reactor design, validation and operating experience (8 conferences), - session 12: Deterministic transport methods 2 (7 conferences), - session 13: Application of expert systems to physical aspects of reactor design and operation.

Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This document describes the specific dispositions relative to the nuclear reactor domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor and of the retraining, the Passerelle formation, are presented. (A.L.B.)

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.)

Monte Carlo modelling of the Belgian materials testing reactor BR2: present status

International Nuclear Information System (INIS)

Verboomen, B.; Aoust, Th.; Raedt, Ch. de; Beeckmans de West-Meerbeeck, A.

2001-01-01

A very detailed 3-D MCNP-4B model of the BR2 reactor was developed to perform all neutron and gamma calculations needed for the design of new experimental irradiation rigs. The Monte Carlo model of BR2 includes the nearly exact geometrical representation of fuel elements (now with their axially varying burn-up), of partially inserted control and regulating rods, of experimental devices and of radioisotope production rigs. The multiple level-geometry possibilities of MCNP-4B are fully exploited to obtain sufficiently flexible tools to cope with the very changing core loading. (orig.)

Expected characteristics of future reactors for human beings

International Nuclear Information System (INIS)

Taketani, Kiyoaki

1992-01-01

Based on four reactor safety components (namely: a) God-given safety, b) Equipment safety, c) Quick-response safety, d) Containing safety), categorical assessment is made of various nuclear reactor concepts ranging from present existing reactors to future reactors based on innovative reactor design. In pursuit of nuclear reactor safety, ultimate characteristics of the ideal nuclear reactor are expected to coincide with those of an inherently safe reactor. A definition of 'inherently safe' has already been proposed by a committee in Japan. As a realistic and existable reactor, which is as close to the ideal reactor, a future reactor which is almost the same as a global reactor, is proposed. This global reactor must be constructable anywhere on earth and must permit easy operation and maintenance by anyone. It is also discussed to identify what behavior is expected of the global reactor under various conditions. At the same time, this future reactor which includes the global reactor, should solve a) the nuclear fuel resource issue, b) efficient utilization of fission energy and c) environmental issues as the greenhouse effect. (author). 7 refs., 2 figs

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

Farewell to a reactor

International Nuclear Information System (INIS)

Skanborg, P.

1976-01-01

Denmark's second reactor, DR 2, whose first criticality took place the night of 18/19 December 1958 was shut down for the last time on 31 October 1975. It was a light-water moderrated and cooled reactor of swimming-pool type with a thermal power of 5 MW, using 90% enriched uranium. The operation is described. The reactor and auxiliary equipment are now being put 'in store' - all fuel elements sent for reprocessing, the reactor tank and cooling circuits emptied, and a lead shielding placed over the tank opening. The rest of the equipment will remain in place. (B.P.)

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

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

International Nuclear Information System (INIS)

Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

1993-01-01

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

Gaseous swelling of B4C and UO2 fuel: similarities and differences

International Nuclear Information System (INIS)

Evdokimov, I.; Khoruzhii, O.; Kourtchatov, S.; Likhanskii, V.; Matweev, L.

2001-01-01

A major factor limiting the resource of control rods (CRs) for WWER-1000 reactors is their radiation damage. Radiation induced embrittlement of the CRs cladding, core swelling and gaseous internal pressure in CRs result in mechanical core-cladding interaction. This work is devoted to the physical analysis of processes that control the structural changes in neutron absorber elements with B 4 C under irradiation in water reactors. Particularly, the analysis of mechanisms of the helium porosity formation in B 4 C is undertaken. In view of the deficiency of experimental data on the subject, a fruitful approach to the problem is a comparative analysis of the swelling mechanisms in B 4 C absorber and UO 2 fuel. Using this similarity a phenomenological model of fission gas behavior in boron carbide is proposed. The model predictions for radial profile of 10 B burnup under influence of thermal and epithermal neutrons are compared with experimental results. The main results show that despite the external similarity of the process of fission gas accumulation in UO 2 and in B 4 C, phenomenology of gaseous swelling is much different for the fuel and the CR core. The reason for that difference is the distinction of physical conditions in irradiated fuel and CR core

TPDWR2: thermal power determination for Westinghouse reactors, Version 2. User's guide

International Nuclear Information System (INIS)

Kaczynski, G.M.; Woodruff, R.W.

1985-12-01

TPDWR2 is a computer program which was developed to determine the amount of thermal power generated by any Westinghouse nuclear power plant. From system conditions, TPDWR2 calculates enthalpies of water and steam and the power transferred to or from various components in the reactor coolant system and to or from the chemical and volume control system. From these results and assuming that the reactor core is operating at constant power and is at thermal equilibrium, TPDWR2 calculates the thermal power generated by the reactor core. TPDWR2 runs on the IBM PC and XT computers when IBM Personal Computer DOS, Version 2.00 or 2.10, and IBM Personal Computer Basic, Version D2.00 or D2.10, are stored on the same diskette with TPDWR2

Once-through CANDU reactor models for the ORIGEN2 computer code

International Nuclear Information System (INIS)

Croff, A.G.; Bjerke, M.A.

1980-11-01

Reactor physics calculations have led to the development of two CANDU reactor models for the ORIGEN2 computer code. The model CANDUs are based on (1) the existing once-through fuel cycle with feed comprised of natural uranium and (2) a projected slightly enriched (1.2 wt % 235 U) fuel cycle. The reactor models are based on cross sections taken directly from the reactor physics codes. Descriptions of the reactor models, as well as values for the ORIGEN2 flux parameters THERM, RES, and FAST, are given

Re criticality assessment following reactor core damage in Fukushima unit 2

International Nuclear Information System (INIS)

Jeong, Hae Sun; Song, Jin Ho; Park, Chang Je; Ha, Kwang Soon; Song, Yong Mann; Ryu, Eun Hyun

2012-01-01

Following the severe core damage accident at the Fukushima nuclear power plants (NPPs), many researchers have studied a possible Re criticality caused by core melting or corium. However, no one can accurately examine the internal conditions of the reactor vessel, and thus there have been different opinions from some organizations depending on their assumption and analysis methods. If there is a potential Re criticality in the reactor vessel, some counter plans for the accident management should be established to prevent and mitigate re criticality, and to return the plant to a safe and stable state. In this study, the criticality level following a severe core damage accident was first analyzed using the MCNPX 2.6.0 code. Based on this result, practical strategies in terms of accident management were obtained by charging soluble boron (H 3B O 3) into re flooded water

Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

International Nuclear Information System (INIS)

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

1988-01-01

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

A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

Directory of Open Access Journals (Sweden)

Hwanyeal Yu

2015-12-01

Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

Irradiation effects on Zr-2.5Nb in power reactors

Energy Technology Data Exchange (ETDEWEB)

Song, C., E-mail: Carol.Song@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, 'High Power Channel-type Reactor') reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr- 2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge. (author)

Specific schedule conditions for the formation of personnel of A or B category working in nuclear facilities. Option nuclear reactor-borne

CERN Document Server

Int. At. Energy Agency, Wien

2002-01-01

This document describes the specific dispositions relative to the nuclear reactor-borne domain, for the formation to the conventional and radiation risks prevention of personnel of A or B category working in nuclear facilities. The application domain, the applicable documents, the liability, the specificity of the nuclear reactor-borne and of the retraining, the Passerelle formation, are presented. (A.L.B.)

A computer based I and C to improve nuclear power plant operation and safety

International Nuclear Information System (INIS)

Appell, Bernard; Beltranda, Guy

1995-01-01

Chooz B1 is the head of 1400 MW N4 series (up to now, a programme of 4 identical units have been ordered) will be put on the French national grid this year (1995). Hot tests were performed last autumn. Loading fuel is scheduled in June 95. Chooz B2 will follow 6 months later. This nuclear power plant series is fully French designed. It comprises, compared to the former 1300 MW series, evolution concerning mainly the turbine, the steam generators, the primary pumps, the I and C and the man machine interface. A major improvement for the plant operation and safety is the computerised I and C which comprises the protection system (specific controllers), the process automation system (off the shelf equipment), the computer based and conventional man machine interfaces associated. These systems were put in service last year. This paper is focused on them. (author)

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

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.)

Characterization of the Three Mile Island Unit-2 reactor building atmosphere prior to the reactor building purge

International Nuclear Information System (INIS)

Hartwell, J.K.; Mandler, J.W.; Duce, S.W.; Motes, B.G.

1981-05-01

The Three Mile Island Unit-2 reactor building atmosphere was sampled prior to the reactor building purge. Samples of the containment atmosphere were obtained using specialized sampling equipment installed through penetration R-626 at the 358-foot (109-meter) level of the TMI-2 reactor building. The samples were subsequently analyzed for radionuclide concentration and for gaseous molecular components (O 2 , N 2 , etc.) by two independent laboratories at the Idaho National Engineering Laboratory (INEL). The sampling procedures, analysis methods, and results are summarized

New ceramics for nuclear industry. Case of fission and fusion reactors

International Nuclear Information System (INIS)

Yvars, M.

1979-10-01

The ceramics used in the nuclear field are described as is their behaviour under radiation. 1) Power reactors - nuclear fission. Ceramics enter into the fabrication of nuclear fuels: oxides, carbides, uranium or plutonium nitrides or oxy-nitrides. Silicon carbide SiC is used for preparing the fuels of helium cooled high temperature reactors. Its use is foreseen in the design of gas high temperature gas thermal exchangers, as is silicon nitride (Si 3 N 4 ). In the materials for safety or control rods, the intense neutron flows induce nuclear reactions which increase the temperature of the neutron absorbing material. Boron carbide B 4 C, rare earth oxides Ln 2 O 3 , or B 4 C-Cu or B 4 C-Al cermets are employed. Burnable poison materials are formed of Al 2 O 3 -B 4 C or Al 2 O 3 -Ln 2 O 3 cermets. The moderators of thermal neutron reactors are in high purety polycrystalline graphite. For the thermal insulation of reactor vessels and jackets, honeycomb ceramics are used as well as ceramic fibres on an increasing scale (kaolin, alumina and other fibres). 2) fusion reactors (Tokomak). These require refractory materials with a low atomic number. Carbon fibres, boron carbide, some borons (Al B 12 ), silicon nitrides and oxy-nitrides and high density alumina are the substances considered [fr

Application of a new cross section library based on ENDF/B-IV to reactor core analysis

International Nuclear Information System (INIS)

Lima Bezerra, J. de.

1991-04-01

The use of the ENDF/B-IV library in the LEOPARD code for the Angra-1 reactor simulation is presented. The results are compared to those obtained using the ENDF/B-II library and show better values for the power distribution but an underestimated global reactivity as compared to experimental results. (F.E.). 1 ref, 55 figs, 1 tab

Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

International Nuclear Information System (INIS)

Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

1998-01-01

Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

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.)

Enrichment reduction calculations for the DIDO reactor. App. B

International Nuclear Information System (INIS)

Constantine, G.; Javadi, M.; Thick, E.

1985-01-01

The possibility has been raised that DIDO/PLUTO type heavy water moderated reactors can be operated with fuel of lower than the 75% enrichment material currently in use with the object of increasing the proliferation resistance of the fuel cycle. This paper sets out to examine the reactor physics aspects of enrichment reductions to 45% and 20% for Harwell's MTR's as part of an IAEA collaborative exercise currently being conducted to examine the topic in a more general way for the whole class of heavy water moderated reactors. The reactor physics tool used at Harwell is WIMSE, the Winfrith Improved Multigroup Scheme, a suite of linked reactor physics codes which has been used extensively for light water, heavy water and graphite moderated thermal reactors. The course of the calculations and the WIMSE modules involved in this study are described briefly

Reactor handbook. 2. rev. ed.

International Nuclear Information System (INIS)

Lederer, B.J.; Wildberg, D.W.

1992-01-01

On the basis of the guidelines on expert knowledge, the book discusses the subjects of atomic physics, heat transfer, nuclear power plants, reactor materials, radiation protection, reactor safety, reactor instrumentation, and reactor operation, with special regard to nuclear power plants with LWR-type reactors. The book is intended for shift personnel, especially gang bosses, reactor operators, and control station operators: for this reason a practical and rather popular style has been chosen. However, the book will also be a manual for other operating personnel, personnel of producer companies, expert organisations, authorities, and students. It can be used as a textbook for staff training, a manual for the practice, and as accompanying book for teaching at nuclear engineering schools. (orig.) With 173 figs [de

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.

Concept on coupled spectrum B/T (burning and/or transmutation) reactor for treatment of minor actinides by thermal and fast neutrons

International Nuclear Information System (INIS)

Aziz, Ferhat; Kitamoto, Asashi

1996-01-01

A conceptual design of B/T (burning and/or transmutation) reactor based on a modified conventional 1150 MWe-PWR system, with core consisted of two concentric regions for thermal and fast neutrons, was proposed herein for B/T treatment of MA (minor actinides). The B/T fuel considered was supposed such that MA discharged from 1 GWe-LWR was blended homogeneously with the composition of LWR fuel. In the outer region 23- Np, 241 Am and 243 Am were loaded and burned by thermal neutron, while in the inner region 244 Cm was loaded and burned mainly by fast neutron. The geometry of B/T fuel and the fuel assembly in the outer region was left in the same condition to those of standard PWR while in the inner region the B/T fuel was arranged in the hexagonal geometry, allowed high fuel to coolant volume ratio (V m /V f ), to keep the harder neutron spectrum. Two cases of the Coupled Spectrum B/T Reactor (CSR) with different (V m 1 f ) ratio in the inner region were studied, and the results for the tight lattice with (V m /V f ) = 0.5 showed that those isotopes approached the equilibrium composition after about 5 recycle period, when the CSR was operated under the reactivity swing of 2.8 % dk/k. The evaluations on the void coefficient of reactivity, the Doppler effect and the reactivity swing showed that the CSR concept has the inherent safety and can burn and/or transmute all kind of MA in a single reactor. This CSR can burn about 808 kg of MA in one recycle period of 3 years, which is equivalent to the discharged fuel from about 12 units of LWR in a year. (author)

2-DB, 2-D Multigroup Diffusion, X-Y, R-Theta, Hexagonal Geometry Fast Reactor, Criticality Search

International Nuclear Information System (INIS)

Little, W.W. Jr.; Hardie, R.W.; Hirons, T.J.; O'Dell, R.D.

1969-01-01

1 - Description of problem or function: 2DB is a flexible, two- dimensional (x-y, r-z, r-theta, hex geometry) diffusion code for use in fast reactor analyses. The code can be used to: (a) Compute fuel burnup using a flexible material shuffling scheme. (b) Perform criticality searches on time absorption (alpha), material concentrations, and region dimensions using a regular or adjoint model. Criticality searches can be performed during burnup to compensate for fuel depletion. (c) Compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Standard source-iteration techniques are used. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy (group) averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes, are formed by the user. The code does not contain built-in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated. The current 1108 version, however, is nominally restricted to 50 energy groups in a 65 K memory. In the 6600 version the power fraction, average burnup rate, and breeding ratio calculations are limited to reactors with a maximum of 50 zones

Fusion reactor materials

International Nuclear Information System (INIS)

Sethi, V.K.; Scholz, R.; Nolfi, F.V. Jr.; Turner, A.P.L.

1980-01-01

Data are given for each of the following areas: (1) effects of irradiation on fusion reactor materials, (2) hydrogen permeation and materials behavior in alloys, (3) carbon coatings for fusion applications, (4) surface damage of TiB 2 coatings under energetic D + and 4 He + irradiations, and (5) neutron dosimetry

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

Energy Technology Data Exchange (ETDEWEB)

Montes Ponce de Leon, J.

1974-07-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Neutronic characteristics of linear-assembly breed-and-burn reactors

International Nuclear Information System (INIS)

Petroski, Robert; Forget, Benoit; Forsberg, Charles

2012-01-01

Highlights: ► Simple models used to characterize general behavior of linear-assembly B and B reactors. ► Diffusion theory model developed to explain axial distributions, height vs. reactivity. ► Neutron excess concept reformulated to include linear-assembly B and B reactors. ► Designed model of B and B reactor started using melt-refined B and B reactor used fuel. ► Computed doubling time of fuel cycle requiring no chemical separations. - Abstract: Linear-assembly breed-and-burn (B and B) reactors are B and B reactors that use axially connected assemblies similar to conventional LWR or fast reactor fuel assemblies. Methods for analyzing linear-assembly B and B reactors and their fuel cycles are developed and applied. General neutronic characteristics of linear-assembly B and B reactors are analyzed, including the effects that burnup, shuffling sequence, and radial and axial size have on equilibrium-cycle k-effective. The mechanisms that give rise to a highly peaked axial burnup distribution are explained, and a method for predicting peak burnup vs. k-effective based on infinite-medium depletion calculations is developed. Next, the neutron excess concept from previous studies of B and B reactors is extended to apply to linear-assembly B and B reactors, which allows the amount of starter fuel needed to establish a given equilibrium cycle to be calculated. Several example applications of the neutron excess formulation are given. First, an example model of a linear-assembly B and B reactor is analyzed to find the neutron excess cost of an equilibrium cycle. Second, simple one-dimensional models are used to predict the neutron excess value obtainable from different starter fuel configurations. Finally, these ideas are applied to design a fuel cycle consisting of linear-assembly B and B reactors and fuel recycling via a melt refining process. The neutron excess concept is used to design an appropriate starter fuel configuration made from melt refined fuel, which

The Vaporization of B2O3(l) to B2O3(g) and B2O2(g)

Science.gov (United States)

Jacobson, Nathan S.; Myers, Dwight L.

2011-01-01

The vaporization of B2O3 in a reducing environment leads to formation of both B2O3(g) and B2O2(g). While formation of B2O3(g) is well understood, many questions about the formation of B2O2(g) remain. Previous studies using B(s) + B2O3(l) have led to inconsistent thermodynamic data. In this study, it was found that after heating, B(s) and B2O3(l) appear to separate and variations in contact area likely led to the inconsistent vapor pressures of B2O2(g). To circumvent this problem, an activity of boron is fixed with a two-phase mixture of FeB and Fe2B. Both second and third law enthalpies of formation were measured for B2O2(g) and B2O3(g). From these the enthalpies of formation at 298.15 K are calculated to be -479.9 +/- 41.5 kJ/mol for B2O2(g) and -833.4 +/- 13.1 kJ/mol for B2O3(g). Ab initio calculations to determine the enthalpies of formation of B2O2(g) and B2O3(g) were conducted using the W1BD composite method and show good agreement with the experimental values.

An extended conventional fuel cycle for the B and W mPower{sup TM} small modular nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Scarangella, M. J. [Babcock and Wilcox Company, 109 Ramsey Place, Lynchburg, VA 24502 (United States)

2012-07-01

The B and W mPower{sup TM} reactor is a small pressurized water reactor (PWR) with an integral once-through steam generator and a thermal output of about 500 MW; it is intended to replace aging fossil power plants of similar output. The core is composed of 69 reduced-height PWR assemblies with the familiar 17 x 17 fuel rod array. The Babcock and Wilcox Company (B and W) is offering a core loading and cycle management plan for a four-year cycle based on its presumed attractiveness to potential customers. This option is a once-through fuel cycle in which the entire core is discharged and replaced after four years. In addition, a conventional fuel utilization strategy, employing a periodic partial reload and shuffle, was developed as an alternative to the four-year once-through fuel cycle. This study, which was performed using the Studsvik core design code suite, is a typical multi-cycle projection analysis of the type performed by most fuel management organizations such as fuel vendors and utilities. In the industry, the results of such projections are used by the financial arms of these organizations to assist in making long-term decisions. In the case of the B and W mPower reactor, this analysis demonstrates flexibility for customers who consider the once-through fuel cycle unacceptable from a fuel utilization standpoint. As expected, when compared to the once-through concept, reloads of the B and W mPower reactor will achieve higher batch average discharge exposure, will have adequate shut-down margin, and will have a relatively flat hot excess reactivity trend at the expense of slightly increased peaking. (authors)

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.

Optimized Control Rods of the BR2 Reactor

Energy Technology Data Exchange (ETDEWEB)

Kalcheva, Silva; Koonen, E.

2007-09-15

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Optimized Control Rods of the BR2 Reactor

International Nuclear Information System (INIS)

Kalcheva, Silva; Koonen, E.

2007-01-01

At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

Plasma-catalyst hybrid reactor with CeO2/γ-Al2O3 for benzene decomposition with synergetic effect and nano particle by-product reduction.

Science.gov (United States)

Mao, Lingai; Chen, Zhizong; Wu, Xinyue; Tang, Xiujuan; Yao, Shuiliang; Zhang, Xuming; Jiang, Boqiong; Han, Jingyi; Wu, Zuliang; Lu, Hao; Nozaki, Tomohiro

2018-04-05

A dielectric barrier discharge (DBD) catalyst hybrid reactor with CeO 2 /γ-Al 2 O 3 catalyst balls was investigated for benzene decomposition at atmospheric pressure and 30 °C. At an energy density of 37-40 J/L, benzene decomposition was as high as 92.5% when using the hybrid reactor with 5.0wt%CeO 2 /γ-Al 2 O 3 ; while it was 10%-20% when using a normal DBD reactor without a catalyst. Benzene decomposition using the hybrid reactor was almost the same as that using an O 3 catalyst reactor with the same CeO 2 /γ-Al 2 O 3 catalyst, indicating that O 3 plays a key role in the benzene decomposition. Fourier transform infrared spectroscopy analysis showed that O 3 adsorption on CeO 2 /γ-Al 2 O 3 promotes the production of adsorbed O 2 - and O 2 2‒ , which contribute benzene decomposition over heterogeneous catalysts. Nano particles as by-products (phenol and 1,4-benzoquinone) from benzene decomposition can be significantly reduced using the CeO 2 /γ-Al 2 O 3 catalyst. H 2 O inhibits benzene decomposition; however, it improves CO 2 selectivity. The deactivated CeO 2 /γ-Al 2 O 3 catalyst can be regenerated by performing discharges at 100 °C and 192-204 J/L. The decomposition mechanism of benzene over CeO 2 /γ-Al 2 O 3 catalyst was proposed. Copyright © 2017 Elsevier B.V. All rights reserved.

Upgrading of the research reactors FRG-1 and FRG-2

International Nuclear Information System (INIS)

Krull, W.

1981-01-01

In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

Research and development of a super fast reactor (12). Considerations for the reactor characteristics

International Nuclear Information System (INIS)

Goto, Shoji; Ishiwatari, Yuki; Oka, Yoshiaki

2008-01-01

A research program aimed at developing the Super Fast Reactor (Super FR) has been entrusted by the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan since December 2005. It includes the following three projects. (A) Development of the Super Fast Reactor concept. (B)Thermal-hydraulic experiments. (C) Materials development. Tokyo Electric Power Company (TEPCO) has joined this program and works on part (A) together with the University of Tokyo. From the utility's viewpoint, it is important to consider the most desirable characteristics for Super FR to have. Four issues were identified in project (A), (1) Fuel design, (2) Reactor core design, (3) Safety, and (4) Plant characteristics of Super FR. This report describes the desired characteristics of Super FR with respect to item (1) fuel design and item (2) Reactor core design, as compared with a boiling water reactor (BWR) plant. The other two issues will be discussed in this project, and will also be considered in the design process of Super FR. (author)

Enhancement of actinide incineration and transmutation rates in Ads EAP-80 reactor core with MOX PuO2 and UO2 fuel

International Nuclear Information System (INIS)

Kaltcheva-Kouzminava, S.; Kuzminov, V.; Vecchi, M.

2001-01-01

Neutronics calculations of the accelerator driven reactor core EAP-80 with UO 2 and PuO 2 MOX fuel elements and Pb-Bi coolant are presented in this paper. Monte Carlo optimisation computations of several schemes of the EAP-80 core with different types of fuel assemblies containing burnable absorber B4 C or H 2 Zr zirconium hydride moderator were performed with the purpose to enhance the plutonium and actinide incineration rate. In the first scheme the reactor core contains burnable absorber B4 C arranged in the cladding of fuel elements with high enrichment of plutonium (up to 45%). In the second scheme H2 Zr zirconium hydride moderated zones were located in fuel elements with low enrichment (∼20%). In both schemes the incineration rate of plutonium is about two times higher than in the reference EAP-80 core and at the same time the power density distribution remains significantly unchanged compared to the reference core. A hybrid core containing two fuel zones one of which is the inner fuel region with UO 2 and PuO 2 high enrichment plutonium fuel and the second one is the outer region with fuel elements containing zirconium hydride layer was also considered. Evolution of neutronics parameters and actinide transmutation rates during the fuel burn-up is presented. Calculations were performed using the MCNP-4B code and the SCALE 4.3 computational system. (author)

Annual report on JEN-1 and JEN-2 Reactors

International Nuclear Information System (INIS)

Montes Ponce de Leon, J.

1974-01-01

In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

International Nuclear Information System (INIS)

Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

1987-01-01

A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

B2B or Not to Be: Does B2B E-Commerce Increase Labour Productivity?

OpenAIRE

Bertschek, Irene; Fryges, Helmut; Kaiser, Ulrich

2004-01-01

We implement an endogeneous switching-regression model for labour productivity and firms' decision to use business-to-business (B2B) e-commerce. Our approach allows B2B usage to affect any parameter of the labour productivity equation and to properly take account of strategic complementarities between the input factors and B2B usage. Empirical evidence from 1,394 German firms shows that firms using B2B e-commerce have a significantly higher output elasticity with respect to ICT-investment and...

System Design of a Supercritical CO_2 cooled Micro Modular Reactor

International Nuclear Information System (INIS)

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

2014-01-01

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

The SLOWPOKE-2 reactor with low enrichment uranium oxide fuel

International Nuclear Information System (INIS)

Townes, B.M.; Hilborn, J.W.

1985-06-01

A SLOWPOKE-2 reactor core contains less than 1 kg of highly enriched uranium (HEU) and the proliferation risk is very low. However, to overcome proliferation concerns a new low enrichment uranium (LEU) fuelled reactor core has been designed. This core contains approximately 180 fuel elements based on the Zircaloy-4 clad UOsub(2) CANDU fuel element, but with a smaller outside diameter. The physics characteristics of this new reactor core ensure the inherent safety of the reactor under all conceivable conditions and thus the basic SLOWPOKE safety philosophy which permits unattended operation is not affected

Grouping of HLW in partitioning for B/T (burning and/or transmutation) treatment with neutron reactors based on three criteria

International Nuclear Information System (INIS)

Kitamoto, Mulyanto; Kitamoto, Asashi

1995-01-01

A grouping concept of HLW in partitioning for B/T (burning and/or transmutation) treatment by fission reactor was developed in order to improve the disposal in waste management from the safety aspect. The selecting and grouping concept was proposed herein, such as Group MA1 (Np, Am, and unrecovered U and Pu), Group MA2 (Cm, and trace quantity of Cf, etc.), Group A (Tc and I), Group B (Cs and Sr) and Group R (the partitioned remains of HLW), judging from the three criteria for B/T treatment, based on (1) the concept of the potential risk estimated by the hazard index for long-term tendency based on ALI (2) the concept of the relative dose factor related to the adsorbed migration rate transferred through ground water, and (3) the concept of the decay acceleration factor, the burning and/or transmutation characteristics for recycle B/T treatment. (author)

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

Operation of the SLOWPOKE-2 reactor in Jamaica

Energy Technology Data Exchange (ETDEWEB)

Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

2001-07-01

Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

Irradiation techniques at BR2 reactor

International Nuclear Information System (INIS)

Hebel, W.

1978-01-01

Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

Radiation protection at the RA Reactor in 1985, Part -2, Annex 2b, Environmental Radioactivity control, Control of air contamination

International Nuclear Information System (INIS)

Patic, D.; Smiljanic, R.; Zaric, M.; Savic, Z.; Ristic, D.

1985-01-01

During the period from November 1984 - November 1985, within the radioactivity control on the Vinca Institute site air contamination radioactive aerosol contents was measured. Control was done on 4 measuring stations, two in the Institute and two locations in the direction of wind i.e. Belgrade, 2 km and 7 km away from the Institute respectively. This position of the measuring locations enables control of radiation safety of the Institute, as well as environment of Belgrade taking into account the existence of the reactor and other possible contaminants in the Institute [sr

Exergy analysis of a hydrogen fired combined cycle with natural gas reforming and membrane assisted shift reactors for CO2 capture

International Nuclear Information System (INIS)

Atsonios, K.; Panopoulos, K.D.; Doukelis, A.; Koumanakos, A.; Kakaras, Em.

2012-01-01

Highlights: ► Exergy analysis of NGCC with CCS. ► WGS-MR: exergetically efficient technology for CCS, less than 2% total exergy losses. ► 10% of total exergy dissipation in the ATR. ► Optimization of ATR operation and CO 2 stream treatment. - Abstract: Hydrogen production from fossil fuels together with carbon capture has been suggested as a means of providing a carbon free power. The paper presents a comparative exergetic analysis performed on the hydrogen production from natural gas with several combinations of reactor systems: (a) oxy or air fired autothermal reforming with subsequent water gas shift reactor and (b) membrane reactor assisted with shift catalysts. The influence of reactor temperature and pressure as well as operating parameter steam-to-carbon ratio, is also studied exergetically. The results indicate optimal power plant configurations with CO 2 capture, or hydrogen delivery for industrial applications.

Photocatalytic Degradation Property of NANO-TiO2/DIATOMITE for Rodamine B Dye Wastewater

Science.gov (United States)

Liu, Yue; Zheng, Shuilin; Du, Gaoxiang; Shu, Feng; Chen, Juntao

The Nano-TiO2/Diatomite compound photocatalyst is used to degrade rhodamine B dye wastewater in photochemical reactor. The test result indicates that the rate of photodegradation of rhodamine B is influenced by reactive conditions. The best technical conditions are concentration of rhodamine B solution 10mg/L, ultraviolet light 300W, the compound photocatalyst amount used 1g/L, the pH 5.8, reaction time 20min. Under these conditions the rate of photodegradation of rhodamine B may reach as high as 97.80%. And the efficiency of photodegradation of catalyst only has a little changed in recycling.

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1985-09-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

TMI-2 reactor vessel plenum final lift

International Nuclear Information System (INIS)

Wilson, D.C.

1986-01-01

Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

B11 NMR in the layered diborides OsB2 and RuB2

Science.gov (United States)

Suh, B. J.; Zong, X.; Singh, Y.; Niazi, A.; Johnston, D. C.

2007-10-01

B11 nuclear magnetic resonance (NMR) measurements have been performed on B11 enriched OsB2 and RuB2 polycrystalline powder samples in an external field of 4.7T and in the temperature range, 4.2KOsB2 and RuB2 , respectively. The experimental results indicate that a p character dominates the conduction electron wave function at the B site with a negligibly small s character in both compounds.

International working group on gas-cooled reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

1981-01-15

The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

Reactivity and neutron flux measurements in IPEN/MB-01 reactor with B4C burnable poison

International Nuclear Information System (INIS)

Fer, Nelson Custodio; Moreira, Joao Manoel Losada

2000-01-01

Burnable poison rods, made of B 4 C- Al 2 O 3 pellets with 5.01 mg/cm 3 10 B concentration, have been manufactured for a set of experiments in the IPEN/MB-01 zero-power reactor. Several core parameters which are affected by the burnable poisons rods have been measured. The principal results, for the situation in which the burnable poison rods are located near the absorber rods of a control rod, are they cause a 29% rod worth shadowing, a reduction of 39% in the local void coefficient of reactivity, a reduction of 4.8% in the isothermal temperature coefficient of reactivity, and a reduction of 9% in the thermal neutron flux in the region where the burnable poison rods are located. These experimental results will be used for the validation of burnable poison calculation methods in the CTMSP. (author)

Refurbishment programme for the BR2-reactor

Energy Technology Data Exchange (ETDEWEB)

Koonen, E [Centre d' Etude de l' Energie Nucleaire, Studiecentrum voor Kernenergie, BR2 Department, Boeretang, Mol (Belgium)

1992-07-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Refurbishment programme for the BR2-reactor

International Nuclear Information System (INIS)

Koonen, E.

1992-01-01

BR2 is a high flux engineering test reactor, which differs from comparable material testing reactors by its specific core array (fig. 1). It is a heterogeneous, thermal, tank-in-pool type reactor, moderated by beryllium and light water, which serves also as coolant. The fuel elements consist of cylindrical assemblies loaded in channels materialized by hexagonal beryllium prisms. The central 200 mm channel is vertical, while all others are inclined and form a hyperbolical arrangement around the central one. This feature combines a very compact core with the requirement of sufficient space for individual access to all channels through penetrations in the top cover of the aluminium pressure vessel. Each channel may hold a fuel element, a control rod, an experiment, an irradiation device or a beryllium plug. The refurbishment Program According to the present programme of C.E.N./S.C.K., BR2 will be in operation until 1996. At that time, the beryllium matrix will reach its foreseen end-of-life. In order to continue operation beyond this point, a thorough refurbishment of the reactor is foreseen, in addition to the unavoidable replacement of the matrix, to ensure quality of the installation and compliance with modern standards. Some fundamental options have been taken as a starting point: BR2 will continue to be used as a classical MTR, i.e. fuel and material irradiations and safety experiments with some additional service-activities. The present configuration is optimized for that use and there is no specific experimental requirement to change the basic concepts and performance characteristics. From the customers viewpoint, it is desirable to go ahead with the well-known features of BR2, to maintain a high degree of availability and reliability and to minimize the duration of the long shutdown. It is also important to limit the amount of nuclear liabilities. So the objective of the refurbishment programme is the life extension of BR2 for about 15 years, corresponding to

Safe Management Of Fast Reactors: Towards Sustainability

International Nuclear Information System (INIS)

Dreimanis, Andrejs

2015-01-01

An interdisciplinary systemic approach to socio-technical optimization of nuclear energy management is proposed, by recognizing a) the rising requirements to nuclear safety being realized using fast reactors (FR), b) the actuality to maintain and educate qualified workforce for fast reactors, c) the reactor safety and public awareness as the keystones for improving attitude to implement novel reactors. Knowledge management and informational support firstly is needed in: 1) technical issues: a) nuclear energy safety and reliability, b) to develop safe and economic technologies; 2) societal issues: a) general nuclear awareness, b) personnel education and training, c) reliable staff renascence, public education, stakeholder involvement, e).risk management. The key methodology - the principles being capable to manage knowledge and information issues: 1) a self-organization concept, 2) the principle of the requisite variety. As a primary source of growth of internal variety is considered information and knowledge. Following questions are analyzed indicating the ways of further development: a) threats in peaceful use of nuclear energy, b) basic features of nuclear risks, including terrorism, c) human resource development: basic tasks and instruments, d) safety improvements in technologies, e) advanced research and nuclear awareness improvement There is shown: public education, social learning and the use of mass media are efficient mechanisms forming a knowledge-creating community thereby reasoning to facilitate solution of key socio-technical nuclear issues: a) public acceptance of novel nuclear objects, b) promotion of adequate risk perception, and c) elevation of nuclear safety level and adequate risk management resulting in energetic and ecological sustainability. (author)

Evaluation 2 of B10 depletion in the WH PWR

International Nuclear Information System (INIS)

Park, Sang Won; Woo, Hae Suk; Kim, Sun Doo; Chae, Hee Dong; Myung, Sun Yup; Jang, Ju Kyung

2001-01-01

This paper presents the methodology to evaluate the B 10 depletion behavior in the pressurized water reactor. And B 10 depletion evaluation is performed based on the prediction program and the measured data of B 10 . The result shows that B 10 depletion during normal operation is not negligible. Therefore, adjustments for this depletion effect should be made to calculate the estimated critical postion(ECP) and determine the boron concentration required to maintain the specified shutdown margin

Refurbishing the BR2 materials testing reactor

International Nuclear Information System (INIS)

Baugnet, J.M.; Dekeyser, J.; Gubel, P.

1995-01-01

SCK/CEN is refurbishing its BR2 reactor to allow its further operation during the next 15 years; in doing so, it chooses to keep BR2 available for future scientific and technological irradiation programs within an international context. (author) 2 figs

Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

International Nuclear Information System (INIS)

Wheeler, J.K.; Pallotta, A.S.

1992-01-01

The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

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

International Nuclear Information System (INIS)

1966-01-01

This Volume 2 of the Proceedings of 2. Yugoslav symposium on reactor physics includes eight papers dealing with the following topics: method for measuring high anti reactivities of a reactor system; integration method for thermal reaction rate calculation; Determination of initial core configuration for BHWR-200 MWe; safety shutdowns and failures of the RA reactor equipment; determining the reactivity of absorption rods; measurements of thermal and fast neutron fluxes at the TRIGA reactor and other measurements during operation of the TRIGA reactor; mathematical modelling of the reactor safety; review of problems and methods for radiation risk assessment in the environment of a nuclear power plant

SIRIUS 2: A versatile medium power research reactor

International Nuclear Information System (INIS)

Rousselle, P.

1992-01-01

Most of the Research Reactors in the world have been critical in the Sixties and operated for twenty to thirty years. Some of them have been completely shut down, modified, or simply refurbished; the total number of RR in operation has decreased but there is still an important need for medium power research reactors in order: - to sustain a power program with fuel and material testing for NPP or fusion reactors; - to produce radioisotopes for industrial or medical purposes, doped silicon, NAA or neutron radiography; - to investigate further the condensed matter, with cold neutrons routed through neutron guides to improved equipment; - to develop new technologies and applications such as medical alphatherapy. Hence, taking advantage of nearly hundred reactor x years operation and backed up by the CEA experience, TECHNICATOME assisted by FRAMATOME has designed a new versatile multipurpose Research Reactor (20-30 Mw) SIRIUS 2 taking into account: - more stringent safety rules; - the lifetime; - the flexibility enabling a wide range of experiments and, - the future dismantling of the facility according to the ALARA criteria

H.B. Robinson-2 pressure vessel benchmark

Energy Technology Data Exchange (ETDEWEB)

Remec, I.; Kam, F.B.K.

1998-02-01

The H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark) is described and analyzed in this report. Analysis of the HBR-2 benchmark can be used as partial fulfillment of the requirements for the qualification of the methodology for calculating neutron fluence in pressure vessels, as required by the U.S. Nuclear Regulatory Commission Regulatory Guide DG-1053, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Section 1 of this report describes the HBR-2 benchmark and provides all the dimensions, material compositions, and neutron source data necessary for the analysis. The measured quantities, to be compared with the calculated values, are the specific activities at the end of fuel cycle 9. The characteristic feature of the HBR-2 benchmark is that it provides measurements on both sides of the pressure vessel: in the surveillance capsule attached to the thermal shield and in the reactor cavity. In section 2, the analysis of the HBR-2 benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed with three multigroup libraries based on ENDF/B-VI: BUGLE-93, SAILOR-95 and BUGLE-96. The average ratio of the calculated-to-measured specific activities (C/M) for the six dosimeters in the surveillance capsule was 0.90 {+-} 0.04 for all three libraries. The average C/Ms for the cavity dosimeters (without neptunium dosimeter) were 0.89 {+-} 0.10, 0.91 {+-} 0.10, and 0.90 {+-} 0.09 for the BUGLE-93, SAILOR-95 and BUGLE-96 libraries, respectively. It is expected that the agreement of the calculations with the measurements, similar to the agreement obtained in this research, should typically be observed when the discrete-ordinates method and ENDF/B-VI libraries are used for the HBR-2 benchmark analysis.

Neutronic study of the two french heavy water reactors

International Nuclear Information System (INIS)

Horowitz, J.

1955-01-01

The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.) [fr

MICROX-2 cross section library based on ENDF/B-VII

International Nuclear Information System (INIS)

Hou, J.; Ivanov, K.; Choi, H.

2012-01-01

New cross section libraries of a neutron transport code MICROX-2 have been generated for advanced reactor design and fuel cycle analyses. A total of 386 nuclides were processed, including 10 thermal scattering nuclides, which are available in ENDF/B-VII release 0 nuclear data. The NJOY system and MICROR code were used to process nuclear data and convert them into MICROX-2 format. The energy group structure of the new library was optimized for both the thermal and fast neutron spectrum reactors based on Contributon and Point-wise Cross Section Driven (CPXSD) method, resulting in a total of 1173 energy groups. A series of lattice cell level benchmark calculations have been performed against both experimental measurements and Monte Carlo calculations for the effective/infinite multiplication factor and reaction rate ratios. The results of MICROX-2 calculation with the new library were consistent with those of 15 reference cases. The average errors of the infinite multiplication factor and reaction rate ratio were 0.31% δk and 1.9%, respectively. The maximum error of reaction rate ratio was 8% for 238 U-to- 235 U fission of ZEBRA lattice against the reference calculation done by MCNP5. (authors)

OTUS - Reactor inventory management system based on ORIGEN2

Energy Technology Data Exchange (ETDEWEB)

Poellaenen, R; Toivonen, H; Lahtinen, J; Ilander, T

1995-10-01

ORIGEN2 is a computer code that calculates nuclide composition and other characteristics of nuclear fuel. The use of ORIGEN2 requires good knowledge in reactor physics. However, once the input has been defined for a particular reactor type, the calculations can be easily repeated for any burnup and decay time. This procedure produces large output files that are difficult to handle manually. A new computer code, known as OTUS, was designed to facilitate the postprocessing of the data. OTUS makes use of the inventory files precalculated with ORIGEN2 in a way that enables their versatile treatment for different safety analysis purposes. A data base is created containing a comprehensive set of ORIGEN2 calculations as a function of fuel burnup and decay time. OTUS is a reactor inventory management system for a microcomputer with Windows interface. Four major data operations are available: (1) Build data modifies ORIGEN2 output data into a suitable format, (2) View data enables flexible presentation of the data as such, (3) Different calculations, such as nuclide ratios and hot particle characteristics, can be performed for severe accident analyses, consequence analyses and research purposes, (4) Summary files contain both burnup dependent and decay time dependent inventory information related to the nuclide and the reactor specified. These files can be used for safeguards, radiation monitoring and safety assessment. (orig.) (22 refs., 29 figs.).

Marketing Optimization for B2B Market

OpenAIRE

Kaynova Tatyana V.

2012-01-01

The article presents market definition B2B, the necessity to optimize marketing B2B market, provides a system for B2B-marketing and developed stages of its formation. On this basis it was identified key factors of customer loyalty and are the stages of development of loyalty programs for customers market B2B.

The 2nd reactor core of the NS Otto Hahn

International Nuclear Information System (INIS)

Manthey, H.J.; Kracht, H.

1979-01-01

Details of the design of the 2nd reactor core are given, followed by a brief report summarising the operating experience gained with this 2nd core, as well as by an evaluation of measured data and statements concerning the usefulness of the knowledge gained for the development of future reactor cores. Quite a number of these data have been used to improve the concept and thus the specifications for the fuel elements of the 3rd core of the reactor of the NS Otto Hahn. (orig./HP) [de

Calculations on neutron irradiation damage in reactor materials

International Nuclear Information System (INIS)

Sone, Kazuho; Shiraishi, Kensuke

1976-01-01

Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

Catalytic Reactor for Inerting of Aircraft Fuel Tanks

Science.gov (United States)

1974-06-01

Aluminum Panels After Triphase Corrosion Test 79 35 Inerting System Flows in Various Flight Modes 82 36 High Flow Reactor Parametric Data 84 37 System...AD/A-000 939 CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS George H. McDonald, et al AiResearch Manufacturing Company Prepared for: Air Force...190th Street 2b. GROUP Torrance, California .. REPORT TITLE CATALYTIC REACTOR FOR INERTING OF AIRCRAFT FUEL TANKS . OESCRIP TIVE NOTEs (Thpe of refpoft

Possible physics modifications to CIRUS reactor core for improved reactor utilization

International Nuclear Information System (INIS)

John, Benjamin; Khosla, S.K.; Narain, Rajendra.

1976-01-01

Two fuelling schemes for uprating the neutron flux in CIRUS reactor at Trombay, are studied. One scheme employs enriched uranium-aluminium alloy boosters, the second envisages employing thorium oxide enriched with 0.2% plutonium oxide. It is seen that the second scheme has the potential of in-situ thorium utilization. (M.G.B.)

Addition of soluble and insoluble neutron absorbers to the reactor coolant system of TMI-2

International Nuclear Information System (INIS)

Hansen, R.F.; Silverman, J.; Queen, S.P.; Ryan, R.F.; Austin, W.E.

1984-07-01

The physical and chemical properties of six elements were studied and combined with cost estimates to determine the feasibility of adding them to the TMI-2 reactor coolant to depress k/sub eff/ to less than or equal to 0.95. Both soluble and insoluble forms of the elements B, Cd, Gd, Li, Sm, and Eu were examined. Criticality calculations were performed by Oak Ridge National Laboratory to determine the absorber concentration required to meet the 0.95 k/sub eff/ criterion. The conclusion reached is that all elements with the exception of boron have overriding disadvantages which preclude their use in this reactor. Solubility experiments in the reactor coolant show that boron solubility is the same as that of boron in pure aqueous solutions of sodium hydroxide and boric acid; consequently, solubility is not a limiting factor in reaching the k/sub eff/ criterion. Examination of the effect of pH on sodium requirements and costs for processing to remove radionuclides revealed a sharp dependence; small decreases in pH lead to a large decrease in both sodium requirements and processing costs. Boron addition to meet any contemplated reactor safety requirements can be accomplished with existing equipment; however, this addition must be made with the reactor coolant system filled and pressurized to ensure uniform boron concentration

AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

International Nuclear Information System (INIS)

Abbott, Michael L.; Daum, Keith A.

2011-01-01

This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

AGR-2 Data Qualification Report for ATR Cycles 147A, 148A, 148B, and 149A

Energy Technology Data Exchange (ETDEWEB)

Michael L. Abbott; Keith A. Daum

2011-08-01

This report presents the data qualification status of fuel irradiation data from the first four reactor cycles (147A, 148A, 148B, and 149A) of the on-going second Advanced Gas Reactor (AGR-2) experiment as recorded in the NGNP Data Management and Analysis System (NDMAS). This includes data received by NDMAS from the period June 22, 2010 through May 21, 2011. AGR-2 is the second in a series of eight planned irradiation experiments for the AGR Fuel Development and Qualification Program, which supports development of the very high temperature gas-cooled reactor (VHTR) under the Next Generation Nuclear Plant (NGNP) Project. Irradiation of the AGR-2 test train is being performed at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and is planned for 600 effective full power days (approximately 2.75 calendar years) (PLN-3798). The experiment is intended to demonstrate the performance of UCO (uranium oxycarbide) and UO2 (uranium dioxide) fuel produced in a large coater. Data qualification status of the AGR-1 experiment was reported in INL/EXT-10-17943 (Abbott et al. 2010).

Lattice dynamical and thermodynamical properties of ReB2, RuB2, and OsB2 compounds in the ReB2 structure

International Nuclear Information System (INIS)

Deligoz, E.; Colakoglu, K.; Ciftci, Y. O.

2012-01-01

Structural and lattice dynamical properties of ReB 2 , RuB 2 , and OsB 2 in the ReB 2 structure are studied in the framework of density functional theory within the generalized gradient approximation. The present results show that these compounds are dynamically stable for the considered structure. The temperature-dependent behaviors of thermodynamical properties such as internal energy, free energy, entropy, and heat capacity are also presented. The obtained results are in good agreement with the available experimental and theoretical data

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

Physics design of advanced steady-state tokamak reactor A-SSTR2

International Nuclear Information System (INIS)

Nishio, Satoshi; Ushigusa, Kenkichi

2000-10-01

Based on design studies on the fusion power reactor such as the DEMO reactor SSTR, the compact power reactor A-SSTR and the DREAM reactor with a high environmental safety and high availability, a new concept of compact and economic fusion power reactor (A-SSTR2) with high safety and high availability is proposed. Employing high temperature superconductor, the toroidal filed coils supplies the maximum field of 23T on conductor which corresponds to 11T at the magnetic axis. A-SSTR2 (R p =6.2m, a p =1.5m, I p =12MA) has a fusion power of 4GW with β N =4. For an easy maintenance and for an enough support against a strong electromagnetic force on coils, a poloidal coils system has no center solenoid coils and consists of 6 coils located on top and bottom of the machine. Physics studies on the plasma equilibrium, controllability of the configuration, the plasma initiation and non-inductive current ramp-up, fusion power controllability and the diverter have shown the validity of the A-SSTR2 concept. (author)

FORE-2, Thermohydraulics and Space-Independent Reactor Kinetics for Transients

International Nuclear Information System (INIS)

Fox, J.N.; Lawler, B.E.; Butz, H.R.; Heames, T.J.

1984-01-01

1 - Description of problem or function: FORE2 is a coupled thermal hydraulics-point kinetics digital computer code designed to calculate significant reactor parameters under steady-state conditions, or as functions of time during transients. The transients may result from a programmed reactivity insertion or a power change. Variable inlet coolant flow rate and temperature are considered. The code calculates the reactor power, the individual reactivity feedbacks, and the temperature of coolant, cladding, fuel, structure, and additional material for up to seven axial positions in three channel types which represent radial zones of the reactor. The heat of fusion, accompanying fuel melting, the liquid metal voiding reactivity, and the spatial and the time variation of the fuel cladding gap coefficient due to changes in gap size are considered. 2 - Method of solution: FORE2 input consists of property data, geometry, power and flow distribution factors, external time varying functions, experimental coefficients, and termination data. The differential equations for fluid flow, heat transfer, and point neutronics are solved by explicit finite-difference procedures. 3 - Restrictions on the complexity of the problem: Reactor excursions which can be calculated are restricted to those transients in which the reactor is not substantially destroyed. As a general rule, changes in reactor geometry and composition during an excursion are limited to those cases in which the reactivity effects of the changes may be considered as small perturbations of the initial system. Thus, accidents involving large-scale disassembly and bulk meltdown of a core are not covered by FORE2. FORE2 is valid only while the core retains its initial geometry

B2B myynnin johtaminen ravintola-alalla

OpenAIRE

Pajari, Katja

2014-01-01

Tämän opinnäytetyön tavoitteena oli selvittää, miten ravintola-alan yrityksissä johdetaan B2B myyntiä. Tarkoituksena oli kartoittaa, miten yrityksissä panostetaan B2B myyntiin ja sen johtamiseen sekä millä tavoin yritysmyyntiä johdetaan. Tutkimus pohjautuu opinnäytetyön tietoperustaan, jossa käsitellään B2B myyntiprosessia ja myynnin johtamista. Työssä käsitellään B2B myyntiä ja selvitetään myyntiprosessin eri vaiheita. Aihe on rajattu koskemaan nimenomaan johtamisen näkökulmaa B2B myynni...

Structural, mechanical, and electronic properties of TaB2, TaB, IrB2, and IrB: First-principle calculations

International Nuclear Information System (INIS)

Zhao Wenjie; Wang Yuanxu

2009-01-01

First-principle calculations were performed to investigate the structural, elastic, and electronic properties of TaB 2 , TaB, IrB 2 , and IrB. The calculated equilibrium structural parameters, shear modulus, and Young's modulus of TaB 2 are well consistent with the available experimental data, and TaB 2 with P6/mmm space group has stronger directional bonding between ions than WB 2 , OsB 2 , IrN 2 , and PtN 2 . For TaB 2 , the hexagonal P6/mmm structure is more stable than the orthorhombic Pmmn one, while for IrB 2 the orthorhombic Pmmn structure is the most stable one. The high shear modulus of P6/mmm phase TaB 2 is mainly due to the strong covalent π-bonding of B-hexagon in the (0001) plane. Such a B-hexagon network can strongly resist against an applied [112-bar0] (0001) shear deformation. Correlation between the hardness and the elastic constants of TaB 2 was discussed. The band structure shows that P6/mmm phase TaB 2 and Pmmn phase IrB 2 are both metallic. The calculations show that both TaB and IrB are elastically stable with the hexagonal P6 3 /mmc structure. - Elastic constant c 44 of TaB 2 is calculated to be 235 GPa. This value is exceptionally high, exceeding those of WB 2 , OsB 2 , WB 4 , OsN 2 , IrN 2 , and PtN 2 .

Maximization of burning and/or transmutation (B/T) capacity in coupled spectrum reactor (CSR) by fuel and core adjustment

International Nuclear Information System (INIS)

Aziz, F.; Kitamoto, Asashi.

1996-01-01

A conceptual design of burning and/or transmutation (B/T) reactor, based on a modified conventional 1150 MWe-PWR system, consisted of two core regions for thermal and fast neutrons, respectively, was proposed herein for the treatments of minor actinides (MA). In the outer region 237 Np, 241 Am, and 243 Am burned by thermal neutrons, while in the inner region 244 Cm was burned mainly by fast neutrons. The geometry of B/T fuel in the outer region was left the same with that of PWR, while in the inner region the B/T fuel was arranged in a tight-lattice geometry that allowed a higher fuel to coolant volume ratio. The maximization of B/T capacity in CSR were done by, first, increasing the radius of the inner region. Second, reducing the coolant to fuel volume ratio, and third, choosing a suitable B/T fuel type. The result of the calculations showed that the equilibrium of main isotopes in CSR can be achieved after about 5 recycle stages. This study also showed that the CSR can burn and transmute up to 808 kg of MA in a single reactor core effectively and safely. (author)

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

Keeping research reactors relevant: A pro-active approach for SLOWPOKE-2

International Nuclear Information System (INIS)

Cosby, L.R.; Bennett, L.G.I.; Nielsen, K.; Weir, R.

2010-01-01

The SLOWPOKE is a small, inherently safe, pool-type research reactor that was engineered and marketed by Atomic Energy of Canada Limited (AECL) in the 1970s and 80s. The original reactor, SLOWPOKE-1, was moved from Chalk River to the University of Toronto in 1970 and was operated until upgraded to the SLOWPOKE-2 reactor in 1973. In all, eight reactors in the two versions were produced and five are still in operation today, three having been decommissioned. All of the remaining reactors are designated as SLOWPOKE-2 reactors. These research reactors are prone to two major issues: aging components and lack of relevance to a younger audience. In order to combat these problems, one SLOWPOKE -2 facility has embraced a strategy that involves modernizing their reactor in order to keep the reactor up to date and relevant. In 2001, this facility replaced its aging analogue reactor control system with a digital control system. The system was successfully commissioned and has provided a renewed platform for student learning and research. The digital control system provides a better interface and allows flexibility in data storage and retrieval that was never possible with the analogue control system. This facility has started work on another upgrade to the digital control and instrumentation system that will be installed in 2010. The upgrade includes new computer hardware, updated software and a web-based simulation and training system that will allow licensed operators, students and researchers to use an online simulation tool for training, education and research. The tool consists of: 1) A dynamic simulation for reactor kinetics (e.g., core flux, power, core temperatures, etc). This tool is useful for operator training and student education; 2) Dynamic mapping of the reactor and pool container gamma and neutron fluxes as well as the vertical neutron beam tube flux. This research planning tool is used for various researchers who wish to do irradiations (e.g., neutron

Reactor lattice codes

International Nuclear Information System (INIS)

Kulikowska, T.

2001-01-01

The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

International Nuclear Information System (INIS)

Ritchie, L.T.; Johnson, J.D.; Blond, R.M.

1983-02-01

The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

1994-01-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

TRUST IN B2B E-MARKETPLACES

Directory of Open Access Journals (Sweden)

SEBASTIAN KOT

2011-01-01

Full Text Available The paper presents background of B2B exchanges and review of their forms and functionalities. The benefits and fails reasons are noticed. European enterprises interest in B2B trade is next aspect of consideration. Finally, the trust barriers of B2B exchanges are presented.

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

International Nuclear Information System (INIS)

2005-01-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Current status of restoration work for obstacle and upper core structure in reactor vessel of experimental fast reactor 'Joyo'. 2-2

International Nuclear Information System (INIS)

Okuda, Eiji; Ito, Hiromichi; Yoshihara, Shizuya

2014-01-01

An accident occurred in experimental fast reactor 'Joyo' in 2007 which is obstruction of fuel change equipment caused by contacting rotating plug and MARICO-2. In addition, we confirmed two happenings in the reactor vessel that (1) Deformation of MARICO-2 subassembly on the in vessel storage rack together with a transfer pot, (2) Deformation of the Upper core structure of 'Joyo' caused by contacting MARICO-2 subassembly and the UCS. We do the restoration work for restoring it. This time, we describe current status of Replacement work of the UCS. (author)

Ship propulsion reactors technology

International Nuclear Information System (INIS)

Fribourg, Ch.

2002-01-01

This paper takes the state of the art on ship propulsion reactors technology. The french research programs with the corresponding technological stakes, the reactors specifications and advantages are detailed. (A.L.B.)

Identification of SH2B2β as an Inhibitor for SH2B1- and SH2B2α-Promoted Janus Kinase-2 Activation and Insulin Signaling

OpenAIRE

Li, Minghua; Li, Zhiqin; Morris, David L.; Rui, Liangyou

2007-01-01

The SH2B family has three members (SH2B1, SH2B2, and SH2B3) that contain conserved dimerization (DD), pleckstrin homology, and SH2 domains. The DD domain mediates the formation of homo- and heterodimers between members of the SH2B family. The SH2 domain of SH2B1 (previously named SH2-B) or SH2B2 (previously named APS) binds to phosphorylated tyrosines in a variety of tyrosine kinases, including Janus kinase-2 (JAK2) and the insulin receptor, thereby promoting the activation of JAK2 or the ins...

Sterilization of E. coli bacterium in a flowing N2-O2 post-discharge reactor

International Nuclear Information System (INIS)

Villeger, S; Cousty, S; Ricard, A; Sixou, M

2003-01-01

Effective destruction of Escherichia coli (E. coli) bacteria has been obtained in a flowing N 2 -O 2 microwave post-discharge reactor. The sterilizing agents are the O atoms and the UV emissions of NOβ which are produced by N and O atoms recombination in the reactor. In the following plasma conditions: pressure 5 Torr, flow rate 1 L n min -1 , microwave power of 100 W in a quartz tube of 5 mm, an O atom density of 2.5x10 15 cm -3 is measured by NO titration in the post-discharge reactor with UV emission in a N 2 -(5%)O 2 gas mixture. Full destruction of 10 13 cfu ml -1 E. coli is observed after a treatment time of 25 min. (rapid communication)

Advanced gas-cooled reactors (AGR)

Energy Technology Data Exchange (ETDEWEB)

Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

1981-01-15

The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

Safe dismantling of the SVAFO research reactors R2 and R2-0 in Sweden

International Nuclear Information System (INIS)

ARNOLD, Hans-Uwe; BROY, Yvonne; Dirk Schneider

2017-01-01

The R2 and R2-0 reactors were part of the Swedish government's research program on nuclear power from the early 1960's. Both reactors were shut down in 2005 following a decision by former operator Studsvik Nuclear AB. The decommissioning of the R2 and R2-0 reactors is divided into three phases. The first phase - awarded to AREVA - involved dismantling of the reactors and associated systems in the reactor pool, treatment of the disassembled components as well as draining, cleaning and emptying the pool. In the second phase, the pool structure itself will be dismantled, while removal of remaining reactor systems, treatment and disposal of materials and clean-up will be carried out in the third stage. The entire work is planned to be completed before the end of this decade. The paper describes the several steps of phase 1 - starting with the team building, followed by the dismantling operations and covers challenges encountered and lessons learned as well. The reactors consist of 5.400 kg aluminum, 6.000 kg stainless steel restraint structures as well as, connection elements of the mostly flanged components (1.000 kg). The most demanding - from a radiological point of view - was the R2-0 reactor that was limited to ∼ 1 m"3 construction volumes but with an extremely heterogeneous activation profile. Based on the calculated radiological entrance data and later sampling, nuclide vectors for both reactors depending on the real placement of the single component and on the material (aluminum and stainless steel) were created. Finally, for the highest activated component from R2 reactor, 85 Sv/h were measured. The dismantling principles - adopted on a safety point of view - were the following: The always protected base area of the ponds served as a flexible buffer area for waste components and packaging. Specific protections were also installed on the walls to protect them from mechanical stress which may occur during dismantling work. A specific work platform was

Effect of U-238 and U-235 cross sections on nuclear characteristics of fast and thermal reactors

Energy Technology Data Exchange (ETDEWEB)

Akie, Hiroshi; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

1997-03-01

Benchmark calculation has been made for fast and thermal reactors by using ENDF/B-VI release 2(ENDF/B-VI.2) and JENDL-3.2 nuclear data. Effective multiplication factors (k{sub eff}s) calculated for fast reactors calculated with ENDF/B-VI.2 becomes about 1% larger than the results with JENDL-3.2. The difference in k{sub eff} is caused mainly from the difference in inelastic scattering cross section of U-238. In all thermal benchmark cores, ENDF/B-VI.2 gives smaller multiplication factors than JENDL-3.2. In U-235 cores, the difference is about 0.3%dk and it becomes about 0.6% in TCA U cores. The difference in U-238 data is also important in thermal reactors, while there are found 0.1-0.3% different v values of U isotopes in thermal energy between ENDF/B-VI.2 and JENDL-3.2. (author)

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

Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

International Nuclear Information System (INIS)

Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

2004-01-01

As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

PCU arrangement of a supercritical CO{sub 2} cooled micro modular reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Seong Gu; Baik, Seungjoon; Cho, Seong Kuk; Oh, Bong Seong; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

2016-05-15

As part of the SMR(Small Modular Reactor)s development effort, the authors propose a concept of supercritical CO{sub 2} (S-CO{sub 2}) cooled fast reactor combined with the S-CO{sub 2} Brayton cycle. The reactor concept is named as KAIST Micro Modular Reactor (MMR). The S-CO{sub 2} Brayton cycle has many strong points when it is used for SMR's power conversion unit. It occupies small footprints due to the compact cycle components and simple layout. Thus, a concept of one module containing the S-CO{sub 2} cooled fast reactor and power conversion system is possible. This module can be shipped via ground transportation (by trailer) or marine transportation. In this study, the authors propose a new conceptual layout for the S-CO{sub 2} cooled direct cycle while considering various issues for arranging cycle components. The new design has an improved cycle efficiency (from 31% to 34%) than the earlier version of MMR by reducing pressure drops in the heat exchangers. As a more efficient option, a recompression recuperated cycle was also designed. It improves 5% of thermal efficiency while 18tons of mass can be added in comparison to the simple recuperated cycle. Even if we adopt recompression cycle as a PCU, the weight of module (152tons) is less than the ground transportable limit (260tons)

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.

TMI-2 reactor vessel head removal

International Nuclear Information System (INIS)

Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

1984-12-01

This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

2017-05-15

Beginning-Of-Life (BOL) spatial distributions of the fission power in the core at 100 MW{sub th}, and the negative temperature reactivity feedback effects. Besides decreasing the UN fuel enrichment, other parameters examined are the materials of the followers to the B{sub 4}C rods for RSS and RC, the rods in the radial blanket assemblies, and the thickness of the scalloped BeO walls of the UN fuel and radial blanket assemblies. Despite ∼13% lower fuel enrichment (15.35%) and ∼22% lower hot-clean excess reactivity ($6.29 versus $8.06 for SLIMM-1.0 at 100 MW{sub th}), the operation life of the SLIMM-1.2 is ∼6.8% longer (6.3 versus 5.9 FPY for SLIMM-1.0), and the total negative temperature reactivity feedback is slightly smaller. At BOL, the radial blanket assemblies in ring 4, and the six UN fuel assembles in ring 1 of the SLIMM-1.2 core generate 2% and 23% of the total reactor thermal power, respectively. The twelve and eighteen UN fuel assemblies in rings 2 and 3 of the SLIMM-1.2 core generate 38% and 37% of the reactor thermal power, respectively. At EOL, the thermal power generation by the UN assemblies in rings 1 and 2 of the SLIMM-1.2 core decreases to 21.5% and 35.9%, respectively, while that by the assemblies in rings 3 and 4 increases to 37.6% and 5%, respectively.

Delayed photoneutrons of the 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; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

Neutronic study using oxide and nitride fuels for the Super Phenix 2 reactor

International Nuclear Information System (INIS)

Batista, J.L.; Renke, C.A.C.

1991-11-01

This report presents a neutronic analysis and a description of the Super Phenix 2 reactor, taken as reference. We present the methodology and results for cell and global reactor calculations for oxide (U O 2 - Pu O 2 ) and nitride (U N - Pu N) fuels. To conclude we compare the performance of oxide and nitride fuels for the reference reactor. (author)

Shadow corrosion evaluation in the Studsvik R2 reactor

International Nuclear Information System (INIS)

Sanders, Ch.; Lysell, G.

2000-01-01

Post-irradiation examination has shown that increased corrosion occurs when zirconium alloys are in contact with or in proximity to other metallic objects. The observations indicate an influence of irradiation from the adjacent component as the enhanced corrosion occurs as a 'shadow' of the metallic object on the zirconium surface. This phenomenon could ultimately limit the lifetime of certain zirconium alloy components in the reactor. The Studsvik R2 materials test reactor has an In-Core Autoclave (INCA) test facility especially designed for water chemistry and materials research. The INCA facility has been evaluated and found suitable for shadow corrosion studies. The R2 reactor core containing the INCA facility was modeled with the Monte Carlo N-Particle (MCNP) code in order to evaluate the electron deposition in various materials and to develop a hypothesis of the shadow corrosion mechanism. (authors)

Research nuclear reactor RA - Annual report 1992

International Nuclear Information System (INIS)

Sotic, O.

1992-12-01

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

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

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

Radionuclide distribution in TMI-2 reactor building basement liquids and solids

International Nuclear Information System (INIS)

Horan, J.T.; McIsaac, C.V.; Keefer, D.G.

1984-01-01

As a result of the TMI-2 accident, approximately 2.46 x 10 6 L of contaminated water were released to the Reactor Building basement. The principal fission product release pathway from the damaged core was through the reactor coolant system (RCS) to the pressurizer, through the pressure-operated relief valve (PORV) on the pressurizer to the Reactor Coolant Drain Tank (RCDT), and then through the RCDT rupture disk to the Reactor Building basement. Since August 1979, a number of efforts have been made to determine the location, quantity, and composition of fission products released to the Reactor Building basement. These efforts have included sampling of the basement water and solids, the basement sump pump recirculation line, the RCDT, and visual surveys using a closed circuit television (CCTV) system. The analysis of basement samples has provided data on the physical and radioisotopic characteristics of the liquids and solids. This paper describes the sample collection techniques and discusses radiochemical analyses results

An experimental investigation of fission product release in SLOWPOKE-2 reactors

International Nuclear Information System (INIS)

Harnden, A.M.C.

1995-09-01

Increasing radiation fields due to a release of fission products in the reactor container of several SLOWPOKE-2 reactors fuelled with a highly-enriched uranium (HEU) alloy core have been observed. It is believed that these increases are associated with the fuel fabrication where a small amount of uranium-bearing material is exposed to the coolant at the end-welds of the fuel element. To investigate this phenomenon samples of reactor water and gas from the headspace above the water have been obtained and examined by gamma spectrometry methods for reactors of various burnups at the University of Toronto, Ecole Polytechnique and Kanata Isotope Production Facility. An underwater visual examination of the fuel core at Ecole Polytechnique has also provided information on the condition of the core. This report (Volume 1) summarizes the equipment, analysis techniques and results of tests conducted at the various reactor sites. The data report is published as Volume 2. (author). 30 refs., 9 tabs., 20 figs

The Oak Ridge Research Reactor: safety analysis: Volume 2, supplement 2

International Nuclear Information System (INIS)

Hurt, S.S.

1986-11-01

The Oak Ridge Research Reactor Safety Analysis was last updated via ORNL-4169, Vol. 2, Supplement 1, in May of 1978. Since that date, several changes have been effected through the change-memo system described below. While these changes have involved the cooling system, the electrical system, and the reactor instrumentation and controls, they have not, for the most part, presented new or unreviewed safety questions. However, some of the changes have been based on questions or recommendations stemming from safety reviews or from reactor events at other sites. This paper discusses those changes which were judged to be safety related and which include revisions to the syphon-break system and changes related to seismic considerations which were very recently completed. The maximum hypothetical accident postulated in the original safety analysis requires dynamic containment and filtered flow for compliance with 10CFR100 limits at the site boundary

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

Loss of coolant analysis for the tower shielding reactor 2

International Nuclear Information System (INIS)

Radcliff, T.D.; Williams, P.T.

1990-06-01

The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

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

International Nuclear Information System (INIS)

Huda, M.Q.

2006-01-01

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

Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

International Nuclear Information System (INIS)

Chatt, A.; Holzbecher, J.

1990-01-01

SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

Thermal hydraulic and neutron kinetic simulation of the Angra 2 reactor using a RELAP5/PARCS coupled model

Energy Technology Data Exchange (ETDEWEB)

Reis, Patricia A.L.; Costa, Antonella L.; Hamers, Adolfo R.; Pereira, Claubia; Rodrigues, Thiago D.A.; Mantecon, Javier G.; Veloso, Maria A.F., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: adolforomerohamers@hotmail.com, E-mail: claubia@nuclear.ufmg.br, E-mail: thiagodanielbh@gmail.com, E-mail: mantecon1987@gmail.com, E-mail: dora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

2015-07-01

The computational advances observed in the last two decades have been provided direct impact on the researches related to nuclear simulations, which use several types of computer codes, including coupled between them, allowing representing with very accuracy the behavior of nuclear plants. Studies of complex scenarios in nuclear reactors have been improved by the use of thermal-hydraulic (TH) and neutron kinetics (NK) coupled codes. This technique consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into codes, mainly to simulate transients that involve asymmetric core spatial power distributions and strong feedback effects between neutronics and reactor thermal-hydraulics. Therefore, this work presents preliminary results of TH RELAP5 and the NK PARCS calculations applied to model of the Angra 2 reactor. The WIMSD-5B code has been used to generate the macroscopic cross sections used in the NK code. The results obtained are satisfactory and represent important part of the development of this methodology. The next step is to couple the codes. (author)

B2B oriented on-line applications generator

OpenAIRE

Vintilă Bogdan-Cătălin

2008-01-01

B2B applications are presented. Quality characteristics of B2B applications are defined. B2B application structure is defined. The application for contracts is developed. The advantages are identified.

Gas-cooled breeder reactor safety

Energy Technology Data Exchange (ETDEWEB)

Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

1981-01-15

The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

Mirror fusion reactor design

International Nuclear Information System (INIS)

Neef, W.S. Jr.; Carlson, G.A.

1979-01-01

Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

International Nuclear Information System (INIS)

Greene, S.R.; Spellman, D.J.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 2 - CANDU heavy water reactor alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Spellman, D.J.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 2 of a four volume report, summarizes the results of these analyses for the CANDU reactor based plutonium disposition alternative.

Dynamic simulation of the 2 MWt slowpoke heating reactor

International Nuclear Information System (INIS)

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

1982-04-01

A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

Distribution of energy of impulses of the modernized IBR-2 REACTOR

International Nuclear Information System (INIS)

Tayibov, L.A; Mehtiyeva, R.N.; )

2011-01-01

Full text: For the modernized IBR-2 reactor there are two main reasons causing fluctuations of energy of impulses [1,3] on low power of stochastic fluctuations, on the nominal - giving rise to fluctuations of external reactance. The fluctuations of pulse energy is quite significant (20%). They affect the dynamics of the reactor, the process of regulation, starting, as well as the work of the experimental apparatus, etc. It is clear that research of fluctuation of energy of impulses has special value for the IBR-2 type reactor. Sufficient information about the statistical properties of the reactor noise gives the density distribution of the energy pulse power. We used the usual procedure of statistical analysis of time series. Calculated pulse energy of density and the parameters of this distribution.

Tritium resources available for fusion reactors

Science.gov (United States)

Kovari, M.; Coleman, M.; Cristescu, I.; Smith, R.

2018-02-01

The tritium required for ITER will be supplied from the CANDU production in Ontario, but while Ontario may be able to supply 8 kg for a DEMO fusion reactor in the mid-2050s, it will not be able to provide 10 kg at any realistic starting time. The tritium required to start DEMO will depend on advances in plasma fuelling efficiency, burnup fraction, and tritium processing technology. It is in theory possible to start up a fusion reactor with little or no tritium, but at an estimated cost of 2 billion per kilogram of tritium saved, it is not economically sensible. Some heavy water reactor tritium production scenarios with varying degrees of optimism are presented, with the assumption that only Canada, the Republic of Korea, and Romania make tritium available to the fusion community. Results for the tritium available for DEMO in 2055 range from zero to 30 kg. CANDU and similar heavy water reactors could in theory generate additional tritium in a number of ways: (a) adjuster rods containing lithium could be used, giving 0.13 kg per year per reactor; (b) a fuel bundle with a burnable absorber has been designed for CANDU reactors, which might be adapted for tritium production; (c) tritium production could be increased by 0.05 kg per year per reactor by doping the moderator with lithium-6. If a fusion reactor is started up around 2055, governments in Canada, Argentina, China, India, South Korea and Romania will have the opportunity in the years leading up to that to take appropriate steps: (a) build, refurbish or upgrade tritium extraction facilities; (b) extend the lives of heavy water reactors, or build new ones; (c) reduce tritium sales; (d) boost tritium production in the remaining heavy water reactors. All of the alternative production methods considered have serious economic and regulatory drawbacks, and the risk of diversion of tritium or lithium-6 would also be a major concern. There are likely to be serious problems with supplying tritium for future

Generation and quenching of NF(a) and NF(b) molecules

International Nuclear Information System (INIS)

Setser, D.W.; Cha, H.; Quinones, E.; Du, K.

1987-01-01

The Ar( 3 Po,2) + NF 2 and 2F + HN 3 reactions have been developed as sources of NF(b 1 Σ + ) and NF(a 1 Δ) molecules, respectively, in a flow reactor. The decay kinetics for these molecules in the presence of added reagent can be studied using standard flow reactor techniques. Room temperature quenching rate constants for both molecules are reported for several reagents and compared to results for the isoelectronic O 2 (a) and O 2 (b) molecules

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.)

Description of the advanced gas cooled type of reactor (AGR)

Energy Technology Data Exchange (ETDEWEB)

Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

1996-11-01

The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

Description of the advanced gas cooled type of reactor (AGR)

International Nuclear Information System (INIS)

Nonboel, E.

1996-11-01

The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Reactor physics studies in the GCFR phase-II critical assembly

International Nuclear Information System (INIS)

Pond, R.B.

1976-09-01

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

A conceptual design of LIB fusion reactor: UTLIF(2)

International Nuclear Information System (INIS)

Madarame, Haruki; Kondo, Shunsuke; Iwata, Shuichi; Oka, Yoshiaki; Miya, Kenzo.

1984-01-01

UTLIF(2) is a conceptual design study on a light ion beam driven fusion reactor based on a concept of rod-bundle blanket. Survivability and maintainability of the first wall and the blanket are regarded as of major importance in the design. The blanket rod is composed of a thick tube which has enough stiffness, a thin wrapping wall which receives high heat flux, and liquid lithium which breeds tritium and removes generated heat. The rod can be pulled out from the outside of the reactor vessel, hence the replacement is very easy. Nuclear and thermal analysis have been made and the performance of the reactor has been shown to be satisfactory. (author)

The BR2 materials testing reactor. Past, ongoing and under-study upgradings

Energy Technology Data Exchange (ETDEWEB)

Baugnet, J M; Roedt, Ch de; Gubel, P; Koonen, E [Centre d' Etude de I' Energie Nucleaire, Studiecentrum voor Kernenergie, C.E.N./S.C.K., Mol (Belgium)

1990-05-01

The BR2 reactor (Mol, Belgium) is a high-flux materials testing reactor. The fuel is 93% {sup 235}U enriched uranium. The nominal power ranges from 60 to 100 MW. The main features of the design are the following: 1) maximum neutron flux, thermal: 1.2 x 10{sup 15} n/cm{sup 2} s; fast (E > 0.1 MeV) : 8.4 x 10{sup 14} n /cm{sup 2} s; 2) great flexibility of utilization: the core configuration and operation mode can be adapted to the experimental loading; 3) neutron spectrum tailoring; 4) availability of five 200 mm diameter channels besides the standard channels (84 mm diameter); 5) access to the top and bottom covers of the reactor authorizing the irradiation of loops. The reactor is used to study the behaviour of fuel elements and structural materials intended for future nuclear power stations of several types (fission and fusion). Irradiations are carried out in connection with performance tests up to very high burn-up or neutron fluence as well as for safety experiments, power cycling experiments, and generally speaking, tests under off-normal conditions. Irradiations for nuclear transmutation (production of high specific activity radio-isotopes and transplutonium elements), neutron-radiography, use of beam tubes for physics studies, and gamma irradiations are also carried out. The BR2 is used in support of Belgian programs, at the request of utilities, industry and universities and in the framework of international agreements. The paper reviews the past and ongoing upgrading and enhancement of reactor capabilities as well as those under study or consideration, namely with regard to: reactor equipment, fuel elements, irradiation facilities, reactor operation conditions and long-term strategy. (author)

TiO2 Solar Photocatalytic Reactor Systems: Selection of Reactor Design for Scale-up and Commercialization—Analytical Review

Directory of Open Access Journals (Sweden)

Yasmine Abdel-Maksoud

2016-09-01

Full Text Available For the last four decades, viability of photocatalytic degradation of organic compounds in water streams has been demonstrated. Different configurations for solar TiO2 photocatalytic reactors have been used, however pilot and demonstration plants are still countable. Degradation efficiency reported as a function of treatment time does not answer the question: which of these reactor configurations is the most suitable for photocatalytic process and optimum for scale-up and commercialization? Degradation efficiency expressed as a function of the reactor throughput and ease of catalyst removal from treated effluent are used for comparing performance of different reactor configurations to select the optimum for scale-up. Comparison included parabolic trough, flat plate, double skin sheet, shallow ponds, shallow tanks, thin-film fixed-bed, thin film cascade, step, compound parabolic concentrators, fountain, slurry bubble column, pebble bed and packed bed reactors. Degradation efficiency as a function of system throughput is a powerful indicator for comparing the performance of photocatalytic reactors of different types and geometries, at different development scales. Shallow ponds, shallow tanks and fountain reactors have the potential of meeting all the process requirements and a relatively high throughput are suitable for developing into continuous industrial-scale treatment units given that an efficient immobilized or supported photocatalyst is used.

Digitaalisen markkinoinnin suunnitelma b2b-yritykselle

OpenAIRE

Harhakoski, Oskari

2011-01-01

Työ käsittelee digitaalisen markkinoinnin suunnitelman tekemistä b2b-yritykselle. Tavoitteena oli kilpailuedun hankkiminen sosiaalisen median tehokkaalla hyödyntämisellä mark-kinoinnissa. Konkreettisemmin yritys halusi lisää näkyvyyttä ja myyntiä. Suunnitelman laatimisessa hyödynnettiin POST-menetelmää. Erityistä huomiota kiinnitettiin b2b-markkinoinnin eroihin b2c-markkinointiin verrattuna. Myös yrityksen toimiminen Suomen markkinoilla huomioitiin. Lisäksi analysoitiin kilpailijoita asia...

Planned Scientific programs around the Triga Mark 2 Reactor

International Nuclear Information System (INIS)

Majah, M Ibn.

2007-01-01

Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

KAFAX-F22 : development and benchmark of multi-group library for fast reactor using JEF-2.2. Neutron 80 group and Photon 24 group

International Nuclear Information System (INIS)

Kim, Jung Do; Gil, Choong Sup.

1997-03-01

The KAFAX-F22 was developed from JEF-2.2, which is a MATXS format, multigroup library of fast reactor. The KAFAX-F22 has 80 and 24 energy group structures for neutron and photon, respectively. It includes 89 nuclide data processed by NJOY94.38. The TRANSX/TWODANT system was used for benchmark calculations of fast reactor and one- and two-dimensional calculations of ONEDANT and TWODANT were carried out with 80 group, P 3 S 16 and with 25 group, P 3 S 8 , respectively. The average values of multiplication factors are 0.99652 for MOX cores, 1.00538 for uranium cores and 1.00032 for total cores. Various central reaction rate ratios also give good agreements with the experimental values considering experimental uncertainties except for VERA-11A, VERA-1B, ZPR-6-7 and ZPR-6-6A cores of which experimental values seem to involve some problems. (author). 13 refs., 18 tabs., 2 figs

Inspection of the Sizewll 'B' reactor coolant pump flywheels

International Nuclear Information System (INIS)

McNulty, A.L.; Cheshire, A.

1992-01-01

The Sizewell ''B'' safety case has categorised some primary circuit items as components for which failure is considered to be incredible. These Incredibility of Failure (IOF) components are particularly critical in their safety function, and specially stringent and all embracing provisions are made in their design, manufacture, inspection and operation. These provisions are such as to limit the probability of failure to levels which are so low that it does not have to be taken into account and no steps are necessary to control the consequences. The reactor coolant pump flywheel is considered to be an IOF component. Consequently there is a need for rigorous inspection during both manufacture and in service (ISI). The ISI requirement results in the need for an automated inspection. There is therefore a prerequisite to perform a Pre-Service Inspection (PSI) for baseline fingerprinting purposes. Furthermore there is a requirement that the inspection procedure, the inspection equipment and the operators are validated at the Inspection Validation Centre (IVC) of the AEA Technology laboratories at Risley. Development work is described. (author)

Power noise spectrum classification in the problem of the IBR-2 reactor

International Nuclear Information System (INIS)

Bargel, M.; Kitowski, J.; Pepelyshev, Yu.N.

1988-01-01

The classification spectrum results of random fluctuations in the IBR-2 energy pulse are presented. The work is performed for the application of the obtained results to the reactor diagnostics and the study of its noise uncontrolled states. For classification of the spectra the method of pattern recognition based upon the ISODATA heuristic algorithm is used. It is shown that a set of noise uncontrolled reactor states, registered during the reactor operation period at power of 0.4-2 MVt with the first variant of moving reflector (1983-1986) is formed into 4(5) most typical states. Each of the states corresponds to the general conditions of the reactor core cooling and provides the normal work of the moving reflector. However, these states differ in coolant flow, power level and peculiarities of the moving reflector rotation regime. One type of anomal power noise, connected with some disorder in the moving reflctor work, is isolated. This work also presents the possibility of control over the state of moving reflectors according to the change in the amplitude of power oscillations at some frequences. The reactor noise classification results can be used as the data bank for the IBR-2 reactor diagnostic system

The experimental and technological developments reactor

International Nuclear Information System (INIS)

Carbonnier, J.L.

2003-01-01

THis presentation concerns the REDT, gas coolant reactor for experimental and technological developments. The specifications and the research programs concerning this reactor are detailed;: materials, safety aspects, core physic, the corresponding fuel cycle, the reactor cycle and the program management. (A.L.B.)

Research reactor FR2 - 20 years chemical and radiochemical measurements

International Nuclear Information System (INIS)

Feuerstein, H.; Graebner, H.; Oschinski, J.; Hoffmann, W.; Beyer, J.

1986-09-01

The FR2 has been a D 2 O cooled and moderated research reactor with a thermal output of 44 MW. It was in operation from 1961 to 1981. Because of the operating conditions of the reactor, only a small number of routine measurements were performed. For these however special techniques had to be developed. During the 20 years of operation a number of special events occured or have been observed, sometimes with very amazing results, e.g. the 'aceton effect'. This report describes the chemical and radiochemical conditions of the reactor systems, as well as the results of the surveilance work. Not described are measurements for the many experiments. The last chapter gives in a short form a description of the most unusual events and observations. (orig.) [de

Venting krypton-85 from the Three Mile Island Unit 2 reactor building

International Nuclear Information System (INIS)

Burton, H.M.

1981-01-01

To permit the less restricted access to the reactor building necessary to maintain instrumentation and equipment, and to proceed towad the total decontamination of the facility, General Public Utilities, operators of the facility referred to hereafter as GPU, asked the United States Nuclear Regulatory Commission, or NRC, for permission to remove the 85 Kr from the reactor building by venting it to the environment. GPU supported their request with the Safety Analysis and Environmental Assessment Report on the proposed reactor building venting plan. On June 12, 1980, after seven months of licensing deliberations and numerous public hearings, the NRC granted GPU's request. The actual venting took place between June 28 and July 11, 1980. This report presents an overview of the detailed effort involved in the TMI-2 reactor building venting program. The findings reported here are condensed from a published report entitled TMI-2 Reactor Building Purge--Kr-85 Venting

Research on economics and CO2 emission of magnetic and inertial fusion reactors

International Nuclear Information System (INIS)

Mori, Kenjiro; Yamazaki, Kozo; Oishi, Tetsutarou; Arimoto, Hideki; Shoji, Tatsuo

2011-01-01

An economical and environment-friendly fusion reactor system is needed for the realization of attractive power plants. Comparative system studies have been done for magnetic fusion energy (MFE) reactors, and been extended to include inertial fusion energy (IFE) reactors by Physics Engineering Cost (PEC) system code. In this study, we have evaluated both tokamak reactor (TR) and IFE reactor (IR). We clarify new scaling formulas for cost of electricity (COE) and CO 2 emission rate with respect to key design parameters. By the scaling formulas, it is clarified that the plant availability and operation year dependences are especially dominant for COE. On the other hand, the parameter dependences of CO 2 emission rate is rather weak than that of COE. This is because CO 2 emission percentage from manufacturing the fusion island is lower than COE percentage from that. Furthermore, the parameters dependences for IR are rather weak than those for TR. Because the CO 2 emission rate from manufacturing the laser system to be exchanged is very large in comparison with CO 2 emission rate from TR blanket exchanges. (author)

Development of a TiO2-coated optical fiber reactor for water decontamination

International Nuclear Information System (INIS)

Danion, A.

2004-09-01

The objective of this study was to built and to study a photo-reactor composed by TiO 2 -coated optical fibers for water decontamination. The physico-chemical characteristics and the optical properties of the TiO 2 coating were first studied. Then, the influences of different parameters as the coating thickness, the coating length and the coating volume were investigated both on the light transmission in the TiO 2 - coated fiber and on the photo-catalytic activity of the fiber for a model compound (malic acid). The photo-catalytic degradation of malic acid was optimized using the experimental design methodology allowing to build a multi-fiber reactor comprising 57 optical fibers. The photo-degradation of malic acid was conducted in the multi-fiber reactor and it was demonstrated that the multi-fiber reactor was more efficient than the single-fiber reactor at the same fibers density. Finally, the multi-fiber reactor was applied to the photo-degradation of a fungicide, called fenamidone, and a degradation pathway was proposed. (author)

Calculation of the geometric buckling for reactors of various shapes

Energy Technology Data Exchange (ETDEWEB)

Sjoestrand, N E

1958-05-15

A systematic investigation is made of the eleven coordinate systems in which the reactor equation {nabla}{sup 2}{phi} + B{sup 2}{phi} = 0 is separable. The fundamental solution and geometric buckling are given for those cases where the separated equations lead to known functions. It is especially shown that reactors of prolate and oblate spheroidal shape can be calculated in detail, and the results are given in extensive tables.

Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO

Energy Technology Data Exchange (ETDEWEB)

Cajko, Frantisek; Secansky, Michal; Chrebet, Tomas; Zajac, Radoslav; Darilek, Petr [VUJE, a.s., Trnava (Slovakia)

2016-09-15

Experimental reactor ALLEGRO is a gas cooled fast reactor in the design stage. The current design of its reactivity control system is based on control rods filled with boron carbide as the absorber. Because of disadvantages connected to high boron enrichment a possibility of using other absorbent materials was explored to lower the boron enrichment and increase the worth of the control rods. The results of neutronic Monte-Carlo analyses in a computational supercell are presented in this paper. Three absorbent materials most suitable for a use in reactor ALLEGRO (B{sub 4}C, EuB{sub 6} and ReB{sub 2}) have been analysed also in a full core model. A possible benefit of a neutron trap concept is explored as well but materials with satisfactory neutronic properties proved to be not suitable for expected high temperatures in the reactor.

Preparing the construction of a school reactor

International Nuclear Information System (INIS)

Matejka, K.

1977-01-01

The possibilities are discussed of teaching and training nuclear reactor operation and control, teaching experimental reactor physics and investigating reactor lattice parameters using a training reactor to be installed at the Faculty of Nuclear Science and Physical Engineering in Prague. Requirements are indicated for the reactor's technical design and the Faculty's possibilities to contribute to its construction. (J.B.)

TMI-2 reactor-vessel head removal and damaged-core-removal planning

International Nuclear Information System (INIS)

Logan, J.A.; Hultman, C.W.; Lewis, T.J.

1982-01-01

A major milestone in the cleanup and recovery effort at TMI-2 will be the removal of the reactor vessel closure head, planum, and damaged core fuel material. The data collected during these operations will provide the nuclear power industry with valuable information on the effects of high-temperature-dissociated coolant on fuel cladding, fuel materials, fuel support structural materials, neutron absorber material, and other materials used in reactor structural support components and drive mechanisms. In addition, examination of these materials will also be used to determine accident time-temperature histories in various regions of the core. Procedures for removing the reactor vessel head and reactor core are presented

Bu-2470, a new peptide antibiotic complex. II. Structure determination of Bu-2470 A, B1, B2a and B2b.

Science.gov (United States)

Sugawara, K; Yonemoto, T; Konishi, M; Matsumoto, K; Miyaki, T; Kawaguchi, H

1983-06-01

The structures of Bu-2470 A, B1, B2a, and B2b have been determined. Bu-2470 A is a simple octapeptide having no fatty acid moiety, while Bu-2470 B1, B2a and B2b are octapeptides that have been acylated with a beta-hydroxy C11 or C10 fatty acid. The octapeptide structure of Bu-2470 components was found identical with that of octapeptin C1, hence generic names of octapeptin C0, C2, C3 and C4 are proposed for Bu-2470 A, B1, B2a and B2b, respectively.

Boron neutron capture therapy (BNCT) translational studies in the hamster cheek pouch model of oral cancer at the new ''B2'' configuration of the RA-6 nuclear reactor

International Nuclear Information System (INIS)

Monti Hughes, Andrea; Trivillin, Veronica A.; Schwint, Amanda E.; Longhino, Juan; Boggio, Esteban; Medina, Vanina A.; Martinel Lamas, Diego J.; Garabalino, Marcela A.; Heber, Elisa M.; Pozzi, Emiliano C.C.; Itoiz, Maria E.; Aromando, Romina F.; Nigg, David W.

2017-01-01

Boron neutron capture therapy (BNCT) is based on selective accumulation of B-10 carriers in tumor followed by neutron irradiation. We demonstrated, in 2001, the therapeutic effect of BNCT mediated by BPA (boronophenylalanine) in the hamster cheek pouch model of oral cancer, at the RA-6 nuclear reactor. Between 2007 and 2011, the RA-6 was upgraded, leading to an improvement in the performance of the BNCT beam (B2 configuration). Our aim was to evaluate BPA-BNCT radiotoxicity and tumor control in the hamster cheek pouch model of oral cancer at the new ''B2'' configuration. We also evaluated, for the first time in the oral cancer model, the radioprotective effect of histamine against mucositis in precancerous tissue as the dose-limiting tissue. Cancerized pouches were exposed to: BPA-BNCT; BPA-BNCT + histamine; BO: Beam only; BO + histamine; CONTROL: cancerized, no-treatment. BNCT induced severe mucositis, with an incidence that was slightly higher than in ''B1'' experiments (86 vs 67%, respectively). BO induced low/moderate mucositis. Histamine slightly reduced the incidence of severe mucositis induced by BPA-BNCT (75 vs 86%) and prevented mucositis altogether in BO animals. Tumor overall response was significantly higher in BNCT (94-96%) than in control (16%) and BO groups (9-38%), and did not differ significantly from the ''B1'' results (91%). Histamine did not compromise BNCT therapeutic efficacy. BNCT radiotoxicity and therapeutic effect at the B1 and B2 configurations of RA-6 were consistent. Histamine slightly reduced mucositis in precancerous tissue even in this overly aggressive oral cancer model, without compromising tumor control. (orig.)

Boron neutron capture therapy (BNCT) translational studies in the hamster cheek pouch model of oral cancer at the new ''B2'' configuration of the RA-6 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Monti Hughes, Andrea; Trivillin, Veronica A.; Schwint, Amanda E. [Constituyentes Atomic Center, National Atomic Energy Commission (CNEA), Department of Radiobiology, San Martin, Province Buenos Aires (Argentina); National Research Council (CONICET), Ciudad Autonoma de Buenos Aires (Argentina); Longhino, Juan; Boggio, Esteban [Bariloche Atomic Center, CNEA, Department of Nuclear Engineering, San Carlos de Bariloche, Province Rio Negro (Argentina); Medina, Vanina A.; Martinel Lamas, Diego J. [National Research Council (CONICET), Ciudad Autonoma de Buenos Aires (Argentina); Pontifical Catholic University of Argentina (UCA), Laboratory of Tumoral Biology and Inflammation, School of Medical Sciences, Institute for Biomedical Research (BIOMED CONICET-UCA), Ciudad Autonoma de Buenos Aires (Argentina); Garabalino, Marcela A.; Heber, Elisa M.; Pozzi, Emiliano C.C. [Constituyentes Atomic Center, National Atomic Energy Commission (CNEA), Department of Radiobiology, San Martin, Province Buenos Aires (Argentina); Itoiz, Maria E. [Constituyentes Atomic Center, National Atomic Energy Commission (CNEA), Department of Radiobiology, San Martin, Province Buenos Aires (Argentina); UBA, Department of Oral Pathology, Faculty of Dentistry, Ciudad Autonoma de Buenos Aires (Argentina); Aromando, Romina F. [UBA, Department of Oral Pathology, Faculty of Dentistry, Ciudad Autonoma de Buenos Aires (Argentina); Nigg, David W. [Idaho National Laboratory, Idaho Falls (United States)

2017-11-15

Boron neutron capture therapy (BNCT) is based on selective accumulation of B-10 carriers in tumor followed by neutron irradiation. We demonstrated, in 2001, the therapeutic effect of BNCT mediated by BPA (boronophenylalanine) in the hamster cheek pouch model of oral cancer, at the RA-6 nuclear reactor. Between 2007 and 2011, the RA-6 was upgraded, leading to an improvement in the performance of the BNCT beam (B2 configuration). Our aim was to evaluate BPA-BNCT radiotoxicity and tumor control in the hamster cheek pouch model of oral cancer at the new ''B2'' configuration. We also evaluated, for the first time in the oral cancer model, the radioprotective effect of histamine against mucositis in precancerous tissue as the dose-limiting tissue. Cancerized pouches were exposed to: BPA-BNCT; BPA-BNCT + histamine; BO: Beam only; BO + histamine; CONTROL: cancerized, no-treatment. BNCT induced severe mucositis, with an incidence that was slightly higher than in ''B1'' experiments (86 vs 67%, respectively). BO induced low/moderate mucositis. Histamine slightly reduced the incidence of severe mucositis induced by BPA-BNCT (75 vs 86%) and prevented mucositis altogether in BO animals. Tumor overall response was significantly higher in BNCT (94-96%) than in control (16%) and BO groups (9-38%), and did not differ significantly from the ''B1'' results (91%). Histamine did not compromise BNCT therapeutic efficacy. BNCT radiotoxicity and therapeutic effect at the B1 and B2 configurations of RA-6 were consistent. Histamine slightly reduced mucositis in precancerous tissue even in this overly aggressive oral cancer model, without compromising tumor control. (orig.)

Development of a methodology for simulation of gas cooled reactors with purpose of transmutation

International Nuclear Information System (INIS)

Silva, Clarysson Alberto da

2009-01-01

This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k eff , neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

Fission product poisoning in KS-150 reactor operation

International Nuclear Information System (INIS)

Rana, S.B.

1978-01-01

A three-dimensional model of the KS-150 reactor was used to study reactivity changes induced by reactor poisoning with fission products Xe 135 and Sm 149 . A comparison of transients caused by the poisoning showed the following differences: (1) the duration of the transient Xe poisoning (2 days) is shorter by one order of magnitude than the duration of Sm poisoning (20 days); however, the level of Xe poisoning is greater approximately by one order than the level of the Sm poisoning; (2) the level of steady-state Xe poisoning depends on the output level of the reactor; steady-state Sm poisoning does not depend on this level; (3) following reactor shutdown Xe poisoning may increase to the maximum value of up to Δrhosub(Xe)=20% and will then gradually decrease; Sm poisoning may reach maximum values of up to Δrhosub(Sm)=2% and does not decrease. (J.B.)

Design of a reactor system for the synthesis of titanium diboride

International Nuclear Information System (INIS)

Tsui, M.E.; Epstein, H.A.

1981-10-01

TiB 2 , a hard, refractory material, is difficult to produce at a purity required for many potential uses. In this study, a laboratory-scale reactor system was designed to produce 4 g/h of very pure TiB 2 powder from a homogeneous-nucleation gas-phase reaction. The system operates at temperatures up to 1700 K, pressures from 1 torr to 1 atm, and incorporates a novel flame-reactor concept in which the heat of reaction for powder formation is provided by a H 2 -Cl 2 flame. The powder is produced in an alumina reactor 3-3/8-in. -ID x 55-in. long, with a feed preheater, and is collected in low-pressure traps. The system is fully instrumented for study of reaction kinetics and powder morphology. System startup, operating, shutdown, and safety procedures as well as a proposed experimental plan are included. The estimated construction cost of the system is $24,500

Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

Investigation of small and modular-sized fast reactor

International Nuclear Information System (INIS)

Kubota, Kenichi; Kawasaki, Nobuchika; Umetsu, Yoichiro; Akatsu, Minoru; Kasai, Shigeo; Konomura, Mamoru; Ichimiya, Masakazu

2000-06-01

In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

Energy Technology Data Exchange (ETDEWEB)

1992-04-01

This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

Fissile fuel production and usage of thermal reactor waste fueled with UO2 by means of hybrid reactor system

International Nuclear Information System (INIS)

Ipek, O.

1997-01-01

The use of Fast Breeder Reactors to produce fissile fuel from nuclear waste and the operation of these reactors with a new neutron source are becoming today' topic. In the thermonuclear reactors, it is possible to use 2.45-14.1 MeV - neutrons which can be obtained by D-T, D-D Semicatalyzed (D-D) and other fusion reactions. To be able to do these, Hybrid Reactor System, which still has experimental and theoretical studies, have to be taken into consideration.In this study, neutronic analysis of hybrid blanket with grafit reflector, is performed. D-T driven fusion reaction is surrounded by UO 2 fuel layer and the production of ''2''3''9Pu fissile fuel from waste ''2''3''8U is analyzed. It is also compared to the other possible fusion reactions. The results show that 815.8 kg/year ''2''3''8Pu with D-T reaction and 1431.6 kg/year ''2''3''8Pu with semicatalyzed (D-D) reaction can be produced for 1000 MW fusion power. This means production of 2.8/ year and 4.94/ year LWR respectively. In addition, 1000 MW fusion flower is is multiplicated to 3415 MW and 4274 MW for D-T and semicatalyzed (D-D) reactions respectively. The system works subcritical and these values are 0.4115 and 0.312 in order. The calculations, ANISN-ORNL code, S 16 -P 3 approach and DLC36 data library are used

Reactor group constants and benchmark test

Energy Technology Data Exchange (ETDEWEB)

Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

Neutron noise measurement technique in a coupled reactor

International Nuclear Information System (INIS)

Genoud, J.P.

1976-01-01

Describes work carried out on the swimming pool reactor at the Physikalisch-Technische Bundesanstalt at Braunschweig. The reactor has two multiplying zones, is light water moderated, with 90% enriched 235 U fuel. There is a D 2 0 reservoir between the two parts of the reactor. Signal/noise ratio obtained by means of ionisation chamber type neutron detectors of 10 -13 amp/u.f. sensitivity is of the order of 40 dB and band frequency 1.5 kHz. Spectral density of the interzone interaction energy was obtained by use of Fourier transforms, previously corrected by a Hanning window. (S.W.)

A theoretical analysis of methanol synthesis from CO2 and H2 in a ceramic membrane reactor

NARCIS (Netherlands)

Gallucci, F.; Basile, A.

2007-01-01

In this theoretical work the CO2 conversion into methanol in both a traditional reactor (TR) and a membrane reactor (MR) is considered. The purpose of this study was to investigate the possibility of increasing CO2 conversion into methanol with respect to a TR. A zeolite MR, able to combine

E-Business Models in B2B: Process Based Categorization and Analysis of B2B Models

OpenAIRE

Mahesh S. Raisinghani; Turan Melemez; Lijie Zou; Chris Paslowski; Irma Kimvidze; Susanne Taha; Klaus Simons

2005-01-01

The business models in business-to-business (B2B) e-commerce and their effectiveness have been a major topic of research in recent years. Due to the variety of existing models, it seems difficult to find a widely accepted categorization that can be analyzed and assessed. An in-depth study that provides a process-based approach to B2B e-commerce is presented and illustrated with examples from industry. A comparative examination of both the buy and the sell side based on a process-related appro...

Pressurised water reactor operation

International Nuclear Information System (INIS)

Birnie, S.; Lamonby, J.K.

1987-01-01

The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

Cronos 2: a neutronic simulation software for reactor core calculations

International Nuclear Information System (INIS)

Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

1999-01-01

The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

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)

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.

Manufacture of components for Canadian reactor programs

International Nuclear Information System (INIS)

Perry, L.P.

Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

Strippable coating used for the TMI-2 reactor building decontamination