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Sample records for cornell university zero power reactor

  1. Safety evaluation report related to the renewal of the operating license for the Zero-Power Reactor at Cornell University, Docket No. 50-97

    International Nuclear Information System (INIS)

    1983-09-01

    This Safety Evaluation Report for the application filed by Cornell University (CU) for a renewal of Operating License R-80 to continue to operate a zero-power reactor (ZPR) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York. The staff concludes that the ZPR facility can continue to be operated by CU without endangering the health and safety of the public

  2. License renewal and power upgrade of the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Aderhold, Howard C.

    1984-01-01

    The Cornell Mark II TRIGA reactor has been a principal facility for instruction and research in nuclear science and engineering at Cornell, and it has been extensively used by other departments at Cornell and by nearby universities and industries. Initially the fuel was low hydride, 8.5w/o 19%-enriched, aluminum clad; in 1974 it was changed to high-hydride, stainless-steel-clad. The maximum power has been 100 kW, with pulses to $2, and operation has been on a one-shift demand basis. Annual energy generation of 50 MWH has been typical. Standard features include a 4-inch tangential port and our 6-inch radial ports, a thermal column with hohlraum and vertical access, a central thimble, a 'rabbit', and a set of dry irradiation tubes, replacing the 'Lazy Susan'. The license was renewed and amended in November 1983; the new limits are 500 kW and $3 pulses. Physical changes to the facility included addition of a water-to-water heat exchanger and of a diffuser at the water outlet ∼ 60 cm above the core. The flow rate is 300 liters per minute in the primary (reactor) side of the heat exchanger. The temperature of the chilled water entering the secondary of the exchanger is ∼ 12?C; its flow rate is adjusted by a servo-controlled by-pass valve to maintain the desired range of pool water temperature. Steps taken to go to higher power included rearrangement of fuel elements to increase excess reactivity, recalibration of control rods, and power vs ion chamber current calibrations at successively higher power by comparing the rate of rise of pool temperature with a known rate using electrical heating elements. Steady-state operation has been done up to 480 kW (nominal) but pulsing at the newly allowed higher levels has not been tested as yet

  3. Use of research reactors in multidisciplinary education at Cornell University

    International Nuclear Information System (INIS)

    Clark, D.D.

    1992-01-01

    Multidisciplinary aspects of nuclear science and technology form a large part of the research and teaching activities of the Nuclear Science and Engineering (NS and E) Program at Cornell, and the two reactors housed in Ward Laboratory - a 500-kW TRIGA and a 100-W critical facility [zero-power reactor (ZPR)]- play a central role in those activities. Several primarily educational and multidisciplinary features of the NS and E program are described in this paper

  4. Programs with societal benefits at the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Clark, D.D.; Aderhold, H.C.; Hossain, T.Z.

    1993-01-01

    In its 30 yr of operation, the Cornell TRIGA reactor has been used for many educational and research programs that provide general benefits to society. In addition to supporting graduate-level education of nuclear scientists and engineers, it has been extensively used in undergraduate and graduate courses and research by nonspecialists and, through the medium of tours, in education of the general public. Some educational functions have been described previously. In this paper, examples are presented of research of societal interest in nonnuclear fields. The first two rely mainly on radiography, and the remaining five on neutron activation analysis (NAA)

  5. Preparation of mandatory documentation before the start up of the RA-0 'zero power' nuclear reactor at Cordoba National University

    International Nuclear Information System (INIS)

    Martin, H.R.; Keil, W.M.; Pezzi, N.

    1991-01-01

    Before the start up of the RA-0 'zero power' nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the '70, a work program for the future operational training personnel was elaborated. Based on the Authority's applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author) [es

  6. Safety-evaluation report related to the renewal of the operating license for the Cornell University TRIGA Research Reactor. Docket No. 50-157

    International Nuclear Information System (INIS)

    1983-08-01

    This Safety Evaluation Report for the application filed by the Cornell University for a renewal of Operating License R-80 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York. The staff concludes that the TRIGA reactor facility can continue to be operated by Cornell without endangering the health and safety of the public

  7. Technology transfer of Cornell university

    International Nuclear Information System (INIS)

    Yoo, Wan Sik

    2010-01-01

    This book introduces technology transfer of Cornell university which deals with introduction of Cornell university, composition of organization and practice of technology transfer : a research contract, research perform, invention report, evaluation and succession of invention, a patent application and management, marketing, negotiation and writing contract, management of contract, compensation, result of technology transfer, cases of success on technical commercialization and daily life of technology transfer center.

  8. The Swedish Zero Power Reactor R0

    Energy Technology Data Exchange (ETDEWEB)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-15

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of {+-} 0. 1 mm.

  9. The Swedish Zero Power Reactor R0

    International Nuclear Information System (INIS)

    Landergaard, Olof; Cavallin, Kaj; Jonsson, Georg

    1961-05-01

    The reactor R0 is a critical facility built for heavy water and natural uranium or fuel of low enrichment,, The first criticality was achieved September 25, 1959. During a first period of more than two years the R0 will be operated as a bare reactor in order to simplify interpretation of results. The reactor tank is 3. 2 m high and 2. 25 m in diameter. The fuel suspension system is quite flexible in order to facilitate fuel exchange and lattice variations. The temperature of the water can be varied between about 10 and 90 C by means of a heater and a cooler placed in the external circulating system. The instrumentation of the reactor has to meet the safety requirements not only during operation but also during rearrangements of the core in the shut-down state. Therefore, the shut-down state is always defined by a certain low 'safe' moderator level in the reactor tank. A number of safety rods are normally kept above the moderator ready for action. For manual or automatic control of the reactor power a specially designed piston pump is needed, by which the moderator level is varied. The pump speed is controlled from the reactor power error by means of a Ward-Leonard system. Moderator level measurement is made by means of a water gauge with an accuracy of ± 0. 1 mm

  10. Status report about the works for the start up of the RA-0 'zero power' nuclear reactor at the Cordoba National University

    International Nuclear Information System (INIS)

    Martin, H.R; Carballido, C.; Oliveras, T.

    1991-01-01

    After two years of works at the Cordoba National University for the new start-up of the RA-0 'zero power' nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author) [es

  11. Design characteristics of zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Popovic, D.; Nikolic, D.; Antic, D.; Zavaljevski, N.

    1987-01-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  12. Design characteristics of research zero power fast reactor Lasta

    International Nuclear Information System (INIS)

    Milosevic, M.; Stefanovic, D.; Pesic, M.; Nikolic, D.; Antic, D.; Zavaljevski, N.; Popovic, D.

    1990-01-01

    LASTA is a flexible zero power reactor with uranium and plutonium fuel designed for research in the neutron physics and in the fast reactor physics. Safety considerations and experimental flexibility led to the choice of a fixed vertical assembly with two safety blocks as the main safety elements, so that safety devices would be operated by gravity. The neutron and reactor physics, the control and safety philosophy adopted in our design, are described in this paper. Developed computer programs are presented. (author)

  13. Preliminary concept of a zero power nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  14. Preliminary concept of a zero power nuclear reactor core

    International Nuclear Information System (INIS)

    Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

    2011-01-01

    The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

  15. The zero-power basis of fast reactor dosimetry

    International Nuclear Information System (INIS)

    Sanders, J.E.

    1978-06-01

    Predictions of reaction rates, atomic displacements, and gamma-ray energy deposition in the Prototype Fast Reactor are based on cross-section data and calculation methods validated against the results of zero-power experiments. The paper reviews work in Zebra relevant to this dosimetry, including neutron spectrometry, power mapping, foil activations within core heterogeneities, and measurements with thermoluminescent detectors. Comparisons of experiment and calculation are discussed in relation to the accuracies required to meet materials testing objectives. (author)

  16. The zero-power basis of fast reactor dosimetry

    International Nuclear Information System (INIS)

    Sanders, J.E.

    1978-06-01

    Predictions of reaction rates, atomic displacements, and gamma-ray energy deposition in the Prototype Fast Reactor are based on cross-section data and calculation methods validated against the results of zero-power experiments. The paper reviews work in Zebra relevant to this dosimetry, including neutron spectrometry, power mapping, foil activations within core heterogeneities, and measurements with thermoluminescent detectors. Comparisons of experiment and calculation are discussed in relation to the accuracies required to meet material testing objectives. (author)

  17. Experimental measurement of zero power reactor transfer function

    International Nuclear Information System (INIS)

    Liang Shuhong

    2011-01-01

    In order to study the zero power reactor (ZPR) transfer function, the ZPR transfer function expression was deduced with the point reactor kinetics equation, which was disturbed by reactivity input response. Based on the Fourier analysis for the input of triangular wave, the relation between the transfer function and reactivity was got. Validating research experiment was made on the DF-VI fast ZPR. After the disturbed reactivity was measured, the experimental value of the transfer function was got. According to the experimental value and the calculated value, the expression of the ZPR transfer function is proved, whereas the disturbed reactivity is got from the transfer function. (authors)

  18. Recent activities at the zero-power teaching reactor CROCUS

    International Nuclear Information System (INIS)

    Girardin, G.; Chawla, R.

    2011-01-01

    CROCUS is a zero-power critical facility used mainly for educational purposes at the Swiss Federal Institute of Technology (EPFL) in Lausanne, Switzerland. It is a low-enriched-uranium fuelled, light-water moderated reactor, with the fission power limited to 100 W. The presentation will discuss the crucial role of CROCUS in teaching -- both as framework for reactor practicals offered to physics students at EPFL and as key educational tool in the recently established Swiss Master of Science in Nuclear Engineering. Regular development work is needed for the various instruments and components associated with the facility. As illustration, the recently completed refurbishment of the control rod system and the related calibration experiments will also be discussed.

  19. The zero power reactor SUR and its application

    International Nuclear Information System (INIS)

    Wesser, U.

    1986-01-01

    This low-power reactor, rated nominally at 100 milliwatts, has a cylindrical core of 26 cm in diameter and 24 cm high consisting of U 3 O 8 powder in a polyethylene matrix. The fuel is 20 percent enriched and the critical mass about 700 g. The excess reactivity is about 3 mk. The reactivity is controlled by two cadmium sheets in addition to a back-up system that drops the inner reflector. The reactor has no active cooling system. Personnel costs include a supervisor and an operator. The reactor is used for training in Reactor Theory (including use of a neutron chopper), reactor kinetics, nuclear technology, reactor operations and for doctoral thesis research. (author)

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

    International Nuclear Information System (INIS)

    Raisic, N.

    1962-09-01

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

  1. Design characteristics of zero power fast reactor Lasta; Osnovne karakteristike brzog reaktora nulte snage Lasta

    Energy Technology Data Exchange (ETDEWEB)

    Milosevic, M; Stefanovic, D; Pesic, M; Popovic, D; Nikolic, D; Antic, D; Zavaljevski, N [Institut za nuklearne nauke Boris Kidric, Vinca, Beograd (Yugoslavia)

    1987-07-01

    The concept, purpose and preliminary design of a zero power fast reactor LASTA are described. The methods of computing the reactor core parameters and reactor kinetics are presented with the basic calculated results and analysis for one selected LASTA configuration. The nominal parameters are determined according to the selected reactor safety criteria and results of calculations. Important aspects related to the overall safety are examined in detail. (author)

  2. Control and instrumentation system of the Zero Power Reactor at IEA, Sao Paulo (Brazil)

    International Nuclear Information System (INIS)

    Peluso, M.A.V.; Matsuda, K.; Hukai, R.

    1974-01-01

    The control and instrumentation system of the Zero Power Reactor at the IEA (Institute of Atomic Energy - Sao Paulo, Brazil) is described. Technical specifications of the main items of equipment are presented in a general way. Information is also given on the connection between the system described and the electrical supply system of the IEA reactor physics laboratory [pt

  3. Determination of the theoretical and experimental zero-power frequency response of Ghana Research Reactor-1

    International Nuclear Information System (INIS)

    Intsiful, J.D.K.; Akaho, E.H.K.; Tetteh, G.K.

    1997-12-01

    The frequency response measurements of a reactor at low power help in determining the kinetic parameters of a reactor and ultimately in investigating its stability with respect to small perturbations in reactivity. In this report, we present the results of the zero-power frequency response measurements of GHARR-1 by rod method and its analytical analogue. The comparison in calculated and measured values is reasonably good in the frequency range used (author)

  4. Cross-disciplinary research programs at the Cornell TRIGA reactor

    International Nuclear Information System (INIS)

    Clark, D.D.

    1995-01-01

    This paper describes cross-disciplinary research efforts at the Cornell TRIGA reactor. A new graduate laboratory course for nonspecialists was developed which brought in graduate students from many fields, and a weekly or bimonthly nuclear methods seminars are being held to describe research methods, sample preparation, irradiation, etc

  5. Zirconium-hydride solid zero power reactor and its application research

    International Nuclear Information System (INIS)

    Lin Shenghuo; Luo Zhanglin; Su Zhuting

    1994-10-01

    The Zirconium Hydride Solid Zero Power Reactor built at China Institute of Atomic Energy is introduced. In the reactor Zirconium-hydride is used as moderator, plexiglass as reflector and U 3 O 8 with enrichment of 20% as the fuel, Since its initial criticality, the physical characteristics and safety features have been measured with the result showing that the reactor has sound stability and high sensitivity, etc. It has been successfully used for the personnel training and for the testing of reactor control instruments and experiment devices. It also presents the special advantage for the pre-research of some applications

  6. The control-and-instrumentation system of the IEA zero power reactor and its reliability calculation

    International Nuclear Information System (INIS)

    Peluso, M.A.V.

    1978-01-01

    The control-and instrumentation system for the Instituto de Energia Atomica Zero Power Reactor is described and the design criteria are presented and discussed. The reliability analysis for the reactor protection system was performed using the fault tree method. This was done using a computer code based on the Monte Carlo simulation. That code is an adaptation of the SAFTE-I, for the IBM 360/155 IEA Computer. (Author) [pt

  7. Design of the zero power reactor core of Instituto de Energia Atomica, SP, Brazil

    International Nuclear Information System (INIS)

    Ferreira, Antonio Carlos de Almeida

    1974-01-01

    The main characteristics of a graphite moderated core of a critical assembly to be installed in the zero power reactor of the Instituto de Energia Atomica have been defined. Several simple geometric configurations have been selected and criticality studies have been made. The necessary quantity of fissile uranium has been calculated. (author)

  8. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  9. Measures of the zero power nuclear reactor's kinetic parameters with application of noise analysis

    International Nuclear Information System (INIS)

    Martins, F.R.

    1992-01-01

    The purpose of this work was to establish an experimental technique based on noise analysis for measuring the ratio of kinetic parameters β/ Λ and the power of the Zero Power Nuclear Reactor IPEN-MB 01. A through study of the microscopic and macroscopic noise analysis techniques has been carried out. The Langevin technique and the point kinetic model were chosen to describe the stochastic phenomena that occur in the zero power reactor. Measurements have been made using two compensated ionization chambers localized in the water reflector at symmetric positions in order to minimize spatial effects on the neutron flux fluctuation. Power calibrations based on the low frequency plateau of the cross-power spectral density has also been carried out. (author)

  10. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    International Nuclear Information System (INIS)

    Novelli, A.

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

  11. Determination of spatially dependent transfer function of zero power reactor by using pseudo-random incentive

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1973-01-01

    Specially constructed fast reactivity oscillator was stimulating the zero power reactor by a stimulus which caused pseudo-random reactivity changes. Measuring system included stochastic oscillator BCR-1 supplied by pseudo-random pulses from noise generator GBS-16, instrumental tape-recorder, system for data acquisition and digital computer ZUSE-Z-23. For measuring the spatially dependent transfer function, reactor response was measured at a number of different positions of stochastic oscillator and ionization chamber. In order to keep the reactor system linear, experiment was limited to small reactivity fluctuations. Experimental results were compared to theoretical ones

  12. Superconducting rf activities at Cornell University

    International Nuclear Information System (INIS)

    Padamsee, H.; Hakimi, M.; Kirchgessner, J.

    1988-01-01

    Development of rf superconductivity for high energy accelerators has been a robust activity at the Cornell Laboratory of Nuclear Studies (LNS) for many years. In order to realize the potential of rf superconductivity, a two-pronged approach has been followed. On the one hand accelerator applications were selected where the existing state-of-the art of superconducting rf is competitive with alternate technologies, then LNS engaged in a program to design, construct and test suitable superconducting cavities, culminating in a full system test in an operating accelerator. On the second front the discovery and invention of ideas, techniques and materials required to make superconducting rf devices approach the ideal in performance has been aggressively pursued. Starting with the development of superconducting cavities for high energy electron synchrotrons, the technology was extended to high energy e + e - storage rings. The LE5 cavity design has now been adopted for use in the Continuous Electron Beam Accelerator Facility (CEBAF). When completed, this project will be one of the largest applications of SRF technology, using 440 LE5 modules[4]. In the last two years, the cavity design and the technology have been transferred to industry and CEBAF. Cornell has tested the early industrial prototypes and cavity pairs. LNS has developed, in collaboration with CEBAF, designs and procedures for cavity pair and cryomodule assembly and testing. Advanced research for future electron accelerators is badly needed if particle physicists hope to expand the energy frontier. Superconducting cavity technology continues to offer attractive opportunities for further advances in achievable voltage at reasonable cost for future accelerators. For Nb, the full potential implies an order of magnitude increase over current capabilities. 20 references, 11 figures

  13. Superconducting RF activities at Cornell University

    International Nuclear Information System (INIS)

    Kirchgessner, J.; Moffat, D.; Padamsee, H.; Rubin, D.; Sears, J.; Shu, Q.S.

    1990-01-01

    This paper outlines the RF superconductivity research and development work that has taken place at Cornell Laboratory of Nuclear Studies over the past years. The work that has been performed since the last RF superconductivity workshop is emphasized together with a discussion of the direction of future efforts. Past work is summarized first, focusing on research and development activities in the area of RF superconductivity. Superconducting TeV linear collider is then discussed focusing on the application of superconducting RF to a future TeV linear collider. Linear collider structure development is then described centering on the development of a simpler (thereby cheaper) structure for a TeV linear collider. B-factory with superconducting RF is outlined focusing on the formulation of a conceptual design for a B-factory. B-factory structure development is discussed in relation to the advancement in the capability of SC cavities to carry beam currents of several amperes necessary for a high luminosity storage ring. High gradients are discussed as the key to the realization of a high energy superconducting linac or a superconducting RF B-factory. (N.K.)

  14. Core concept of fast power reactor with zero sodium void reactivity

    International Nuclear Information System (INIS)

    Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

    1991-01-01

    The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

  15. Accident at the zero power reactor which happened on October 15 1958

    International Nuclear Information System (INIS)

    Savic, P.

    1959-01-01

    During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine

  16. Preparation of mandatory documentation before the start up of the RA-0 `zero power` nuclear reactor at Cordoba National University; Preparacion de la documentacion mandatoria para la puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Keil, W M; Pezzi, N

    1992-12-31

    Before the start up of the RA-0 `zero power` nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the `70, a work program for the future operational training personnel was elaborated. Based on the Authority`s applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author). [Espanol] Con motivo de la nueva puesta en servicio del REACTOR NUCLEAR RA-0 fue necesario elaborar la documentacion mandatoria requerida por la Autoridad Regulatoria Nacional. Siguiendo los lineamientos de las normas y recomendaciones vigentes e incluyendo criterios propios en lo que debia ser el contenido final de dicha documentacion, fue preparado lo que se ha denominado el INFORME DE SEGURIDAD DEL REACTOR NUCLEAR RA-0. Este documento que se describe en este trabajo, si bien contiene las habituales descripciones de todos los Informes de Seguridad, incluye otros aspectos que no siendo requeridos expresamente en el mismo, han dado una mayor coherencia a la conformacion de todos los aspectos que interrelacionan las areas de seguridad fisica, radiologica, nuclear y de control de materiales nucleares bajo salvaguardias. (Autor).

  17. Zero energy reactor 'RB'

    Energy Technology Data Exchange (ETDEWEB)

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

    1959-03-15

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

  18. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N J.H.

    1966-12-15

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.

  19. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    International Nuclear Information System (INIS)

    Schuh, N.J.H.

    1966-12-01

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed

  20. Status report about the works for the start up of the RA-0 `zero power` nuclear reactor at the Cordoba National University; Estado actual de avance de las tareas para la nueva puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

    Energy Technology Data Exchange (ETDEWEB)

    Martin, H R; Carballido, C; Oliveras, T

    1992-12-31

    After two years of works at the Cordoba National University for the new start-up of the RA-0 `zero power` nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author). [Espanol] Luego de aproximadamente dos anos de trabajo para la nueva puesta en marcha del REACTOR NUCLEAR RA-0, se han alcanzado los resultados presentados en este trabajo. Partiendo de una infraestructura practicamente inexistente en cuanto a recursos humanos y estado de las instalaciones, los avances logrados son significativos. Comenzando por la capacitacion y el entrenamiento del futuro personal de operacion y pasando por la adecuacion de los equipos y componentes, hasta la confeccion de la documentacion mandatoria, se muestran los aspectos mas destacables de los trabajos realizados. Una atencion especial se dedica a la insercion de una instalacion de este tipo en el ambito universitario, el cual por sus particulares caracteristicas, ha debido ser tenido en cuenta permanentemente para la futura operacion de las instalaciones. (Autor).

  1. Fabrication of zero power reactor fuel elements containing 233U3O8 powder

    International Nuclear Information System (INIS)

    Nicol, R.G.; Parrott, J.R.; Krichinsky, A.M.; Box, W.D.; Martin, C.W.; Whitson, W.R.

    1982-05-01

    Oak Ridge National Laboratory, under contract with Argonne National Laboratory, completed the fabrication of 1743 fuel elements for use in their Zero Power Reactor. The contract also included recovery of 20 kg of 233 U from rejected elements. This report describes the steps associated with conversion of purified uranyl nitrate (as solution) to U 3 O 8 powder (suitable for fuel) and subsequent charging, sealing, decontamination, and testing of the fuel elements (packets) preparatory to shipment. The nuclear safety, radiation exposures, and quality assurance aspects of the program are discussed

  2. Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

    International Nuclear Information System (INIS)

    Almeida Ferreira, A.C. de; Hukai, R.Y.

    1975-01-01

    The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity

  3. Verification Survey of the Building 315 Zero Power Reactor-6 Facility, Argonne National Laboratory-East, Argonne, Illinois

    International Nuclear Information System (INIS)

    W. C. Adams

    2007-01-01

    Oak Ridge Institute for Science and Education (ORISE) conducted independent verification radiological survey activities at Argonne National Laboratory's Building 315, Zero Power Reactor-6 facility in Argonne, Illinois. Independent verification survey activities included document and data reviews, alpha plus beta and gamma surface scans, alpha and beta surface activity measurements, and instrumentation comparisons. An interim letter report and a draft report, documenting the verification survey findings, were submitted to the DOE on November 8, 2006 and February 22, 2007, respectively (ORISE 2006b and 2007). Argonne National Laboratory-East (ANL-E) is owned by the U.S. Department of Energy (DOE) and is operated under a contract with the University of Chicago. Fundamental and applied research in the physical, biomedical, and environmental sciences are conducted at ANL-E and the laboratory serves as a major center of energy research and development. Building 315, which was completed in 1962, contained two cells, Cells 5 and 4, for holding Zero Power Reactor (ZPR)-6 and ZPR-9, respectively. These reactors were built to increase the knowledge and understanding of fast reactor technology. ZPR-6 was also referred to as the Fast Critical Facility and focused on fast reactor studies for civilian power production. ZPR-9 was used for nuclear rocket and fast reactor studies. In 1967, the reactors were converted for plutonium use. The reactors operated from the mid-1960's until 1982 when they were both shut down. Low levels of radioactivity were expected to be present due to the operating power levels of the ZPR's being restricted to well below 1,000 watts. To evaluate the presence of radiological contamination, DOE characterized the ZPRs in 2001. Currently, the Melt Attack and Coolability Experiments (MACE) and Melt Coolability and Concrete Interaction (MCCI) Experiments are being conducted in Cell 4 where the ZPR-9 is located (ANL 2002 and 2006). ANL has performed final

  4. Measurements of the Reactivity Properties of the Aagesta Nuclear Power Reactor at Zero Power

    Energy Technology Data Exchange (ETDEWEB)

    Bernander, G

    1967-07-15

    The moderator level and temperature coefficients of reactivity and control rod differential reactivity worths have been determined at zero power by means of period measurements. The moderator level coefficient and the corresponding critical level have been measured for the 32, 68 and 136 fuel assembly cores at room temperature for cores with and without control rods. From these results the worths of control rods have been derived. HETERO calculations give up to 15 % lower values than the experimental results. The cold fresh core has an excess reactivity of 9.0 {+-} 0.2 %. The temperature coefficient and differential control rod worths were measured for the fully loaded core with filled tank in the temperature range between 30 and 210 deg C. Critical positions as a function of temperature were obtained for the corresponding control rod groups. No relevant calculations of the temperature coefficient for comparison with the experimental values have yet been made, but the experimental results together with measured critical control rod positions give good opportunities to check calculational programs. HETERO has been shown in these cases to reproduce differential control rod worths and critical positions fairly well. However, a certain underestimation of the rod effectiveness is quite noticeable. The relative increase in control rod effectiveness with a temperature change from 20 to 220 deg C has been estimated to be 0.29 {+-} 0.06.

  5. Accident at the zero power reactor which happened on October 15 1958; Sur l'accident avec le reacteur de puissance zero du 15 octobre 1958

    Energy Technology Data Exchange (ETDEWEB)

    Savic, P [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

    1959-03-15

    During an experiment on the zero power heavy water reactor with natural uranium fuel in the Boris Kidric Institute of Nuclear Sciences, the reactor escaped control. Six staff members in the immediate surrounding of the bare assembly were exposed to high neutron and ionising irradiation. Other two employees who were at some bigger distance were exposed to doses higher than permitted. This paper deals with the circumstances that caused the accident, status of the dosimetry, control and alarm systems. Individual exposure doses were estimated according to the calculated neutron flux values obtained from measuring the activities of personal belongings made of gold and copper as well as radioactive phosphorous from urine.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-02-15

    Numerous research, central station power, propulsion, isotope production, and test reactor designs have been investigated in Argonne's zero-power reactor facilities, and related exponential and clean critical assemblies have provided basic data. To present a representative account of recent experiments and to demonstrate the wide variety of reactor design information obtainable in low flux systems, the following experimental programmes are reviewed: 1. A study of the properties of thoria-urania fuel in heavy water, with particular attention to the requirements for design of a second core for Argonne's Experimental Boiling Water Reactor; 2. A mock-up of a proposed high flux research reactor to confirm the design calculations, optimize the geometry and estimate the effect of fuel burn-up; 3. A determination of the power distribution patterns and reactivity effect of fuel element flooding for a combined boiling-superheat reactor test; 4. The design of a sodium cooled. U{sup 235} fueled, plutonium producing fast breeder reactor core as a first loading for Argonne's Experimental Breeder Reactor II; and 5. An investigation of the characteristics of a reactor with interacting thermal and fast neutron zones. In the discussion of these programmes, the circumstances which influenced the choice among exponentials, clean criticals, zero-power mock-ups and in situ experiments for the acquisition of the required data are explained, as is the role played by supporting analytical effort. The extent to which reactor design data can be attained before actual operation at power is illustrated by specific examples. Such data include shutdown margin, excess reactivity for operational requirements, temperature coefficients, control and safety rods' effectiveness, reactor kinetics, power production patterns, requirements for start-up source and instrument sensitivity, shielding needs and neutron economy. This review of recent activities in zero-power experimentation reveals the strong

  7. Current Mode Neutron Noise Measurements in the Zero Power Reactor CROCUS

    Science.gov (United States)

    Pakari, O.; Lamirand, V.; Perret, G.; Braun, L.; Frajtag, P.; Pautz, A.

    2018-01-01

    The present article is an overview of developments and results regarding neutron noise measurements in current mode at the CROCUS zero power facility. Neutron noise measurements offer a non-invasive method to determine kinetic reactor parameters such as the prompt decay constant at criticality α = βeff / λ, the effective delayed neutron fraction βeff, and the mean generation time λ for code validation efforts. At higher detection rates, i.e. above 2×104 cps in the used configuration at 0.1 W, the previously employed pulse charge amplification electronics with BF3 detectors yielded erroneous results due to dead time effects. Future experimental needs call for higher sensitivity in detectors, higher detection rates or higher reactor powers, and thus a generally more versatile measurement system. We, therefore, explored detectors operated with current mode acquisition electronics to accommodate the need. We approached the matter in two ways: 1) By using the two compensated 10B-coated ionization chambers available in CROCUS as operational monitors. The compensated current signal of these chambers was extracted from coremonitoring output channels. 2) By developing a new current mode amplification station to be used with other available detectors in core. Characteristics and first noise measurements of the new current system are presented. We implemented post-processing of the current signals from 1)and 2) with the APSD/CPSD method to determine α. At two critical states (0.5 and 1.5 W), using the 10B ionization chambers and their CPSD estimate, the prompt decay constant was measured after 1.5 hours to be α=(156.9 ± 4.3) s-1 (1σ). This result is within 1σ of statistical uncertainties of previous experiments and MCNPv5-1.6 predictions using the ENDF/B-7.1 library. The newsystem connected to a CFUL01 fission chamber using the APSDestimate at 100 mW after 33 min yielded α = (160.8 ± 6.3) s-1, also within 1σ agreement. The improvements to previous neutron noise

  8. Microelectronics materials characterization studies at the Cornell TRIGA Reactor

    International Nuclear Information System (INIS)

    McGuire, Stephen C.

    1992-01-01

    The Cornell program of microelectronics materials characterization by neutron activation analysis (NAA) is described. Experimental details and results from the successful application of NAA to silicon germanium circuit structures and nickel silicide layers are presented. In doing so, the potential for using X rays from isotopes that decay by electron capture is demonstrated. (author)

  9. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

  10. Feline dermatology at Cornell University: 1407 cases (1988-2003).

    Science.gov (United States)

    Scott, Danny W; Miller, William H; Erb, Hollis N

    2013-04-01

    Medical records of 1407 cats with dermatologic diagnoses made at Cornell University teaching hospital from 1988 to 2003 were tabulated. We expressed the diagnoses as counts, percentages of the cats with dermatologic disease (1407) and percentages of all cats seen at the university hospital (22,135) during the same period. A total of 1887 diagnoses were made in the 1407 cats. We compared the age, sex and breed group of our cases with all those 22,135 cats in ('1-by-c') χ(2) tests in which the hospital population was considered a standard (rather than a 'sample'). The 10 most common dermatoses, their counts, and the proportions of dermatologic diagnoses and of the total cat population that the cats with these dermatoses represented were: allergy (298; 15.8%; 1.35%), atopic dermatitis (194; 10.3%; 0.88%), bacterial folliculitis/furunculosis (189; 10.0%; 0.85%), otodectic mange (115; 6.1%; 0.52%), flea infestation (99; 5.2%; 0.45%), feline acne (74; 3.9%; 0.33%), flea-bite allergy (70; 3.7%; 0.32%), cutaneous adverse drug reaction (56; 3.0%; 0.25%), idiopathic eosinophilic-granuloma complex (55; 2.9%; 0.25%) and abscess (51; 2.7%; 0.23%). Allergies of all types, combined, accounted for 32.7% of all the feline dermatoses. Relative to the standard of the total hospital population, cats <2 years old and females (both intact and spayed) were significantly under-represented (all P≤0.001) in the dermatologic case series. In contrast, Himalayans (compared with domestic short- or longhair, Persian, Siamese and other breeds) and males (both intact and neutered) were significantly over-represented (all P ≤0.001).

  11. Measurement of zero power reactor dynamic response by cross correlation method; Merenje dinamickog odziva reaktora nulte snage kros korelacionom metodom

    Energy Technology Data Exchange (ETDEWEB)

    Kostic, Lj; Petrovic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1969-07-15

    Pulse response is comprehensive description of linear system dynamics. In this paper, cross correlation method was used for measuring the response of zero power reactor. Reactor system was perturbed by pseudo-random signal, which was cross correlated with the reactor signal responding to this perturbation on the digital ZUSE Z-23 computer. Cross-correlation functions were measured for different positions of stochastic oscillator and ionization chamber in the critical system. From numerical processing of performed experimental data, it was concluded that a more powerful faster computer would be needed for processing statistical experiments. In that case it would be possible to obtain information about spatial effects in the reactor and propagation of neutron waves in the multiplication medium. Impulsni odziv je potpuni opis dinamike linearnog sistema. Za merenje impulsnog odziva nultog reaktora, u ovom radu, koriscena je kros korelaciona metoda. Reaktorski sistem je perturbovan pseudoslucajnim signalom, koji je u digitalnom racunaru ZUSE Z-23 kroskorelisan sa signalom odziva reaktora na ove perturbacije. Merene su kroskorelacione funkcije za razlicite polozaje stohastickog oscilatora i jonizacione komore u kriticnom sistemu. Iz numericki obradjivanih eksperimenta namece se kao zakljucak da bi za obradu statistickih eksperimenata kod nultih reaktora bio potreban racunar veceg kapaciteta i brzine. U tom slucaju bi se iz ovako postavljenog eksperimenta moglo doci i do informacija o prostornim efektima u reaktoru i prostiranju neutronskih talasa kroz multiplikativnu sredinu. (author)

  12. State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

    International Nuclear Information System (INIS)

    Le Guillou, M.; Gruel, A.; Destouches, C.; Blaise, P.

    2017-01-01

    The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs) and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n-γ field. A recently developed methodology based on the use of "7Li and "6Li-enriched TLDs, pre-calibrated both in photon and neutron fields, is a promising approach to de-convolute the 2 components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fiber-using setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1 σ). (authors)

  13. State of the art on nuclear heating measurement methods and expected improvements in zero power research reactors

    Directory of Open Access Journals (Sweden)

    Le Guillou Mael

    2017-01-01

    Full Text Available The paper focuses on the recent methodological advances suitable for nuclear heating measurements in zero power research reactors. This bibliographical work is part of an experimental approach currently in progress at CEA Cadarache, aiming at optimizing photon heating measurements in low-power research reactors. It provides an overview of the application fields of the most widely used detectors, namely thermoluminescent dosimeters (TLDs and optically stimulated luminescent dosimeters. Starting from the methodology currently implemented at CEA, the expected improvements relate to the experimental determination of the neutron component, which is a key point conditioning the accuracy of photon heating measurements in mixed n–γ field. A recently developed methodology based on the use of 7Li and 6Li-enriched TLDs, precalibrated both in photon and neutron fields, is a promising approach to deconvolute the two components of nuclear heating. We also investigate the different methods of optical fiber dosimetry, with a view to assess the feasibility of online photon heating measurements, whose primary benefit is to overcome constraints related to the withdrawal of dosimeters from the reactor immediately after irradiation. Moreover, a fibered setup could allow measuring the instantaneous dose rate during irradiation, as well as the delayed photon dose after reactor shutdown. Some insights from potential further developments are given. Obviously, any improvement of the technique has to lead to a measurement uncertainty at least equal to that of the currently used methodology (∼5% at 1σ.

  14. Study of reactor parameters of on critical systems, Phase I: Safety report for RB zero power reactor; Ispitivanje reaktorskih parametara na kriticnim sistemima, I faza: Izvestaj o sigurnosti reaktora nulte snage RB

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1962-09-15

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

  15. Investigating heavy water zero power reactors with a new core configuration based on experiment and calculation results

    Energy Technology Data Exchange (ETDEWEB)

    Nasrazadani, Zahra; Salimi, Raana; Askari, Afrooz; Khorsandi, Jamshid; Mirvakili, Mohammad; Mashayekh, Mohammad [Reactor Research School, Nuclear Science and Technology Research Institute, Atomic Energy Organization of Iran, Esfahan (Iran, Islamic Republic of)

    2017-02-15

    The heavy water zero power reactor (HWZPR), which is a critical assembly with a maximum power of 100 W, can be used in different lattice pitches. The last change of core configuration was from a lattice pitch of 18-20 cm. Based on regulations, prior to the first operation of the reactor, a new core was simulated with MCNP (Monte Carlo N-Particle)-4C and WIMS (Winfrith Improved Multigroup Scheme)-CITATON codes. To investigate the criticality of this core, the effective multiplication factor (Keff) versus heavy water level, and the critical water level were calculated. Then, for safety considerations, the reactivity worth of D2O, the reactivity worth of safety and control rods, and temperature reactivity coefficients for the fuel and the moderator, were calculated. The results show that the relevant criteria in the safety analysis report were satisfied in the new core. Therefore, with the permission of the reactor safety committee, the first criticality operation was conducted, and important physical parameters were measured experimentally. The results were compared with the corresponding values in the original core.

  16. An FFAG-ERL at Cornell University for eRHIC prototyping and bright-beam applications

    Energy Technology Data Exchange (ETDEWEB)

    Hoffstaetter, Georg [Cornell University, Ithaca, NY (United States)

    2016-07-01

    Cornell University has prototyped technology essential for any high-brightness electron ERL. This includes a DC gun and an SRF injector Linac with world-record current and normalized brightness in a bunch train, a high-current CW cryomodule for 70 MeV energy gain, a high-power beam stop, and several diagnostics tools for high-current and high-brightness beams, e.g. slid measurements for 6-D phase-space densities, a fast wire scanner for beam profiles, and beam loos diagnostics. All these are now available to equip a one-cryomodule ERL, and laboratory space has been cleared out and is radiation shielded to install this ERL at Cornell. BNL has designed a multi-turn ERL for eRHIC, where beam is transported more than 20 times around the RHIC tunnel. The number of transport lines is minimized by using two non-scaling (NS) FFAG arcs. A collaboration between BNL and Cornell has been formed to investigate the new NS-FFAG optics and the multi-turn eRHIC ERL design by building a 4-turn, one-cryomodule ERL at Cornell. It has a NS-FFAG return loop built with permanent magnets and is meant to accelerate 40 mA beam to 250 MeV.

  17. Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

    2012-07-01

    The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

  18. Use of zero power plutonium reactor measurements as a support of criticality prediction for the SNR-300

    International Nuclear Information System (INIS)

    Pilate, S.; de Wouters, R.; Wehmann, U.; Helm, F.; Scholtyssek, W.

    1978-01-01

    Evaluations of criticality measurements performed in various SNEAK and Zero Power Plutonium Reactor (ZPPR) cores are compared. The best available methods of calculations (including transport theory) are used. The ZPPR results support well the trend indicated by the SNEAK evaluations for clean cores and for cores with followers; for cores with absorbers partially inserted, the agreement is only rough. Evaluations of control rod worth measurements are therefore also compared, using the routine method of calculation for SNR-300 (diffusion theory). The control rod worths are largely underestimated in SNEAK (C/E = 0.89), but only slightly underestimated in the ZPPR (C/E = 0.97). The difference in the nature of core fuel (uranium in SNEAK, plutonium in the ZPPR) could be at the origin of this discrepancy

  19. MSR - SPHINX concept program Eros (Experimental zero power Salt reactor SR-0) - The proposed experimental program as a basis for validation of reactor physics methods

    Energy Technology Data Exchange (ETDEWEB)

    Hron, M.; Juricek, V.; Kyncl, J.; Mikisek, M.; Rypar, V. [Nuclear Research Institute Rez plc, Rez (Czech Republic)

    2007-07-01

    The Molten Salt Reactor (MSR) - SPHINX (SPent Hot fuel Incinerator by Neutron fluX) concept solves this principal problem of spent fuel treatment by means of so-called nuclear incineration. It means the burning of fissionable part of its inventory and transmutation of other problematic radionuclides by use of nuclear reactions with neutrons in a MSR-SPHINX system. This reactor system is an actinide burner (most in resonance neutron spectrum) and a radionuclide transmuter in a well-thermalized neutron spectrum. In the frame of the physical part, there are computational analyses and experimental activities. The experimental program has been focused, in its first stage, on a short-term irradiation of small size samples of molten-salt systems as well as structural materials proposed for the MSR blanket in the field of high neutron flux of research reactors. The proposed next stage of the program will focus on a large-scale experimental verification of design inputs by use of MSR-type inserting zones into the existing light water moderated experimental reactor LR-0, which may allow us to modify it into the experimental zero power salt reactor SR-0. There will be a detail description of the proposed program given in the paper together with the so far performed experiments and their first results. These realized experiments help us also to verify computational codes used, and to recognize some anomalies related to molten fluorides utilization. (authors)

  20. A visit to Cornell University, Ithaca, USA : Notes on the International Workplace Studies Program IWSP

    NARCIS (Netherlands)

    van der Voordt, Theo

    2004-01-01

    In October 2004 I had the opportunity to visit Cornell University in Ithaca, New York. The purpose of my visit was to learn more about the International Workplace Studies Program (IWSP) that was launched in 1989 by Franklin Becker and William (Bill) Sims. Frank is the present chair (Bill the former)

  1. Educating the Employee Assistance Professional: Cornell University's Employee Assistance Education and Research Program.

    Science.gov (United States)

    Quick, R. C.; And Others

    1987-01-01

    Outlines Cornell University's Employee Assistance Education and Research Program, which uses an academic curriculum and field experience to further develop the Employee Assistance Program (EAP) profession. Addresses the dilemma of personnel executives in ensuring quality in EAP programs and staff. (JOW)

  2. Theoretical Work for the Fast Zero-Power Reactor FR-0

    Energy Technology Data Exchange (ETDEWEB)

    Haeggblom, H

    1965-08-15

    The theoretical part of the fast reactor physics work in Sweden, has mainly been connected with the FR-0 reactor. The report describes the principal features of this reactor, evaluation of cross sections, calculations of critical masses, reactivity of the air gap and of control rods and calculations of neutron generation time and effective beta values. Carlson codes in spherical and in cylindrical geometry are used to evaluate critical masses and fluxes. In cases when reactivity changes are calculated, complementary methods are perturbation theory and variational calculus. The agreement with experiments is in some cases good, especially the determination of critical mass, but in other cases discrepancies are observed, e.g. the activation of U-238 in the reflector is much larger than the theoretical spectrum predicts.

  3. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Bess, John D.; Fujimoto, Nozomu

    2014-01-01

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments

  4. The Texts of the Instruments relating to a Project for a Joint Agency-Norwegian Program of Research with the Zero Power Reactor 'NORA'

    International Nuclear Information System (INIS)

    1961-01-01

    The text of the Supply Agreement between the Agency and the Governments of Norway and of the United States of America, and the text of the related Project Agreement between the Agency and the Government of Norway concerning an Agency project for cooperation in carrying out a joint program of research in reactor physics with the zero power reactor 'NORA', are reproduced in this document for the information of all Members of the Agency

  5. The Texts of the Instruments relating to a Project for a Joint Agency-Norwegian Program of Research with the Zero Power Reactor 'NORA'

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-06-22

    The text of the Supply Agreement between the Agency and the Governments of Norway and of the United States of America, and the text of the related Project Agreement between the Agency and the Government of Norway concerning an Agency project for cooperation in carrying out a joint program of research in reactor physics with the zero power reactor 'NORA', are reproduced in this document for the information of all Members of the Agency.

  6. Cold neutron PGAA facility developments at university research reactors in the USA

    International Nuclear Information System (INIS)

    Uenlue, K.; Rios-Martinez, C.

    2005-01-01

    The PGAA applications can be enhanced by using subthermal neutrons, cold neutrons at university research reactors. Only two cold neutron beam facilities were developed at the U.S. university research reactors, namely at Cornell University and the University of Texas at Austin. Both facilities used mesitylene moderator. The mesitylene moderator in the Cornell Cold Neutron Beam Facility (CNBF) was cooled by a helium cryorefrigerator via copper cold fingers to maintain the moderator below 30 K at full power reactor operation. Texas Cold Neutron Source (TCNS) also uses mesitylene moderator that is cooled by a cryorefrigerator via a neon thermosiphon. The operation of the TCNS is based on a helium cryorefrigerator, which liquefies neon gas in a 3-m long thermosiphon. The thermosiphon cools and maintains mesitylene moderator at about 30 K in a chamber. Neutrons streaming through the mesitylene chamber are moderated and thus reduce their energy to produce a cold neutron distribution. (author)

  7. Analysis of the Temporal Response of Coupled Asymmetrical Zero-Power Subcritical Bare Metal Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Klain, Kimberly L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-21

    The behavior of symmetrical coupled-core systems has been extensively studied, yet there is a dearth of research on asymmetrical systems due to the increased complexity of the analysis of such systems. In this research, the multipoint kinetics method is applied to asymmetrical zeropower, subcritical, bare metal reactor systems. Existing research on asymmetrical reactor systems assumes symmetry in the neutronic coupling; however, it will be shown that this cannot always be assumed. Deep subcriticality adds another layer of complexity and requires modification of the multipoint kinetics equations to account for the effect of the external neutron source. A modified set of multipoint kinetics equations is derived with this in mind. Subsequently, the Rossi-alpha equations are derived for a two-region asymmetrical reactor system. The predictive capabilities of the radiation transport code MCNP6 for neutron noise experiments are shown in a comparison to the results of a series of Rossi-alpha measurements performed by J. Mihalczo utilizing a coupled set of symmetrical bare highly-enriched uranium (HEU) cylinders. The ptrac option within MCNP6 can generate time-tagged counts in a cell (list-mode data). The list-mode data can then be processed similarly to measured data to obtain values for system parameters such as the dual prompt neutron decay constants observable in a coupled system. The results from the ptrac simulations agree well with the historical measured values. A series of case studies are conducted to study the effects of geometrical asymmetry in the coupling between two bare metal HEU cylinders. While the coupling behavior of symmetrical systems has been reported on extensively, that of asymmetrical systems remains sparse. In particular, it appears that there has been no previous research in obtaining the coupling time constants for asymmetrically-coupled systems. The difficulty in observing such systems is due in part to the inability to determine the

  8. Development of An Embedded FPGA-Based Data Acquisition System Dedicated to Zero Power Reactor Noise Experiments

    Directory of Open Access Journals (Sweden)

    Arkani Mohammad

    2014-08-01

    Full Text Available An embedded time interval data acquisition system (DAS is developed for zero power reactor (ZPR noise experiments. The system is capable of measuring the correlation or probability distribution of a random process. The design is totally implemented on a single Field Programmable Gate Array (FPGA. The architecture is tested on different FPGA platforms with different speed grades and hardware resources. Generic experimental values for time resolution and inter-event dead time of the system are 2.22 ns and 6.67 ns respectively. The DAS can record around 48-bit x 790 kS/s utilizing its built-in fast memory. The system can measure very long time intervals due to its 48-bit timing structure design. As the architecture can work on a typical FPGA, this is a low cost experimental tool and needs little time to be established. In addition, revisions are easily possible through its reprogramming capability. The performance of the system is checked and verified experimentally.

  9. Bridging Water Resources Policy and Environmental Engineering in the Classroom at Cornell University

    Science.gov (United States)

    Walter, M. T.; Shaw, S. B.; Seifert, S.; Schwarz, T.

    2006-12-01

    Current university undergraduate students in environmental sciences and engineering are the next generation of environmental protection practitioners. Recognizing this, Cornell's Biological and Environmental Engineering department has developed a popular class, Watershed Engineering (BEE 473), specifically designed to bridge the too-common gap between water resources policy and state-of-art science and technology. Weekly homework assignments are to design real-life solutions to actual water resources problems, often with the objective of applying storm water policies to local situations. Where appropriate, usually in conjunction with recent amendments to the Federal Clean Water Act, this course introduces water resource protection tools and concepts developed in the Cornell Soil and Water Lab. Here we present several examples of how we build bridges between university classrooms and the complex world of water resources policy.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  11. The Design of Universal Soft Control for Advanced Power Reactor

    International Nuclear Information System (INIS)

    Ye, Song Hae; Shin, Yeong Cheol; Kang, Sung Kon; Lyou, Joon

    2011-01-01

    Highly-Integrated Control Room (HICR) has been adopted for APR+. APR+ is being developed as the evolutionary improvement of APR1400. In HICR, the traditional control panels, with their assorted gauges, indicating lights, control switches, annunciators, etc., are replaced by computer-driven consolidated operator workstations. Soft control plays a key role in consolidating control in the workstation. Universal Soft Control (USC) has the ability to control plant functions and components in multiple divisions of safety systems and non-safety systems from a single operator workstation. The USC further contribute to the level of consolidation in the workstation and provide better human machine interface with reduced operational burden. However, the implementation of USC poses challenges to the independence among safety divisions and between safety system and non-safety system. This paper presents design features and approach to prevent adverse affects of the safety systems and to minimize the potential for spurious actuations by USC

  12. The Design of Universal Soft Control for Advanced Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Song Hae; Shin, Yeong Cheol; Kang, Sung Kon [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of); Lyou, Joon [Chungnam National University, Daejeon (Korea, Republic of)

    2011-08-15

    Highly-Integrated Control Room (HICR) has been adopted for APR+. APR+ is being developed as the evolutionary improvement of APR1400. In HICR, the traditional control panels, with their assorted gauges, indicating lights, control switches, annunciators, etc., are replaced by computer-driven consolidated operator workstations. Soft control plays a key role in consolidating control in the workstation. Universal Soft Control (USC) has the ability to control plant functions and components in multiple divisions of safety systems and non-safety systems from a single operator workstation. The USC further contribute to the level of consolidation in the workstation and provide better human machine interface with reduced operational burden. However, the implementation of USC poses challenges to the independence among safety divisions and between safety system and non-safety system. This paper presents design features and approach to prevent adverse affects of the safety systems and to minimize the potential for spurious actuations by USC.

  13. Requalification of SPERT [Special Power Excursion Reactor Test] pins for use in university reactors

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Domagala, R.F.; Dates, L.R.

    1986-12-01

    A series of nondestructive and destructive examinations have been performed on a representative sample of stainless steel-clad UO 2 fuel pins procured in the early-to-mid 1960s for the SPERT program. These examinations were undertaken in order to requalify the SPERT pins for use in converting university research reactors from the use of highly enriched uranium to the use of low-enriched uranium. The requalification program included visual and dimensional inspections of fuel pins and fuel pellets, radiographic inspections of welds, fill gas analyses, and chemical and spectrographic analyses of fuel and cladding materials. In general all attributes tested were within or very close to specified values, although some weld defects not covered by the original specifications were found. 1 ref., 4 figs., 11 tabs

  14. Determination of the decay constants and relative abundances of delayed neutrons by noise analysis in zero-power reactors

    International Nuclear Information System (INIS)

    Diniz, Ricardo

    2005-01-01

    A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters β i and λ i in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the β eff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)

  15. The Opportunities Map at Cornell University: finding direction in dairy production medicine.

    Science.gov (United States)

    Mitchell, Hilda M; Nydam, Daryl V; Reyher, Kristen; Gilbert, Robert O

    2004-01-01

    Discussion between faculty and interested students revealed the existence of a multitude of opportunities in dairy production medicine at the College of Veterinary Medicine at Cornell University. Many of these were not well known to students, or even to some of the faculty, and the means of accessing specific learning experiences were sometimes obscure. Together, an informal group of faculty, students, and alumni set about cataloging available educational opportunities, resulting in a 31-page publication referred to as the "Opportunities Map." Essentially a student handbook for production medicine students, the Opportunities Map at Cornell helps guide the travel of food animal-interested students through the curriculum without missing the important highlights along the way. The map was originally developed to chronicle the opportunities and resources available to students, but it has also been used to foster face-to-face communications between students and faculty, to welcome incoming students with production animal interests, and to provide a baseline description for further discussion about the curriculum.

  16. Zero-Power Radio Device.

    Energy Technology Data Exchange (ETDEWEB)

    Brocato, Robert W. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-02-01

    This report describes an unpowered radio receiver capable of detecting and responding to weak signals transmit ted from comparatively long distances . This radio receiver offers key advantages over a short range zero - power radio receiver previously described in SAND2004 - 4610, A Zero - Power Radio Receiver . The device described here can be fabricated as an integrated circuit for use in portable wireless devices, as a wake - up circuit, or a s a stand - alone receiver operating in conjunction with identification decoders or other electroni cs. It builds on key sub - components developed at Sandia National Laboratories over many years. It uses surface acoustic wave (SAW) filter technology. It uses custom component design to enable the efficient use of small aperture antennas. This device uses a key component, the pyroelectric demodulator , covered by Sandia owned U.S. Patent 7397301, Pyroelectric Demodulating Detector [1] . This device is also described in Sandia owned U.S. Patent 97266446, Zero Power Receiver [2].

  17. Gamma-ray spectrometric measurements of fission rate ratios between fresh and burnt fuel following irradiation in a zero-power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kröhnert, H., E-mail: hanna.kroehnert@ensi.ch [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland); Perret, G.; Murphy, M.F. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); Chawla, R. [Paul Scherrer Institut (PSI), CH-5232 Villigen (Switzerland); École Polytechnique Fédérale de Lausanne (EPFL), CH-1015 Lausanne (Switzerland)

    2013-01-11

    The gamma-ray activity from short-lived fission products has been measured in fresh and burnt UO{sub 2} fuel samples after irradiation in a zero-power reactor. For the first time, short-lived gamma-ray activity from fresh and burnt fuel has been compared and fresh-to-burnt fuel fission rate ratios have been derived. For the measurements, well characterized fresh and burnt fuel samples, with burn-ups up to 46 GWd/t, were irradiated in the zero-power research reactor PROTEUS. Fission rate ratios were derived based on the counting of high-energy gamma-rays above 2200 keV, in order to discriminate against the high intrinsic activity of the burnt fuel. This paper presents the measured fresh-to-burnt fuel fission rate ratios based on the {sup 142}La (2542 keV), {sup 89}Rb (2570 keV), {sup 138}Cs (2640 keV) and {sup 95}Y (3576 keV) high-energy gamma-ray lines. Comparisons are made with the results of Monte Carlo modeling of the experimental configuration, carried out using the MCNPX code. The measured fission rate ratios have 1σ uncertainties of 1.7–3.4%. The comparisons with calculated predictions show an agreement within 1–3σ, although there appears to be a slight bias (∼3%).

  18. Interactive Virtual Reactor and Control Room for Education and Training at Universities and Nuclear Power Plants

    International Nuclear Information System (INIS)

    Satoh, Yoshinori; Li, Ye; Zhu, Xuefeng; Rizwan, Uddin

    2014-01-01

    Efficient and effective education and training of nuclear engineering students and nuclear workers are critical for the safe operation and maintenance of nuclear power plants. With an eye toward this need, we have focused on the development of 3D models of virtual labs for education, training as well as to conduct virtual experiments. These virtual labs, that are expected to supplement currently available resources, and have the potential to reduce the cost of education and training, are most easily developed on game-engine platforms. We report some recent extensions to the virtual model of the University of Illinois TRIGA reactor

  19. Interactive Virtual Reactor and Control Room for Education and Training at Universities and Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Yoshinori; Li, Ye; Zhu, Xuefeng; Rizwan, Uddin [University of Illinois, Urbana (United States)

    2014-08-15

    Efficient and effective education and training of nuclear engineering students and nuclear workers are critical for the safe operation and maintenance of nuclear power plants. With an eye toward this need, we have focused on the development of 3D models of virtual labs for education, training as well as to conduct virtual experiments. These virtual labs, that are expected to supplement currently available resources, and have the potential to reduce the cost of education and training, are most easily developed on game-engine platforms. We report some recent extensions to the virtual model of the University of Illinois TRIGA reactor.

  20. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  1. Operating experience with the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Aderhold, H.C.

    1970-01-01

    As a result of our investigations, we believed the damage to be mechanical in origin and not to cladding failure. A new handling tool of modified design was put into service in July 1963, and since that time one element S/N 3075 has been dropped. This we believe was caused by operator error. At the request of prospective users, a high intensity, high energy gamma-ray irradiation facility has been added to the TRIGA equipment. This apparatus is simple to construct and use, either temporarily or permanently, with the TRIGA. Adjustment of relative neutron and gamma ray fluxes is possible by either shielding or changing rate of water flow. No attempt was made to improve performance by guiding water flow through the core, and higher yields should be obtainable by this means and by increasing the size of the holdup tank

  2. Modifications of the Cornell University TRIGA reactor thermal column

    International Nuclear Information System (INIS)

    Aderhold, Howard C.

    1984-01-01

    Full text: The thermal column has been modified to provide a horizontal beam suitable for neutron radiography. A hole of circular cross-section was cut along the axis of the thermal column through graphite and lead from the outer surface of the 'sliding block' to a point about 10 cm short of the curved end of the thermal column. The section through the sliding block is 15.2 cm in diameter and the remaining section is 10.2 cm in diameter. The outer or 15.2-cm section is lined with a 6-mm thick Boral sleeve, and the inner or 10.2-cm section contains a tapered collimator of Pd-Cd alloy inside a 3-mm thick Boral sleeve, a defining aperture formed by a 2.5-cm diameter hole in a 6-mm Boral plate, and, at the core end, a 5.1-cm thick bismuth absorber disk. All of these components are enclosed in an aluminum housing. From the sliding block outwards - a distance of 122 cm across the hohlraum and through the graphite at the exit end, plus another 122 cm through the rolling door - no drilling was necessary since removable plugs could be taken out to form a tapered hole of square cross section with stepped widths of 30.5, 33.0, and 38.1 cm. An aluminum housing fitting snugly in this hole is lined with Boral or a gadolinium-bearing paint and is joined with a rubber gasket to the inner housing, so that the entire length can be purged with helium. This system provides an l/d ratio of 1/140 and a useful beam area at the exposure point of 38 cm x 38 cm. At 480 kW, the neutron flux there is 10 6 n/cm 2 sec and the gamma flux is 17 R/hr. Figure 2 is a reproduction of the first radiograph taken with the facility. (author)

  3. Towards zero-power ICT

    Science.gov (United States)

    Gammaitoni, Luca; Chiuchiú, D.; Madami, M.; Carlotti, G.

    2015-06-01

    Is it possible to operate a computing device with zero energy expenditure? This question, once considered just an academic dilemma, has recently become strategic for the future of information and communication technology. In fact, in the last forty years the semiconductor industry has been driven by its ability to scale down the size of the complementary metal-oxide semiconductor-field-effect transistor, the building block of present computing devices, and to increase computing capability density up to a point where the power dissipated in heat during computation has become a serious limitation. To overcome such a limitation, since 2004 the Nanoelectronics Research Initiative has launched a grand challenge to address the fundamental limits of the physics of switches. In Europe, the European Commission has recently funded a set of projects with the aim of minimizing the energy consumption of computing. In this article we briefly review state-of-the-art zero-power computing, with special attention paid to the aspects of energy dissipation at the micro- and nanoscales.

  4. The text of the Agreement of 7 October 1983 between Cuba and the Agency for the application of safeguards in connection with the supply of a zero-power nuclear reactor from the Hungarian People's Republic

    International Nuclear Information System (INIS)

    1984-01-01

    The full text of the agreement of 7 October 1983 between Cuba and the Agency for the application of safeguards in connection with the supply of a zero-power nuclear reactor from the Hungarian People's Republic and to the nuclear material to be used therein to be supplied by the Union of Soviet Socialist Republics is presented

  5. Reactor power control device

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  6. Literacy in Action: A Carbon-Neutral Field Program at Cornell University

    Science.gov (United States)

    Moore, A.; Derry, L.

    2010-12-01

    The Cornell Earth and Environmental Systems (EES) Field Program is a semester-length undergraduate field program located on the island of Hawai`i. The Hawaiian Islands are the world’s most dynamic natural laboratory and the premier location for Earth systems research and education. While there are compelling reasons for students and faculty to travel from the US mainland to Hawai`i, the air and ground travel that comprises the program carries a large carbon footprint. This liability is also an extraordinary educational opportunity. For the past two years EES students have been challenged to make the program carbon-neutral. They are asked to devise a set of criteria for a credible and defensible zero-CO2 footprint and then to put their plan into action. The C-neutral project consists of three elements: (1) quantifying CO2 emissions, (2) reducing emissions wherever possible, and (3) offsetting emissions that cannot be eliminated. In quantifying emissions six areas are identified: air travel, ground travel, domestic electricity, natural gas, food, and waste. Emissions reductions include all of the standard “carpool--turn it down--turn it off “ conservation behaviors, with special emphasis on food and waste; eating local and organic, shopping at re-use centers, and compost and recycling of garbage. Our program facility utilizes solar hot water and is equipped with neither heat nor air conditioning, thus domestic energy use is low. Students tabulate all of our energy use and calculate the resulting CO2 emissions for all program participants for a period of four months. The CO2 offsetting strategy is conducted in collaboration with a native ecosystem restoration project. Students participate in all aspects of forest restoration, including seed collection, germination and outplanting of native plant species and removal of invasive pest species. The initial goal of this locally-supported project was to restore degraded pasture to native forest. The EES students have

  7. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

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

  8. Guide to power reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-15

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

  9. Reactor power measuring device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  10. University Reactor Instrumentation Program

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1992-11-01

    Recognizing that the University Reactor Instrumentation Program was developed in response to widespread needs in the academic community for modernization and improvement of research and training reactors at institutions such as the University of Florida, the items proposed to be supported by this grant over its two year period have been selected as those most likely to reduce foreed outages, to meet regulatory concerns that had been expressed in recent years by Nuclear Regulatory Commission inspectors or to correct other facility problems and limitations. Department of Energy Grant Number DE-FG07-90ER129969 was provided to the University of Florida Training Reactor(UFTR) facility through the US Department of Energy's University Reactor Instrumentation Program. The original proposal submitted in February, 1990 requested support for UFTR facility instrumentation and equipment upgrades for seven items in the amount of $107,530 with $13,800 of this amount to be the subject of cost sharing by the University of Florida and $93,730 requested as support from the Department of Energy. A breakdown of the items requested and total cost for the proposed UFTR facility instrumentation and equipment improvements is presented

  11. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Dogen, Ayumi; Ozawa, Michihiro.

    1983-01-01

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

  12. The first university research reactor in India

    International Nuclear Information System (INIS)

    Murthy, G.S.

    1999-01-01

    At low power research reactor is being set up in Andhra University to cater to the needs of researchers and isotope users by the Department of Atomic Energy in collaboration with Andhra University. This reactor is expected to be commissioned by 2001-02. Departments like Chemistry, Earth Sciences, Physics, Life Sciences, Pharmacy, Medicine and Engineering would be the beneficiaries of the availability of this reactor. In this paper, details of the envisaged research programme and training activities are discussed. (author)

  13. Reactor power control device

    International Nuclear Information System (INIS)

    Doi, Kazuyori.

    1981-01-01

    Purpose: To automatically control the BWR type reactor power by simple and short-time searching the load pattern nearest to the required pattern at a nuclear power plant side. Constitution: The reactor power is automatically regulated by periodical modifying of coefficients fitting to a reactor core model, according as a required load pattern. When a load requirement pattern is given, a simulator estimates the total power change and the axial power distribution change from a xenon density change output calculated by a xenon dynamic characteristic estimating device, and a load pattern capable of being realized is searched. The amount to be recirculated is controlled on the basis of the load patteren thus searched, and the operation of the BWR type reactor is automatically controlled at the side of the nuclear power plant. (Kamimura, M.)

  14. Arkansas Tech University TRIGA nuclear reactor

    International Nuclear Information System (INIS)

    Sankoorikal, J.; Culp, R.; Hamm, J.; Elliott, D.; Hodgson, L.; Apple, S.

    1990-01-01

    This paper describes the TRIGA nuclear reactor (ATUTR) proposed for construction on the campus of Arkansas Tech University in Russellville, Arkansas. The reactor will be part of the Center for Energy Studies located at Arkansas Tech University. The reactor has a steady state power level of 250 kW and can be pulsed with a maximum reactivity insertion of $2.0. Experience gained in dismantling and transporting some of the components from Michigan State University, and the storage of these components will be presented. The reactor will be used for education, training, and research. (author)

  15. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Hoizumi, Atsushi.

    1986-01-01

    Purpose: To grasp the margin for the limit value of the power distribution peaking factor inside the reactor under operation by using the reactor power distribution monitor. Constitution: The monitor is composed of the 'constant' file, (to store in-reactor power distributions obtained from analysis), TIP and thermocouple, lateral output distribution calibrating apparatus, axial output distribution synthesizer and peaking factor synthesizer. The lateral output distribution calibrating apparatus is used to make calibration by comparing the power distribution obtained from the thermocouples to the power distribution obtained from the TIP, and then to provide the power distribution lateral peaking factors. The axial output distribution synthesizer provides the power distribution axial peaking factors in accordance with the signals from the out-pile neutron flux detector. These axial and lateral power peaking factors are synthesized with high precision in the three-dimensional format and can be monitored at any time. (Kamimura, M.)

  16. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  17. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  18. Major accident analyses for experimental zero-power fast reactor assemblies; Analyse des accidents graves pouvant survenir dans les reacteurs experimentaux a neutrons rapides de puissance zero; Analiz krupnoj avarii dlya ehksperimental'ny kh reaktornykh ustanovok nulevoj moshchnosti na bystrykh nejtronakh; Analisis de los accidentes graves que pueden producirse en los reactores experimentales rapidos de potencia cero

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, G.; Barts, E. W.; Kapil, S.; Tomabechi, K. [Argonne National Laboratory, Argonne, IL (United States)

    1962-03-15

    A study has been made of the possibility, mechanism, and consequence of melt-down and other major nuclear accidents for a ZPR-III type experimental zero-power fast reactor of the two-half type. This study has been supplemented by an evaluation of the importance of the Doppler effect for a wide range of nuclear reactor assemblies for such a reactor. A melt-down event is highly improbable because of the restricted sequence of events which must be postulated. A discussion of the mechanism of the collapse is followed by the results of coupled neutronics-hydrodynamic s calculations for two zero-power assemblies. A 1200-l core has been examined because it represents a relatively large reactor of common core composition. A smaller core with a high-void fraction has been examined as a potentially more dangerous system. Very different time-wise behaviour has been found for the two systems. For sharp accidents in zero-power assemblies, the U{sup 235}-atoms, separated as plates of enriched uranium, will heat very rapidly while the remainder of the core remains essentially cold, so that a gas of U{sup 235}-vapour will provide the disassembly pressure. The adaption of the neutronics-hydrodynamic s code AX-I to the use of a Van der Waals gas is described. Another important change in the equation of state used in the code is to employ a Mie-Griineisen type equation derivable from solid state theory. This change provides a more satisfactory way to evaluate the pressure term for cores of variable composition. Because the highly enriched U{sup 235} plates of a zero-power assembly will heat much more rapidly than the depleted uranium plates, the possibility of a net positive Doppler effect is much larger for an experimental assembly than for the equivalent power breeder reactor. This hazard has been examined for a range of possible assemblies. These calculations indicate that the Doppler coefficient for a zero-power assembly does not become important as a hazard until one approaches

  19. University Reactor Sharing Program

    International Nuclear Information System (INIS)

    Reese, W.D.

    2004-01-01

    Research projects supported by the program include items such as dating geological material and producing high current super conducting magnets. The funding continues to give small colleges and universities the valuable opportunity to use the NSC for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy. The Reactor Sharing Program has supported the construction of a Fast Neutron Flux Irradiator for users at New Mexico Institute of Mining and Technology and the University of Houston. This device has been characterized and has been found to have near optimum neutron fluxes for A39/Ar 40 dating. Institution final reports and publications resulting from the use of these funds are on file at the Nuclear Science Center

  20. Thermodynamic analysis of the advanced zero emission power plant

    Directory of Open Access Journals (Sweden)

    Kotowicz Janusz

    2016-03-01

    Full Text Available The paper presents the structure and parameters of advanced zero emission power plant (AZEP. This concept is based on the replacement of the combustion chamber in a gas turbine by the membrane reactor. The reactor has three basic functions: (i oxygen separation from the air through the membrane, (ii combustion of the fuel, and (iii heat transfer to heat the oxygen-depleted air. In the discussed unit hot depleted air is expanded in a turbine and further feeds a bottoming steam cycle (BSC through the main heat recovery steam generator (HRSG. Flue gas leaving the membrane reactor feeds the second HRSG. The flue gas consist mainly of CO2 and water vapor, thus, CO2 separation involves only the flue gas drying. Results of the thermodynamic analysis of described power plant are presented.

  1. HOMOGENEOUS NUCLEAR POWER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  2. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  3. A Center of Excellence in the Mathematical Sciences - at Cornell University

    Science.gov (United States)

    1992-03-01

    S.L. PHOENIX 62 PAGES 87. GROEBNER BASES: THE ANCIENT SECRET MYSTIC POWER OF THE ALGU COMPUBRAICUS ,A REVELATION WHOSE SIMPLICITY WILL MAKE LADIES...Equations, October 1988; Groebner Basis, October 1988; Theoretical Aspects of Multiphase Flow, October 1988; Mathematical Theory of Queuing Systems

  4. Research progress in intense ion beam production for inertial confinement fusion at Cornell University

    International Nuclear Information System (INIS)

    Bluhm, H.; Greenly, J.B.; Hammer, D.A.

    1983-01-01

    Recent results obtained in the generation of intense pulsed light ion beams and their application to inertial confinement fusion are described. Studies of time-integrated and time-dependent beam divergence using a magnetically insulated ion diode with a ''flashboard'' anode at 11 W diode power show a directionality which is apparently due to electron dynamics in the diode. Nevertheless, ion beams having divergence angle as small as 0.5 0 have been produced at >10 8 W.cm - 2 . In another experiment with a similar diode, the anode plasma formation time varied with the detailed anode configuration, the diode voltage and the insulating magnetic field, with the longer times obtained at lower voltage and higher insulating magnetic field strength. The anode plasma density was determined to be in the 10 15 cm - 3 density range and to move away from the anode at approx.2 cm.μs - 1 in another similar experiment. Preliminary experiments performed on a 10 12 W accelerator show reasonable power coupling to a magnetically insulated ion diode, with >10 9 W.cm - 2 beams at approx.1.5 MV being generated. Computer simulations suggest that if such a beam can be focused into a plasma channel, most of its energy can be delivered to a pellet one to two metres away. In experiments on the applied Bsub(theta) diode, microwave radiation, ion production efficiency, and ion beam fluctuations all reach a maximum when the insulating magnetic field is about 1.4 times the critical field for magnetic insulation. Finally, relatively pure beams of heavy ions have been produced by making the anode with hydrocarbon-free dielectric material which contains the desired species together with other ions having substantially higher ionization potential. The sum of these results suggests that flashboard anodes operated at the few-MV level can be used to produce beams with properties suitable for inertial confinement fusion experiments on sufficiently powerful pulsed power generators. (author)

  5. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  6. Compact power reactor

    International Nuclear Information System (INIS)

    Wetch, J.R.; Dieckamp, H.M.; Wilson, L.A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector

  7. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  8. Power reactor design trends

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1985-01-01

    Cascade and Pulse Star represent new trends in ICF power reactor design that have emerged in the last few years. The most recent embodiments of these two concepts, and that of the HYLIFE design with which they will compare them, are shown. All three reactors depend upon protecting structural elements from neutrons, x rays and debris by injecting massive amounts of shielding material inside the reaction chamber. However, Cascade and Pulse Star introduce new ideas to improve the economics, safety, and environmental impact of ICF reactors. They also pose different development issues and thus represent technological alternatives to HYLIFE

  9. Local, zero-power void coefficient measurements in the ACPR

    Energy Technology Data Exchange (ETDEWEB)

    Rivard, J B; Thome, F V [Sandia Laboratories (United States)

    1974-07-01

    Changes in reactivity may be stimulated in the ACPR by the local introduction of voids into the reactor coolant. The local void coefficients of reactivity which describe this effect are of interest from a reactor safety point-of-view, and their determination is the subject of this presentation. Bottled nitrogen gas was used to produce the voids. The gas was forced out of a small diameter tube which was positioned vertically in the core lattice with its open end below the fuel. The gas was passed through a pressure regulator, a valve, and a flowmeter to establish a steady flow condition, following which a delayed-critical (zero-power) reactor state was established. Correlation of the average volume of core void created by the nitrogen flow with the reactivity worth of the delayed-critical control-rod bank position produced the values of the zero-power void coefficients of reactivity. The void coefficients were determined at various core positions from {approx}6 mm to 142 mm beyond the central irradiation space and for three different flow rates. For the range of void fractions investigated, these coefficients are negative, with values ranging between -$0.02 and -$0.12. Tabular and graphical results of the measurements are presented, and details of the coefficient determination are explained. (author)

  10. Local, zero-power void coefficient measurements in the ACPR

    International Nuclear Information System (INIS)

    Rivard, J.B.; Thome, F.V.

    1974-01-01

    Changes in reactivity may be stimulated in the ACPR by the local introduction of voids into the reactor coolant. The local void coefficients of reactivity which describe this effect are of interest from a reactor safety point-of-view, and their determination is the subject of this presentation. Bottled nitrogen gas was used to produce the voids. The gas was forced out of a small diameter tube which was positioned vertically in the core lattice with its open end below the fuel. The gas was passed through a pressure regulator, a valve, and a flowmeter to establish a steady flow condition, following which a delayed-critical (zero-power) reactor state was established. Correlation of the average volume of core void created by the nitrogen flow with the reactivity worth of the delayed-critical control-rod bank position produced the values of the zero-power void coefficients of reactivity. The void coefficients were determined at various core positions from ∼6 mm to 142 mm beyond the central irradiation space and for three different flow rates. For the range of void fractions investigated, these coefficients are negative, with values ranging between -$0.02 and -$0.12. Tabular and graphical results of the measurements are presented, and details of the coefficient determination are explained. (author)

  11. Neutrino Mass Models: impact of non-zero reactor angle

    International Nuclear Information System (INIS)

    King, Stephen F.

    2011-01-01

    In this talk neutrino mass models are reviewed and the impact of a non-zero reactor angle and other deviations from tri-bi maximal mixing are discussed. We propose some benchmark models, where the only way to discriminate between them is by high precision neutrino oscillation experiments.

  12. Reactor power distribution monitor

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To improve the performance and secure the safety of a nuclear reactor by rapidly computing and display the power density in the nuclear reactor by using a plurality of processors. Constitution: Plant data for a nuclear reactor containing the measured values from a local power monitor LPRM are sent and recorded in a magnetic disc. They are also sent to a core performance computer in which burn-up degree distribution and the like are computed, and the results are sent and recorded in the magnetic disc. A central processors loads programs to each of the processors and applies data recorded in the magnetic disc to each of the processors. Each of the processors computes the corresponding power distribution in four fuel assemblies surrounding the LPRM string by the above information. The central processor compiles the computation results and displays them on a display. In this way, power distribution in the fuel assemblies can rapidly be computed to thereby secure the improvement of the performance and safety of the reactor. (Seki, T.)

  13. Nuclear power reactor safety

    International Nuclear Information System (INIS)

    Pon, G.A.

    1976-10-01

    This report is based on the Atomic Energy of Canada Limited submission to the Royal Commission on Electric Power Planning on the safety of CANDU reactors. It discusses normal operating conditions, postulated accident conditions, and safety systems. The release of radioactivity under normal and accident conditions is compared to the limits set by the Atomic Energy Control Regulations. (author)

  14. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

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

  15. Feasibility study of the university of Utah TRIGA reactor power upgrade - part II: Thermohydraulics and heat transfer study in respect to cooling system requirements and design

    Directory of Open Access Journals (Sweden)

    Babitz Philip

    2013-01-01

    Full Text Available The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much more pronounced at higher power levels with vortex shedding also occurring. A departure from nucleate boiling analysis showed that while nucleate boiling begins near 210 kW it remains in this state and does not approach the critical heat flux at powers up to 500 kW. Based on these studies, two upgrades are proposed for extended operation and possibly higher reactor power level. Together with the findings from Part I studies, we conclude that increase of the reactor power is highly feasible yet dependable on its purpose and associated investments.

  16. Zero energy reactor RB technical characteristics and experimental possibilities

    Energy Technology Data Exchange (ETDEWEB)

    Jovanovic, S; Takac, S; Raisic, N; Lolic, B; Markovic, H [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1963-04-15

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility.

  17. Zero energy reactor RB technical characteristics and experimental possibilities

    International Nuclear Information System (INIS)

    Jovanovic, S.; Takac, S.; Raisic, N.; Lolic, B.; Markovic, H.

    1963-04-01

    The zero energy reactor RB was constructed in 1958 in accordance with the nuclear reactor development programme of the Boris Kidric Institute of Nuclear Sciences. The reactor was in operation until the middle of 1959 when the heavy water, serving as the moderator, was transported to the high flux reactor RA, built at the same time at the Boris Kidric Institute. Owing to the fact that the purchase of new quantities of heavy water was planned for 1961 it was decided to reconstruct the RB reactor in order to improve the safety of the system and to obtain better flexibility in performing the experiments. New control, safety and radiation monitoring systems were constructed. Some changes were also made on the reactor tank, water circulation system and the water level monitoring equipment. The reconstruction was completed in 1961. and the heavy water was delivered early in 1962. The reconstructed reactor was critical for the first time in summer 1962, and from that time was in continuous operation. This report presents an outline of the design and construction characteristics of the reactor. The main intention is to inform potential users of the reactor about experimental possibilities, advantages and disadvantages of such a critical facility

  18. Status of Japanese university reactors

    International Nuclear Information System (INIS)

    Fujita, Yoshiaki

    1999-01-01

    Status of Japanese university reactors, their role and value in research and education, and the spent fuel problem are presented. Some of the reactors are now faced by severe difficulties in continuing their operation services. The point of measures to solve the difficulties is suggested. (author)

  19. DOE/university reactor sharing

    International Nuclear Information System (INIS)

    Young, H.H.

    1985-01-01

    The objective of the US Department of Energy's program of reactor sharing is to strengthen nuclear science and engineering instruction and nuclear research opportunities in non-reactor-owning colleges and universities. The benefits of the program and need for the continuation of the program in the future are discussed

  20. Reactor power monitoring device

    International Nuclear Information System (INIS)

    Kono, Shigehiro.

    1990-01-01

    Among a plurality of power monitoring programs in a reactor power monitoring device, rapid response is required for a scram judging program for the power judging processing of scram signals. Therefore, the scram judging program is stored independently from other power monitoring programs, applied with a priority order, and executed in parallel with other programs, to output scram signals when the detected data exceeds a predetermined value. As a result, the capacity required for the scram judging program is reduced and the processing can be conducted in a short period of time. In addition, since high priority is applied to the scram judging program which is divided into a small capacity, it is executed at higher frequency than other programs when they are executed in parallel. That is, since the entire processings for the power monitoring program are repeated in a short cycle, the response speed of the scram signals required for high responsivity can be increased. (N.H.)

  1. Design and safety aspects of the Cornell cold neutron source

    International Nuclear Information System (INIS)

    Ouellet, Carol G.; Clark, David D.

    1992-01-01

    The cold neutron beam facility at the Cornell University TRIGA Mark II reactor will begin operational testing in early 1993. It is designed to provide a low background subthermal neutron beam that is as free as possible of fast neutrons and gamma rays for applied research and graduate-level instruction. The Cornell cold neutron source differs from the more conventional types of cold sources in that it is inherently safer because it uses a safe handling material (mesitylene) as the moderator instead of hydrogen or methane, avoids the circulation of cryogenic fluids by removing heat from the system by conduction through a 99.99% pure copper rod attached to a cryogenic refrigerator, and is much smaller in its size and loads. The design details and potential hazards are described, where it is concluded that no credible accident involving the cold source could cause damage to the reactor or personnel, or cause release of radioactivity. (author)

  2. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    Barjon, Robert

    1975-01-01

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude [fr

  3. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

  4. The first university research reactor in India

    International Nuclear Information System (INIS)

    Murty, G.S.

    1999-01-01

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  5. The first university research reactor in India

    Energy Technology Data Exchange (ETDEWEB)

    Murty, G.S. [Co-ordinator, Low Power Research Reactor, Andhra Univ., Visakapatnam (India)

    1999-08-01

    As the first university research reactor in India, the low power, pool type with fixed core and low enriched uranium fuel research reactor is under construction in the Andhra university campus, Andhra Pradesh, India. The reactor is expected to be commissioned during 2001-2002. The mission of the reactor is to play the research center as a regional research facility catering to the needs of academic institutions and industrial organizations of this region of the country. Further, to encourage interdisplinary and multidisplinary research activities, to supply radioisotope and labelled compounds to the user institutions and to create awareness towards the peaceful uses of atomic energy. This report describes its objectives, status and future plans in brief. (H. Itami)

  6. Power reactor noise

    International Nuclear Information System (INIS)

    Thie, J.A.

    1981-01-01

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

  7. Nuclear reactor power supply

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

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

  8. MIT research reactor. Power uprate and utilization

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Lin-Wen [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, Cambridge, MA (United States)

    2012-03-15

    The MIT Research Reactor (MITR) is a university research reactor located on MIT campus. and has a long history in supporting research and education. Recent accomplishments include a 20% power rate to 6 MW and expanding advanced materials fuel testing program. Another important ongoing initiative is the conversion to high density low enrichment uranium (LEU) monolithic U-Mo fuel, which will consist of a new fuel element design and power increase to 7 MW. (author)

  9. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Brooks, J.N.

    1978-01-01

    A tokamak experimental power reactor has been designed that is capable of producing net electric power over a wide range of possible operating conditions. A net production of 81 MW of electricity is expected from the design reference conditions that assume a value of 0.07 for beta-toroidal, a maximum toroidal magnetic field of 9 T and a thermal conversion efficiency of 30%. Impurity control is achieved through the use of a low-Z first wall coating. This approach allows a burn time of 60 seconds without the incorporation of a divertor. The system is cooled by a dual pressurized water/steam system that could potentially provide thermal efficiencies as high as 39%. The first surface facing the plasma is a low-Z coated water cooled panel that is attached to a 20 cm thick blanket module. The vacuum boundary is removed a total of 22 cm from the plasma, thereby minimizing the amount of radiation damage in this vital component. Consideration is given in the design to the possible use of the EPR as a materials test reactor. It is estimated that the total system could be built for less than 550 million dollars

  10. The GUINEVERE-project: the first zero-power fast lead reactor coupled to a 14 MeV neutron generator (GENEPI)

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    The GUINEVERE project is an European project in the framework of FP6 IP-EUROTRANS. The IP-EUROTRANS project aims at addressing the main issues for ADS development in the framework of partitioning and transmutation for nuclear waste volume and radio toxicity reduction. The GUINEVERE-project is carried out in the context of domain 2 of IP-EUROTRANS, ECATS, devoted to specific experiments for the coupling of an accelerator, a target and a subcritical core. These experiments should provide an answer to the questions of on-line reactivity monitoring, sub-criticality determination and operational procedures (loading, start-up, shut-down) in an ADS by 2009-2010. During the definition of the experimental programme ECATS, it was judged that there was a strong need for a European managed experiment in the line of the FP5 MUSE-project. Reanalyzing the outcome of MUSE, two points were left open for significant improvement. To validate the methodology for reactivity monitoring, a continuous beam is needed, which was not present in the MUSE-project. In the definition of the MUSE-project, from the beginning a strong request was made for a lead core in order to have representative conditions of a lead-cooled ADS which was only partially answered by the MUSE-programme. Therefore, there is a need for a lead fast critical facility connected to a continuous beam accelerator. Since such a programme/installation is not present at the European nor at the international level, SCK-CEN has proposed to use a modified VENUS critical facility located at its Mol-site and to couple it to a modified GENEPI deuteron accelerator (used in MUSE) working in current mode delivering 14 MeV neutrons by bombardment of deuterons on a tritium-target: the GUINEVERE-project (Generator of Uninterrupted Intense NEutrons at the lead VEnus REactor). This proposal was formally accepted by the Governing Council of IP-Eurotrans in December 2006. This project represents a close collaboration between SCK-CEN, CEA and

  11. Power reactors in member states

    International Nuclear Information System (INIS)

    1975-01-01

    This is the first issue of a periodical computer-based listing of civilian nuclear power reactors in the Member States of the IAEA, presenting the situation as of 1 April 1975. It is intended as a replacement for the Agency's previous annual publication of ''Power and Research Reactors in Member States''. In the new format, the listing contains more information about power reactors in operation, under construction, planned and shut down. As far as possible all the basic design data relating to reactors in operation have been included. In future these data will be included also for other power reactors, so that the publication will serve to give a clear picture of the technical progress achieved. Test and research reactors and critical facilities are no longer listed. Of interest to nuclear power planners, nuclear system designers, nuclear plant operators and interested professional engineers and scientists

  12. Fast reactors in nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Kazachkovskii, O

    1981-02-01

    The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

  13. Tokamak reactor startup power

    International Nuclear Information System (INIS)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

  14. Fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Vold, E.L.; Conn, R.W.

    1986-01-01

    Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

  15. Calibration of RB reactor power

    International Nuclear Information System (INIS)

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

    1976-09-01

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

  16. Tendencies in operating power reactors

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1987-01-01

    A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

  17. Tokamak experimental power reactor

    International Nuclear Information System (INIS)

    Stacey, W.M. Jr.; Abdou, M.A.; Bertoncini, P.J.

    1976-01-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 yr. The EPR operates in a pulsed mode at a frequency of approximately 1/min, with approximately 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2 cm thick stainless steel, and has 2 cm thick detachable, beryllium-coated coolant panels mounted on the interior. A 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H 2 O. Sixteen niobium-titanium superconducting toroidal field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic heating and equilibrium field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam injectors which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-convertors

  18. Cell heterogeneity problems in the analysis of zero power experiments

    International Nuclear Information System (INIS)

    Grimstone, M.J.; Stevenson, J.M.

    1979-01-01

    Methods are described for treating plate and pin cell heterogeneity in the preparation of broad group cross-sections used in the analysis of zero power fast reactor experiments. Methods used at Karlsruhe and Winfrith are summarised and compared, with particular reference to the treatment of resonance shielding, the calculation of broad group spatial fine structure, the treatment of leakage and the calculation of anisotropic diffusion coefficients. The problems of cells near boundaries such as core-breeder interfaces and of singularities such as control rods are also considered briefly. Numerical studies carried out to investigate approximations in the methods are described. These include tests of the accuracy of one-dimensional cell modelling techniques, and the validation by Monte Carlo of methods for treating streaming in the calculation of diffusion coefficients. Comparisons are shown between the heterogeneity effects calculated by the Karlsruhe and Winfrith methods for typical pin and plate cells used in the BIZET experimental programme, and their effect in a whole reactor calculation is indicated. Comparisons are given with measurements which provide tests of the heterogeneity calculations. These include reaction rate scans within pin and plate cells, and reaction rate measurements across sectors of pin and plate fuel, where the flux tilt is determined by the relative reactivity of the pin and plate cells. Finally, the heterogeneity problems arising in the interpretation of reaction rate measurements are discussed. (author)

  19. Simulations at Cornell

    International Nuclear Information System (INIS)

    Avery, P.

    1992-01-01

    I summarize recent work at Cornell on B Factory simulations. My talk includes a brief review of the Cornell proposal, some estimates we have developed for measuring sinγ and a progress report on the use of kinematic fitting. Because of the recent accumulation of data at CLEO II, the author devote a substantial part of my summary to recent B physics results, especially those connected with CP violation

  20. New training reactor at Dresden Technical University

    International Nuclear Information System (INIS)

    Hansen, W.; Knorr, J.; Wolf, T.

    2006-01-01

    A total of 14 low-power (up to 10 W) training reactors have been operated at German universities, 9 of them officially classified as being operational in 2004, though for very different uses. This number is expected to drop sharply. The only comprehensive upgrading of a training reactor took place at Dresden Technical University: AKR-2, the most modern facility in Germany, started routine operation in April 2005, under a newly granted license pursuant to Sec. 7, Subsec. 1 of the German Atomic Energy Act, for training students in nuclear technology, for suitable research projects, and a a center of information about reactor technology and nuclear technology for the interested public. One special aspect of this refurbishment was the installation of digital safety I and C systems of the TELEPERM XS line, which are used also in other modern plants. This fact, plus the easy possibility to use the plant for many basic experiments in reactor physics and radiation protection, make the AKR-2 attractive also to other users (e.g. for training reactor personnel or other persons working in nuclear technology). (orig.)

  1. Safety of nuclear power reactors

    International Nuclear Information System (INIS)

    MacPherson, H.G.

    1982-01-01

    Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

  2. A nuclear power reactor

    International Nuclear Information System (INIS)

    Borrman, B.E.; Broden, P.; Lundin, N.

    1979-12-01

    The invention consists of shock absorbing support beams fastened to the underside of the reactor tank lid of a BWR type reactor, whose purpose is to provide support to the steam separator and dryer unit against accelerations due to earthquakes, without causing undue thermal stresses in the unit due to differential expansion. (J.I.W.)

  3. Program of RA reactor start-up to nominal power

    International Nuclear Information System (INIS)

    1959-01-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is described in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members

  4. 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

  5. Fractals in Power Reactor Noise

    International Nuclear Information System (INIS)

    Aguilar Martinez, O.

    1994-01-01

    In this work the non- lineal dynamic problem of power reactor is analyzed using classic concepts of fractal analysis as: attractors, Hausdorff-Besikovics dimension, phase space, etc. A new non-linear problem is also analyzed: the discrimination of chaotic signals from random neutron noise signals and processing for diagnosis purposes. The advantages of a fractal analysis approach in the power reactor noise are commented in details

  6. Power reactors in Member States. 1978 edition

    International Nuclear Information System (INIS)

    1978-01-01

    The computer-based reactor listing gives information on reactor core characteristics and plant systems for all power reactors in operation under construction and planned. The following two tables are included to give a general picture of the overall situation: Reactor types and net electrical power; Reactor units and net electrical power by country and cumulated by year

  7. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    1989-06-01

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

  8. Oregon State University TRIGA Reactor annual report

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

    1979-08-31

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included.

  9. Oregon State University TRIGA Reactor annual report

    International Nuclear Information System (INIS)

    Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

    1979-01-01

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included

  10. Power oscillations in BWR reactors

    International Nuclear Information System (INIS)

    Espinosa P, G.

    2002-01-01

    One of the main problems in the operation of BWR type reactors is the instability in power that these could present. One type of oscillations and that is the objective of this work is the named density wave, which is attributed to the thermohydraulic processes that take place in the reactor core. From the beginnings of the development of BWR reactors, the stability of these has been an important aspect in their design, due to its possible consequences on the fuel integrity. The reactor core operates in two phase flow conditions and it is observed that under certain power and flow conditions, power instabilities appear. Studying this type of phenomena is complex, due to that a reactor core is constituted approximately by 27,000 fuel bars with different distributions of power and flow. The phenomena that cause the instability in BWR reactors continue being matter of scientific study. In the literature mainly in nuclear subject, it can be observed that exist different methods and approximations for studying this type of phenomena, nevertheless, their results are focused to establish safety limits in the reactor operation, instead of studying in depth of the knowledge about. Also in this line sense of the reactor data analysis, the oscillations characteristic frequencies are obtained for trying to establish if the power is growing or decreasing. In addition to that before mentioned in this paper it is presented a rigorous study applying the volumetric average method, for obtaining the vacuum waves propagation velocities and its possible connection with the power oscillations. (Author)

  11. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    Spiegelberg, R.

    1992-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-15

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

  13. Nuclear power reactor technology

    International Nuclear Information System (INIS)

    1978-09-01

    Risoe National Laboratory was established more than twenty years ago with research and development of nuclear reactor technology as its main objective. The Laboratory has by now accumulated many years of experience in a number of areas vital to nuclear reactor technology. The work and experience of, and services offered by the Laboratory within the following fields are described: Health physics site supervision; Treatment of low and medium level radioactive waste; Core performance evaluation; Transient analysis; Accident analysis; Fuel management; Fuel element design, fabrication and performance evaluation; Non-destructive testing of nuclear fuel; Theoretical and experimental structural analysis; Reliability analysis; Site evaluation. Environmental risk and hazard calculation; Review and analysis of safety documentation. Risoe has already given much assistance to the authorities, utilities and industries in such fields, carrying out work on both light and heavy water reactors. The Laboratory now offers its services to others as a consultant, in education and training of staff, in planning, in qualitative and quantitative analysis, and for the development and specification of fabrication techniques. (author)

  14. Cascade ICF power reactor

    International Nuclear Information System (INIS)

    Hogan, W.J.; Pitts, J.H.

    1986-01-01

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO 2 . The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor

  15. Magnet powering with zero downtime - a dream?

    CERN Document Server

    Zerlauth, Markus

    2012-01-01

    Despite a number of improvements already applied in the course of the year, the magnet powering system of the LHC still accounts for around 50% of the premature beam dumps. This number might even further increase when moving to higher beam energies in the next years. With mitigations of radiation effects and the prospects for beam induced magnet quenches being discussed elsewhere, we aim at identifying possible mid- and long-term improvements within the various equipment systems to further reduce the number of equipment failures leading to a loss of the particle beams. Amongst others, this includes the sensitivity of equipment to external causes such as electromagnetic perturbations or perturbations on the electrical network. To conclude, the gain of the identified mitigations will have to be balanced against the potential impact on schedule and cost.

  16. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

    1987-01-01

    Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

  17. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  18. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, ''walk-away safe'' design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (OandM) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  19. Low power unattended defense reactor

    International Nuclear Information System (INIS)

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System

  20. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    Regulatory and research evaluations of embrittlement predication models and of pressure vessel integrity can be greatly expedited by the use of a well-designed, computerized data base. The Power Reactor Embrittlement Data Base (PR-EDB) is such a comprehensive collection of data for US commercial nuclear reactors. The Nuclear Regulatory Commission (NRC) has provided financial support, and the Electric Power Research Institute (EPRI) has provided technical assistance in the quality assurance (QA) of the data to establish an industry-wide data base that will be maintained and updated on a long-term basis. Successful applications of the data base to several of NRC's evaluations have received favorable response and support for its continuation. The future direction of the data base has been designed to include the test reactor and other types of data of interest to the regulators and the researchers. 1 ref

  1. Power reactor core safety research

    International Nuclear Information System (INIS)

    Rim, C.S.; Kim, W.C.; Shon, D.S.; Kim, J.

    1981-01-01

    As a part of nuclear safety research program, a project was launched to develop a model to predict fuel failure, to produce the data required for the localizaton of fuel design and fabrication technology, to establish safety limits for regulation of nuclear power plants and to develop reactor operation method to minimize fuel failure through the study of fuel failure mechanisms. During 1980, the first year of this project, various fuel failure mechanisms were analyzed, an experimental method for out-of-pile tests to study the stress corrosion cracking (SCC) behaviour of Zircaloy cladding underiodine environment was established, and characteristics of PWR and CANDU Zircaloy specimens were examined. Also developed during 1980 were the methods and correlations to evaluate fuel failures in the reactor core based on operating data from power reactors

  2. Experimental power reactor

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The following five topics are discussed using figures and diagrams: (1) energy storage and transfer program, (2) thermomechanical analysis, (3) a steam dual-cycle power conversion system for the EPR, (4) EPR tritium facility scoping studies, and (5) vacuum systems

  3. Cornell and Marseille

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    A major theme running through the Cornell meeting, as with most physics meetings these days, was the need to look hard at suppressed or forbidden reactions for signs of an excess or positive signal, suggesting that at last something new might be happening. At Cornell, Jack Ritchie (Texas) covered the kaon front, while Anthony Sanda (Nagoya) gave an overall status report. Rare processes mediated by second order effects (two quarks mechanisms coupled back-to-back) have been seen by the CLEO detector at Cornell (June page 1), and Sanda suggested that other such effects could now begin to show up. Last year's major international meeting at Dallas marked the debut of physics results from the Zeus and H1 experiments at the new and unique HERA electron-proton collider at DESY, Hamburg. This year HERA's collision rate is climbing, and the entire 1992 collision score was matched in just one weekend at the end of July. Cornell's plenary-only programme featured back-to-back presentations from H1 (John Dainton, Liverpool) and Zeus (John Martin, Toronto). Both experiments are beginning to see signs of the tight particle clusters ('jets') indicative of constituent quark/ gluon interactions deep inside the 'target' protons. HERA probes the structure of the proton in an unexplored kinematical region (quark momentum fraction x approaching 10 -4 ), and H1 was the first to see how this structure develops as x decreases. As the proton is probed in increasingly finer detail, it shows a richer quark content as more transient ('virtual') quark/gluon contributions come into play. This structure evolution provides important information for quark field theory, with additional mechanisms opening up

  4. CORNELL: CESR and beyond

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The electron-positron storage ring CESR, at Cornell is now operating regularly with a luminosity of over 10 31 per cm 2 per s (400 inverse nanobarns) and, thanks to the money saved by operating the magnet of the CLEO detector with a superconducting coil, the number of hours available for physics is not restricted. The research remains concentrated around the upsilon resonances (operating the storage ring at just over 5 GeV per beam). (orig./HSI).

  5. Tokamak power reactor ignition and time dependent fractional power operation

    International Nuclear Information System (INIS)

    Vold, E.L.; Mau, T.K.; Conn, R.W.

    1986-06-01

    A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transport power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve

  6. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  7. Reactor power peaking information display

    International Nuclear Information System (INIS)

    Book, T.L.; Kochendarfer, R.A.

    1986-01-01

    This patent describes a system for monitoring operating conditions within a nuclear reactor. The system consists of a method for measuring the operating parameters within the nuclear reactor, including the position of axial power shaping rods and regulating control rod. It also includes a method for determining from the operating parameters the operating limits before a power peaking condition exists within the nuclear reactor, and a method for displaying the operating limits which consists of a visual display permitting the continuous monitoring of the operating conditions within the nuclear reactor as a graph of the shaping rod position vs the regulating rod position having a permissible area and a restricted area. The permissible area is further divided into a recommended operating area for steady state operation and a cursor located on the graph to indicate the present operating condition of the nuclear reactor to allow an operator to view any need for corrective action based on the movement of the cursor out of the recommended operating area and to take any corrective transient action within the permissible area

  8. Reactor power region measuring device

    International Nuclear Information System (INIS)

    Kashiwa, Takao.

    1996-01-01

    The device of the present invention can rapidly detect abnormality of a local power region monitor (LPRM) even at a low power region caused such as upon start-up of a BWR type reactor. Namely, the present invention comprises (1) an LPRM detector for measuring neutron fluxes in the reactor, (2) a gamma thermo detector for calibrating the sensitivity of the LPRM detector, (3) a comparison circuit for comparing the detected values of the detectors (1) and (2), and (4) an alarm circuit for outputting an alarm when the comparative difference of the output of the circuit (3) exceeds a predetermined value. Signals of an alarm for a lower limit of the LPRM detector have been issued continuously upon start-up and shut down of the reactor since neutron fluxes in the reactor are reduced. However, the gamma thermo detector is always secured in the inside of the reactor different from a travelling-type incore probe monitor (TIP) disposed so far for the same purpose. Accordingly, the alarm generated upon usual start-up can be eliminated by comparing the detected values of the detector (2) and abnormality of the detector (1) can be rapidly detected by judging the abnormality of the comparative difference. (I.S.)

  9. Time-zero efficiency of European power derivatives markets

    International Nuclear Information System (INIS)

    Peña, Juan Ignacio; Rodriguez, Rosa

    2016-01-01

    We study time-zero efficiency of electricity derivatives markets. By time-zero efficiency is meant a sequence of prices of derivatives contracts having the same underlying asset but different times to maturity which implies that prices comply with a set of efficiency conditions that prevent profitable time-zero arbitrage opportunities. We investigate whether statistical tests, based on the law of one price, and trading rules, based on price differentials and no-arbitrage violations, are useful for assessing time-zero efficiency. We apply tests and trading rules to daily data of three European power markets: Germany, France and Spain. In the case of the German market, after considering liquidity availability and transaction costs, results are not inconsistent with time-zero efficiency. However, in the case of the French and Spanish markets, limitations in liquidity and representativeness are challenges that prevent definite conclusions. Liquidity in French and Spanish markets should improve by using pricing and marketing incentives. These incentives should attract more participants into the electricity derivatives exchanges and should encourage them to settle OTC trades in clearinghouses. Publication of statistics on prices, volumes and open interest per type of participant should be promoted. - Highlights: •We test time-zero efficiency of derivatives power markets in Germany, France and Spain. •Prices in Germany, considering liquidity and transaction costs, are time-zero efficient. •In France and Spain, limitations in liquidity and representativeness prevent conclusions. •Liquidity in France and Spain should improve by using pricing and marketing incentives. •Incentives attract participants to exchanges and encourage them to settle OTC trades in clearinghouses.

  10. Surveillance of nuclear power reactors

    International Nuclear Information System (INIS)

    Marini, J.

    1983-01-01

    Surveillance of nuclear power reactors is now a necessity imposed by such regulatory documents as USNRC Regulatory Guide 1.133. In addition to regulatory requirements, however, nuclear reactor surveillance offers plant operators significant economic advantages insofar as a single day's outage is very costly. The economic worth of a reactor surveillance system can be stated in terms of the improved plant availability provided through its capability to detect incidents before they occur and cause serious damage. Furthermore, the TMI accident has demonstrated the need for monitoring certain components to provide operators with clear information on their functional status. In response to the above considerations, Framatome has developed a line of products which includes: pressure vessel leakage detection systems, loose part detection systems, component vibration monitoring systems, and, crack detection and monitoring systems. Some of the surveillance systems developed by Framatome are described in this paper

  11. Advances in small zero-leak valves point to better nuclear power-plant reliability

    Energy Technology Data Exchange (ETDEWEB)

    Eacott, K B; Kin, J C; Hotta, Y [Dresser Japan, Ltd.

    1978-04-01

    In the selection of small valves less than two inches used for nuclear power plants, sufficient consideration must be given to the reliability to radioactive material, the easy operability, and the significant function, especially zero leak. These valves are classified into bellows and diaphragm seal types which must satisfy zero leak, 4000 cycles life test and good maintainability. Welded bellows, formed bellows, and metal diaphragms are actually used for these requirements. The construction of these types are shown. The requirements and principal specifications for these small valves are explained, and some examples are given. These zero leak valves are installed in reactor coolant loop system, borated water from B. A. system, pressurizer instrument system, containment spray system, high head system and off gas system for PWRS, and main steam line system, diesel generator cooling water system, re-circulation system, clean up water system, etc. for BWRS.

  12. Measurement of the physics properties of gas-cooled fast reactors in the zero energy reactor PROTEUS and analysis of the results

    International Nuclear Information System (INIS)

    Richmond, R.

    1982-12-01

    The main aim of the fast reactor physics measurements carried out in the zero energy reactor PROTEUS was to check the performance of data sets and calculation methods used in the design of fast breeder reactors. This allowed the accuracy of the power reactor calculations to be determined and enabled an assessment to be made of whether this accuracy would be sufficient to allow the design, construction and licensing of the GCFR power reactor. In order to carry out the physics measurements an existing zero energy reactor was converted to a form in which a central fast reactor lattice was surrounded by thermal zones to drive the reactor critical. One of the most important measuring techniques used to check the performance of data sets and calculation methods was the determination of reaction rate ratios and, by using an appropriate range of nuclides, it was possible to obtain a detailed picture covering 70% of reactions taking place in the central part of the fast reactor zone and with an accuracy of +-1.5% in a typical ratio. A further technique used during the work on GCFR-PROTEUS was the measurement of neutron spectrum which was carried out in a wide range of environments and, in the later stages of the work, covered the energy range from 9 keV to 2.3 MeV. These measurements, in particular, indicated significant errors in the FGL4 scattering cross-sections. A third technique, which was developed to a high degree of accuracy, was the measurement of reactivity worths. This was used in measurements of the worths of small samples and also in the application of the null reactivity technique to determine k-infinity and hence the absorption cross-sections of reactor structural materials. (Auth.)

  13. CORNELL: Synchrotron 25

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    A recent celebration marked the twenty-fifth anniversary of the Cornell Electron Synchrotron. The major milestone in the commissioning of the synchrotron was on October 11, 1967 when Helen Edwards, Boyce McDaniel, and Maury Tigner achieved a 7 GeV beam, a worldrecord energy for electron synchrotrons at that time. Like so many advances in experimental physics, this occurred early in the morning - 3 a.m.! The transition from accelerator commissioning to high energy physics operation was extremely rapid; 7 GeV operation for data collection was routine just five weeks later. Throughout its life as a source of photon and electron beams for fixed target experiments, the synchrotron maintained energy leadership for circular electron machines. Originally designed for operation at 10 GeV, eventually it consistently provided beams for experiments at energies up to 11.6 GeV. It now operates at 5 GeV, serving as the injector for the CESR electron-positron storage ring. Robert Wilson was director of the laboratory during the design and most of the construction of the machine. He left near the end of the construction to become the first director of Fermilab and was replaced by Boyce McDaniel, who guided the laboratory from the completion of the synchrotron to the construction and early operation of CESR. Wilson recalled how the laboratory had originally proposed a 3 GeV turnkey machine to be built entirely by industry and would fit in the space previously occupied by earlier Cornell accelerators. However, members of the laboratory realized that 3 GeV would not open new physics frontiers, that the construction of the accelerator was much of the fun of doing high energy physics experiments, and that a more challenging project was needed. This led to the proposal for the 10 GeV synchrotron which was built in the ''Cornell Style'' with many of the components fabricated and nearly all of the assembly done at Cornell

  14. CORNELL: Synchrotron 25

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1993-03-15

    A recent celebration marked the twenty-fifth anniversary of the Cornell Electron Synchrotron. The major milestone in the commissioning of the synchrotron was on October 11, 1967 when Helen Edwards, Boyce McDaniel, and Maury Tigner achieved a 7 GeV beam, a worldrecord energy for electron synchrotrons at that time. Like so many advances in experimental physics, this occurred early in the morning - 3 a.m.! The transition from accelerator commissioning to high energy physics operation was extremely rapid; 7 GeV operation for data collection was routine just five weeks later. Throughout its life as a source of photon and electron beams for fixed target experiments, the synchrotron maintained energy leadership for circular electron machines. Originally designed for operation at 10 GeV, eventually it consistently provided beams for experiments at energies up to 11.6 GeV. It now operates at 5 GeV, serving as the injector for the CESR electron-positron storage ring. Robert Wilson was director of the laboratory during the design and most of the construction of the machine. He left near the end of the construction to become the first director of Fermilab and was replaced by Boyce McDaniel, who guided the laboratory from the completion of the synchrotron to the construction and early operation of CESR. Wilson recalled how the laboratory had originally proposed a 3 GeV turnkey machine to be built entirely by industry and would fit in the space previously occupied by earlier Cornell accelerators. However, members of the laboratory realized that 3 GeV would not open new physics frontiers, that the construction of the accelerator was much of the fun of doing high energy physics experiments, and that a more challenging project was needed. This led to the proposal for the 10 GeV synchrotron which was built in the ''Cornell Style'' with many of the components fabricated and nearly all of the assembly done at Cornell.

  15. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  16. Universal Nonequilibrium Signatures of Majorana Zero Modes in Quench Dynamics

    Directory of Open Access Journals (Sweden)

    R. Vasseur

    2014-10-01

    Full Text Available The quantum evolution that occurs after a metallic lead is suddenly connected to an electron system contains information about the excitation spectrum of the combined system. We exploit this type of “quantum quench” to probe the presence of Majorana fermions at the ends of a topological superconducting wire. We obtain an algebraically decaying overlap (Loschmidt echo L(t=|⟨ψ(0|ψ(t⟩|^{2}∼t^{-α} for large times after the quench, with a universal critical exponent α=1/4 that is found to be remarkably robust against details of the setup, such as interactions in the normal lead, the existence of additional lead channels, or the presence of bound levels between the lead and the superconductor. As in recent quantum-dot experiments, this exponent could be measured by optical absorption, offering a new signature of Majorana zero modes that is distinct from interferometry and tunneling spectroscopy.

  17. The Optimization of power reactor control system

    International Nuclear Information System (INIS)

    Danupoyo, S.D.

    1997-01-01

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

  18. Feasibility study of the university of Utah TRIGA reactor power upgrade - Part I: Neutronics-based study in respect to control rod system requirements and design

    Directory of Open Access Journals (Sweden)

    Ćutić Avdo

    2013-01-01

    Full Text Available We present a summary of extensive studies in determining the highest achievable power level of the current University of Utah TRIGA core configuration in respect to control rod requirements. Although the currently licensed University of Utah TRIGA power of 100 kW provides an excellent setting for a wide range of experiments, we investigate the possibility of increasing the power with the existing fuel elements and core structure. Thus, we have developed numerical models in combination with experimental procedures so as to assess the potential maximum University of Utah TRIGA power with the currently available control rod system and have created feasibility studies for assessing new core configurations that could provide higher core power levels. For the maximum determined power of a new University of Utah TRIGA core arrangement, a new control rod system was proposed.

  19. University Reactor Matching Grants Program

    International Nuclear Information System (INIS)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-01-01

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

  20. Power reactors in Member States. 1979 edition

    International Nuclear Information System (INIS)

    1979-01-01

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

  1. Computerized reactor power regulation with logarithmic controller

    International Nuclear Information System (INIS)

    Gossanyi, A.; Vegh, E.

    1982-11-01

    A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

  2. Integration of coal gasification and packed bed CLC for high efficiency and near-zero emission power generation

    NARCIS (Netherlands)

    Spallina, V.; Romano, M.C.; Chiesa, P.; Gallucci, F.; Sint Annaland, van M.; Lozza, G.

    2014-01-01

    A detailed thermodynamic analysis has been carried out of large-scale coal gasification-based power plant cycles with near zero CO2 emissions, integrated with chemical looping combustion (CLC). Syngas from coal gasification is oxidized in dynamically operated packed bed reactors (PBRs), generating a

  3. Initial performance of the Cornell cold neutron beam

    International Nuclear Information System (INIS)

    Clark, D.D.; Spern, S.A.; Atwood, A.G.

    1997-01-01

    The cold source for a guided neutron beam has been installed in a Cornell TRIGA beamport and has successfully undergone thermal tests up to full power (normally 480 kW). Tests to date (8/1/96) include spectral and yield measurements at 10 kW with the first three meters of the 2-cm by 5-cm Ni-on-glass guide in place. A 110-cm 3 Al chamber, located 17 cm from the core, contains solid mesitylene and is cooled by conduction through a 269-cm long Cu rod connected to a cryorefrigerator outside the reactor shield. Distributions of flux per unit velocity have been measured at 10 kW by time-of-flight. Anticipated properties of the complete 13 m long beam at full power are discussed. (author)

  4. Operational power reactor health physics

    International Nuclear Information System (INIS)

    Watson, B.A.

    1987-01-01

    Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

  5. Power Reactor Embrittlement Data Base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1990-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: (1) to compile and to verify the quality of the PR-EDB; (2) to provide user-friendly software to access and process the data; (3) to explore or confirm embrittlement prediction models; and (4) to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. To achieve these goals, the data base architecture was designed after much discussion and planning with prospective users, namely, material scientists and members of the research staff. The current compilation of the PR-EDB (Version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points for 110 different irradiated base materials and 161 data points for 79 different welds. Results from heat-affected zone materials are also listed. The time and effort required to process and evaluate different types of data in the PR-EDB have been drastically reduced from previous data bases. The Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of PR-EDB and will be supplementing the data base with additional data and documentation

  6. Reactor power reduction system and method

    International Nuclear Information System (INIS)

    Bruno, S.J.; Dunn, S.A.; Raber, M.

    1978-01-01

    A method of operating a nuclear power reactor is disclosed which enables an accelerated power reduction of the reactor without completely shutting the reactor down. The method includes monitoring the incidents which, upon their occurrence, would require an accelerated power reduction in order to maintain the reactor in a safe operation mode; calculating the power reduction required on the occurrence of such an incident; determining a control rod insertion sequence for the normal operation of the reactor, said sequence being chosen to optimize reactor power capability; selecting the number of control rods necessary to respond to the accelerated power reduction demand, said selection being made according to a priority determined by said control rod insertion sequence; and inserting said selected control rods into the reactor core. 11 claims, 13 figures

  7. Some properties of zero power neutron noise in a time-varying medium with delayed neutrons

    International Nuclear Information System (INIS)

    Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.

    2008-01-01

    The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed

  8. Physical protection of power reactors

    International Nuclear Information System (INIS)

    Darby, J.L.

    1979-01-01

    Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  11. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

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

  12. Nuclear power plant with several reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

    1972-05-10

    A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

  13. Nuclear reactor instrumentation power monitor

    International Nuclear Information System (INIS)

    Suzuki, Shigeru.

    1989-01-01

    The present invention concerns a nuclear reactor instrumentation power monitor that can be used in, for example, BWR type nuclear power plants. Signals from multi-channel detectors disposed on field units are converted respectively by LPRM signal circuits. Then, the converted signals are further converted by a multiplexer into digital signals and transmitted as serial data to a central monitor unit. The thus transmitted serial data are converted into parallel data in the signal processing section of the central monitor unit. Then, LPRM signals are taken out from each of channel detectors to conduct mathematical processing such as trip judgment or averaging. Accordingly, the field unit and the central monitor unit can be connected by way of only one data transmission cable thereby enabling to reduce the number of cables. Further, since the data are transmitted on digital form, it less undergoes effect of noises. (I.S.)

  14. Reactor power control method upon accidents of electrical power system

    International Nuclear Information System (INIS)

    Hirose, Masao.

    1983-01-01

    Purpose: To enable to continue the operation of a BWR type reactor by avoiding the scram while suppressing the reactor power, just after the external disturbance such as earth-trouble in power-transmission network. Method: Steep power drop of an electrical generator is to be detected not only by a current-type power-load-unbalance relay but also with a power-type power-load-unbalance-relay. If steep power-drop was detected by the latter relay, a previously selected control rod is rapidly inserted into the reactor. In this way, in the case where there is a possibility of the reactor scram, the scram can be avoided by suppressing the reactor power, thus the reactor operation can be continued. (Kamimura, M.)

  15. BN-1200 Reactor Power Unit Design Development

    International Nuclear Information System (INIS)

    Vasilyev, B.A.; Shepelev, S.F.; Ashirmetov, M.R.; Poplavsky, V.M.

    2013-01-01

    Main goals of BN-1200 design: • Develop a reliable new generation reactor plant for the commercial power unit with fast reactor to implement the first-priority objectives in changing over to closed nuclear fuel cycle; • Improve technical and economic indices of BN reactor power unit to the level of those of Russian VVER of equal power; • Enhance the safety up to the level of the requirements for the 4th generation RP

  16. Mathematical game type optimization of powerful fast reactors

    International Nuclear Information System (INIS)

    Pavelesku, M.; Dumitresku, Kh.; Adam, S.

    1975-01-01

    To obtain maximum speed of putting into operation fast breeders it is recommended on the initial stage of putting into operation these reactors to apply lower power which needs less fission materials. That is why there is an attempt to find a configuration of a high-power reactor providing maximum power for minimum mass of fission material. This problem has a structure of the mathematical game with two partners of non-zero-order total and is solved by means of specific aids of theory of games. Optimal distribution of fission and breeding materials in a multizone reactor first is determined by solution of competitive game and then, on its base, by solution of the cooperation game. The second problem the solution for which is searched is developed from remark on the fact that a reactor with minimum coefficient of flux heterogenity has a configuration different from the reactor with power coefficient heterogenity. Maximum burn-up of fuel needs minimum heterogenity of the flux coefficient and the highest power level needs minimum coefficient of power heterogenity. That is why it is possible to put a problem of finding of the reactor configuration having both coefficients with minimum value. This problem has a structure of a mathematical game with two partners of non-zero-order total and is solved analogously giving optimal distribution of fuel from the new point of view. In the report is shown that both these solutions are independent which is a result of the aim put in the problem of optimization. (author)

  17. Further analysis of the zero-energy experiment on the Dragon reactor

    International Nuclear Information System (INIS)

    Woloch, F.; Neuberger, W.

    1978-01-01

    The analysis of the Zero-Energy Experiments performed on the Dragon reactor, a high-temperature reactor of the Organization for Economic Cooperation and Development, has been continued. The first analysis established the main route of calculations within the WIMS-E scheme and was reported elsewhere. This Note presents further calculations showing the merits of a refinement in the number of neutron energy groups, of the use of different condensation spectra, and of transport calculations

  18. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

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

  19. Power reactor embrittlement data base

    International Nuclear Information System (INIS)

    Kam, F.B.K.; Stallmann, F.W.; Wang, J.A.

    1989-01-01

    Regulatory and research evaluations of embrittlement prediction models and of vessel integrity under load can be greatly expedited by the use of a well-designed, computerized embrittlement data base. The Power Reactor Embrittlement Data Base (PR-EDB) is a comprehensive collection of data from surveillance reports and other published reports of commercial nuclear reactors. The uses of the data base require that as many different data as available are collected from as many sources as possible with complete references and that subsets of relevant data can be easily retrieved and processed. The objectives of this NRC-sponsored program are the following: to compile and to verify the quality of the PR-EDB; to provide user-friendly software to access and process the data; to explore or confirm embrittlement prediction models; and to interact with standards organizations to provide the technical bases for voluntary consensus standards that can be used in regulatory guides, standard review plans, and codes. 9 figs

  20. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    Gillies, C.; White, M.

    1995-11-01

    A review of Canadian and worldwide performance indicator definitions and data was performed to identify a set of indicators that could be used for comparison of performance among nuclear power plants. The results of this review are to be used as input to an AECB team developing a consistent set of performance indicators for measuring Canadian power reactor safety performance. To support the identification of performance indicators, a set of criteria was developed to assess the effectiveness of each indicator for meaningful comparison of performance information. The project identified a recommended set of performance indicators that could be used by AECB staff to compare the performance of Canadian nuclear power plants among themselves, and with international performance. The basis for selection of the recommended set and exclusion of others is provided. This report provides definitions and calculation methods for each recommended performance indicator. In addition, a spreadsheet has been developed for comparison and trending for the recommended set of indicators. Example trend graphs are included to demonstrate the use of the spreadsheet. (author). 50 refs., 11 tabs., 3 figs

  1. Cornell Mixing Zone Expert System

    Science.gov (United States)

    This page provides an overview Cornell Mixing Zone Expert System water quality modeling and decision support system designed for environmental impact assessment of mixing zones resulting from wastewater discharge from point sources

  2. Operational safety and reactor life improvements of Kyoto University Reactor

    International Nuclear Information System (INIS)

    Utsuro, M.; Fujita, Y.; Nishihara, H.

    1990-01-01

    Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

  3. Transmutation of actinides in power reactors.

    Science.gov (United States)

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides.

  4. Power controlling method for BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1983-01-01

    Purpose: To enable reactor operation exactly following after an aimed curve in the high power resuming and maintaining period without failures in cladding tubes. Method: Upon recovery of the reactor power to a high power level after changing the reactor power from the high power to the low power level, control rod is operated under such conditions that the linear power density after operation of the control rod does not exceed the PC envelope in the low power period, and the core flow rate is coordinated to the control rod operation. The linear power density can be suppressed within an allowable linear power density by the above operation during high power resuming and maintaining period and, as the result, PCI failures can be prevented. (Kamimura, M.)

  5. A WIMS E analysis of zero energy experiments performed on the Dragon reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lancefield, M. J.; Broadhouse, B.; Woloch, F.

    1974-10-15

    UKAEA methods embodied in the WINS-E modular scheme of codes are described in their application to the analysis of zero energy experiments performed on the DRAGON reactor. Measured reactivity and reaction rate distributions are compared with the predictions of the analysis.

  6. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shwageraus, E.; Fridman, E. [Department of Nuclear Engineering, Ben-Gurion University of the Negev Beer-Sheva 84105 (Israel)

    2008-07-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO{sub 2} fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO{sub 2} LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  7. Decay Power Calculation for Safety Analysis of Innovative Reactor Systems

    International Nuclear Information System (INIS)

    Shwageraus, E.; Fridman, E.

    2008-01-01

    In this work, we verified the decay heat calculation capabilities of BGCore computer code system developed recently at Ben-Gurion University. Decay power was calculated for a typical UO 2 fuel in Pressurized Water Reactor environment using BGCore code and using procedure prescribed by the ANS/ANSI-2005 standard. Very good agreement between the two methods was obtained. Once BGCore calculation capabilities were verified, we calculated decay power as a function of time after shutdown for various reactors with innovative fuels, for which no standard procedure is currently available. Notable differences were observed for decay power of the advanced reactors as compared with conventional UO 2 LWR. The observed differences suggest that the design of new reactors safety systems must be based on corresponding decay power curves for each individual case in order to assure the desired performance of such systems. (authors)

  8. Power generator in BWR type reactors

    International Nuclear Information System (INIS)

    Yoshida, Kenji.

    1984-01-01

    Purpose: To enable to perform stable and dynamic conditioning operation for nuclear fuels in BWR type reactors. Constitution: The conditioning operation for the nuclear fuels is performed by varying the reactor core thermal power in a predetermined pattern by changing the predetermined power changing pattern of generator power, the rising rate of the reactor core thermal power and the upper limit for the rising power of the reactor core thermal power are calculated and the power pattern for the generator is corrected by a power conditioning device such that the upper limit for the thermal power rising rate and the upper limit for the thermal power rising rate are at the predetermined levels. Thus, when the relation between the reactor core thermal power and the generator electrical power is fluctuated, the fluctuation is detected based on the variation in the thermal power rising rate and the limit value for the thermal power rising rate, and the correction is made to the generator power changing pattern so that these values take the predetermined values to thereby perform the stable conditioning operation for the nuclear fuels. (Moriyama, K.)

  9. Impacts on power reactor health physics programs

    International Nuclear Information System (INIS)

    Meyer, B.A.

    1991-01-01

    The impacts on power reactor health physics programs form implementing the revised 10 CFR Part 20 will be extensive and costly. Every policy, program, procedure and training lesson plan involving health physics will require changes and the subsequent retraining of personnel. At each power reactor facility, hundreds of procedures and thousands of people will be affected by these changes. Every area of a power reactor health physics program will be affected. These areas include; ALARA, Respiratory Protection, Exposure Control, Job Coverage, Dosimetry, Radwaste, Effluent Accountability, Emergency Planning and Radiation Worker Training. This paper presents how power reactor facilities will go about making these changes and gives possible examples of some of these changes and their impact on each area of power reactor health physics program

  10. Investigation of zero-release cycle using fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The task force was organized for the main purpose of offering quantitative basic data to the study group on nuclear fuel cycle in February, 1997. The effect of so-called frontier technologies such as the isotope separation by laser method, the FP annihilation with electron beam accelerators and so on in the FBR cycle based on MOX fuel and PUREX reprocessing method was expected. It is aimed at to recycle the total amount of minor actinides. The object of recycling is the nuclides which contribute largely to toxicity, namely 11 elements, 12 nuclides. The preconditions and the target to be attained of the investigation are explained. As the results of investigation, the amount of reloading MA and FP into a reactor, squeezing the recycling scenario, the effect of reducing toxicity and the subject of the countermeasures to the nuclides with long half-life which cannot be reloaded are reported. As the technical evaluation required for realizing the concept, the concept of the core which excludes recriticality, the advance of reprocessing technology, isotope separation, the fabrication into the optimal form for recycling and so on are discussed. The economical efficiency of the recycling based on MOX and PUREX and the proposal of the development scenario are described. (K.I.)

  11. A unified model of quarks and leptons with a universal texture zero

    Science.gov (United States)

    de Medeiros Varzielas, Ivo; Ross, Graham G.; Talbert, Jim

    2018-03-01

    We show that a universal texture zero in the (1,1) position of all fermionic mass matrices, including heavy right-handed Majorana neutrinos driving a type-I see-saw mechanism, can lead to a viable spectrum of mass, mixing and CP violation for both quarks and leptons, including (but not limited to) three important postdictions: the Cabibbo angle, the charged lepton masses, and the leptonic `reactor' angle. We model this texture zero with a non-Abelian discrete family symmetry that can easily be embedded in a grand unified framework, and discuss the details of the phenomenology after electroweak and family symmetry breaking. We provide an explicit numerical fit to the available data and obtain excellent agreement with the 18 observables in the charged fermion and neutrino sectors with just 9 free parameters. We further show that the vacua of our new scalar familon fields are readily aligned along desired directions in family space, and also demonstrate discrete gauge anomaly freedom at the relevant scale of our effective theory.

  12. Fast reactors: potential for power

    International Nuclear Information System (INIS)

    1983-02-01

    The subject is discussed as follows: basic facts about conventional and fast reactors; uranium economy; plutonium and fast reactors; cooling systems; sodium coolant; safety engineering; handling and recycling plutonium; safeguards; development of fast reactors in Britain and abroad; future progress. (U.K.)

  13. Thermionic reactors for space nuclear power

    Science.gov (United States)

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

    1985-01-01

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

  14. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

  15. Method and device for controlling reactor power

    International Nuclear Information System (INIS)

    Oohashi, Masahisa; Masuda, Hiroyuki.

    1982-01-01

    Purpose: To enable load following-up operation of a reactor adapted to perform power conditioning by the control of the liquid poison density in the core and by the control rods. Constitution: In a case where the reactor power is repeatedly changed in a reactor having a liquid poison density control device and control rods, the time period for the power control is divided depending on the magnitude of the change with time in the reactivity and the optimum values are set for the injection and removal amount of the liquid poison within the divided period. Then, most parts of the control required for the power change are alloted to the liquid poison that gives no effect on the power distribution while minimizing the movement of the control rods, whereby the power change in the reactor as in the case of the load following-up operation can be practiced with ease. (Kawakami, Y.)

  16. The Texts of the Instruments relating to a Project for a Joint Agency-Norwegian Program of Research with the Zero Power Reactor 'NORA'. Modification of the Supply Agreement

    International Nuclear Information System (INIS)

    1961-01-01

    Pursuant to Section 15 of the 'Contract for the Lease of Enriched Uranium' for the NORA reactor, the United States Atomic Energy Commission has reduced, with effect from 1 January 1962, the rates of the Use Charge specified in Section II(a) and of the Consumption Charge specified in Section II(b)

  17. Nuclear power reactors of new generation

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

    1988-01-01

    The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

  18. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  19. Small and medium power reactors 1987

    International Nuclear Information System (INIS)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347 Small and Medium Power Reactors Project Initiation Study - Phase I published in 1985 and TECDOC-376 Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power programme. It consists of two parts: 1) Guidelines for the Introduction of Small and Medium Power Reactors in Developing Countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of Small and Medium Power Reactors in developing countries; 2) Up-dated Information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex I of the above mentioned TECDOC-347. Figs

  20. Small and medium power reactors 1987

    Science.gov (United States)

    1987-12-01

    This TECDOC follows the publication of TECDOC-347: Small and Medium Power Reactors (SMPR) Project Initiation Study, Phase 1, published in 1985 and TECDOC-376: Small and Medium Power Reactors 1985 published in 1986. It is mainly intended for decision makers in Developing Member States interested in embarking on a nuclear power program. It consists of two parts: (1) guidelines for the introduction of small and medium power reactors in developing countries. These Guidelines were established during the Advisory Group Meeting held in Vienna from 11 to 15 May 1987. Their purpose is to review key aspects relating to the introduction of small and medium power reactors in developing countries; (2) up-dated information on SMPR Concepts Contributed by Supplier Industries. According to the recommendations of the Second Technical Committee Meeting on SMPRs held in Vienna in March 1985, this part contains the up-dated information formerly published in Annex 1 of the above mentioned TECDOC-347.

  1. Radiation protection in a university TRIGA reactor

    International Nuclear Information System (INIS)

    Tschurlovits, M. . Author

    2004-01-01

    Radiation protection in a university institute operating a research reactor and other installations has different constraints as a larger facility. This is because the legal requirements apply in full, but the potential of exposure is low, and accesses has to be made available for students, but also for temporary workers. Some of the problems in practical radiation protection are addressed and solutions are discussed. In addition, experience with national radiation protection legislation recently to be issued is addressed and discussed. (author)

  2. Multiple microprocessor based nuclear reactor power monitor

    International Nuclear Information System (INIS)

    Lewis, P.S.; Ethridge, C.D.

    1979-01-01

    The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

  3. Power distribution forecasting device for reactors

    International Nuclear Information System (INIS)

    Tsukii, Makoto

    1981-01-01

    Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

  4. Power distribution monitor in a nuclear reactor

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi

    1983-01-01

    Purpose: To enable accurate monitoring for the reactor power distribution within a short time in a case where abnormality occurs in in-core neutron monitors or in a case where the reactor core state changes after the calibration for the neutron monitors. Constitution: The power distribution monitor comprises a power distribution calculator adapted to be inputted counted values from a reactor core present state data instruments and calculate the neutron flux distribution in the reactor core and the power distribution based on previously incorporated physical models, an RCF calculator adapted to be inputted with the counted values from the in-core neutron monitors and the neutron flux distribution and the power distribution calculated in the power distribution calculator and compensate the counted errors included in the counted values form the in-core neutron monitors and the calculation errors included in the power distribution calculated in the power distribution calculator to thereby calculate the power distribution within the reactor core, and an input/output device for the input of the data required for said power distribution calculator and the display for the calculation result calculated in the RCF calculator. (Ikeda, J.)

  5. Power conditioning system for a nuclear reactor

    International Nuclear Information System (INIS)

    Higashigawa, Yuichi; Joge, Toshio.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Hvinden, T.

    1976-01-01

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

  7. Identification of fast power reactivity effect in nuclear power reactor

    International Nuclear Information System (INIS)

    Efanov, A.I.; Kaminskas, V.A.; Lavrukhin, V.S.; Rimidis, A.P.; Yanitskene, D.Yu.

    1987-01-01

    A nuclear power reactor is an object of control with distributed parameters, characteristics of which vary during operation time. At the same time the reactor as the object of control has internal feedback circuits, which are formed as a result of the effects of fuel parameters and a coolant (pressure, temperature, steam content) on the reactor breeding properties. The problem of internal feedback circuit identification in a nuclear power reactor is considered. Conditions for a point reactor identification are obtained and algorithms of parametric identification are constructed. Examples of identification of fast power reactivity effect for the RBMK-1000 reactor are given. Results of experimental testing have shown that the developed method of fast power reactivity effect identification permits according to the data of normal operation to construct adaptive models for the point nuclear reactor, designed for its behaviour prediction in stationary and transition operational conditions. Therefore, the models considered can be used for creating control systems of nuclear power reactor thermal capacity (of RBMK type reactor, in particular) which can be adapted to the change in the internal feedback circuit characteristics

  8. MODERATOR ELEMENTS FOR UNIFORM POWER NUCLEAR REACTOR

    Science.gov (United States)

    Balent, R.

    1963-03-12

    This patent describes a method of obtaining a flatter flux and more uniform power generation across the core of a nuclear reactor. The method comprises using moderator elements having differing moderating strength. The elements have an increasing amount of the better moderating material as a function of radial and/or axial distance from the reactor core center. (AEC)

  9. A nuclear power reactor concept for Brazil

    International Nuclear Information System (INIS)

    Sefidvash, F.

    1980-01-01

    For the purpose of developing an independent national nuclear technology and effective manner of transferring such a technology, as well as developing a modern reactor, a new nuclear power reactor concept is proposed which is considered as a suitable and viable project for Brazil to support its development and finally construct its prototype as an indigeneous venture. (Author) [pt

  10. SEISMIC DESIGN CRITERIA FOR NUCLEAR POWER REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R. A.

    1963-10-15

    The nature of nuclear power reactors demands an exceptionally high degree of seismic integrity. Considerations involved in defining earthquake resistance requirements are discussed. Examples of seismic design criteria and applications of the spectrum technique are described. (auth)

  11. Reactor power control method and device

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

    1997-01-01

    The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

  12. Low power reactor for remote applications

    International Nuclear Information System (INIS)

    Meier, K.L.; Palmer, R.G.; Kirchner, W.L.

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical output of 25 kWe from an organic Rankine cycle (ORC) engine. By intentional selection of design features stressing inherent safety, operation in an unattended mode is possible with minimal risk to the environment. Reliability is achieved through the use of components representing existing, proven technology. Low enrichment uranium particle fuel, in graphite core blocks, cooled by heat pipes coupled to an ORC converter insures long-term, virtually maintenance free, operation of this reactor for remote applications. 10 refs., 7 figs., 3 tabs

  13. Neutron energy spectra calculations in the low power research reactor

    International Nuclear Information System (INIS)

    Omar, H.; Khattab, K.; Ghazi, N.

    2011-01-01

    The neutron energy spectra have been calculated in the fuel region, inner and outer irradiation sites of the zero power research reactor using the MCNP-4C code and the combination of the WIMS-D/4 transport code for generation of group constants and the three-dimensional CITATION diffusion code for core analysis calculations. The neutron energy spectrum has been divided into three regions and compared with the proposed empirical correlations. The calculated thermal and fast neutron fluxes in the low power research reactor MNSR inner and outer irradiation sites have been compared with the measured results. Better agreements have been noticed between the calculated and measured results using the MCNP code than those obtained by the CITATION code. (author)

  14. WWER-440 control assembly local power peaking investigation on LR-0 reactor

    International Nuclear Information System (INIS)

    Mikus, J.

    2002-01-01

    This paper presents information concerning the local power peaking problem induced by the WWER-440 control assembly and the investigation possibilities on the light water, zero power reactor LR-0 at the Nuclear Research Institute (NRI) Rez plc. A brief description is given about the disposable control assembly model, experimental arrangement and conditions on the LR-0 reactor with regard to the earlier performed investigations as well as to the relevant measurements to be realized in the near future.(abstract)

  15. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  16. reactor power control using fuzzy logic

    International Nuclear Information System (INIS)

    Ahmed, A.E.E.

    2001-01-01

    power stabilization is a critical issue in nuclear reactors. convention pd- controller is currently used in egypt second testing research reactor (ETRR-2). two fuzzy controllers are proposed to control the reactor power of ETRR-2 reactor. the design of the first one is based on a set of linguistic rules that were adopted from the human operators experience. after off-line fuzzy computations, the controller is a lookup table, and thus, real time controller is achieved. comparing this f lc response with the pd-controller response, which already exists in the system, through studying the expected transients during the normal operation of ETRR-2 reactor, the simulation results show that, fl s has the better response, the second controller is adaptive fuzzy controller, which is proposed to deal with system non-linearity . The simulation results show that the proposed adaptive fuzzy controller gives a better integral square error (i se) index than the existing conventional od controller

  17. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  18. Measuring the productivity of university research reactors

    International Nuclear Information System (INIS)

    Voth, M.H.

    1989-01-01

    University Research Reactors (URRs) on 33 campuses in the United States provide valuable contributions to academic instruction and research programs. In most cases, there are no alternative diagnostic techniques to supplant the need for a reactor and associated facilities. Since URRs constitute a major financial commitment, it is important that they be operated in a productive manner. Productivity may be defined as the sum of new knowledge generated, existing knowledge transferred to others, and analytical services provided to assist in the generation of new knowledge; another definition of productivity is this sum expressed as a function of the cost incurred. In either case, a consistent measurement is difficult and more qualitative than quantitative. A uniform reporting system has been proposed that defines simplified categories through which meaningful comparisons can be performed

  19. University Reactor Sharing Program. Final report, September 30, 1992--September 29, 1994

    International Nuclear Information System (INIS)

    Wehring, B.W.

    1995-01-01

    Over the past 20 years, the number of nuclear reactors on university campuses in the US declined from more than 70 to less than 40. Contrary to this trend, The University of Texas at Austin constructed a new reactor facility at a cost of $5.8 million. The new reactor facility houses a new TRIGA Mark II reactor which replaces an in-ground TRIGA Mark I reactor located in a 50-year old building. The new reactor facility was constructed to strengthen the instruction and research opportunities in nuclear science and engineering for both undergraduate and graduate students at The University of Texas. On January 17, 1992, The University of Texas at Austin received a license for operation of the new reactor. Initial criticality was achieved on March 12, 1992, and full power operation, on March 25, 1992. The UT-TRIGA research reactor provides hands-on education, multidisciplinary research and unique service activities for academic, medical, industrial, and government groups. Support by the University Reactor Sharing Programs increases the availability of The University of Texas reactor facility for use by other educational institutions which do not have nuclear reactors

  20. New generation of reactors for space power

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Buden, D.

    1982-01-01

    Space nuclear reactor power is expected to enable many new space missions that will require several times to several orders of magnitude anything flown in space to date. Power in the 100-kW range may be required in high earth orbit spacecraft and planetary exploration. The technology for this power system range is under development for the Department of Energy with the Los Alamos National Laboratory responsible for the critical components in the nuclear subsystem. The baseline design for this particular nuclear sybsystem technology is described in this paper; additionally, reactor technology is reviewed from previous space power programs, a preliminary assessment is made of technology candidates covering an extended power spectrum, and the status is given of other reactor technologies

  1. Compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine Cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for component development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analysis have aided in the power source design. The analyses have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high. 10 refs

  2. A compact reactor/ORC power source

    International Nuclear Information System (INIS)

    Meier, K.L.; Kirchner, W.L.; Willcutt, G.J.

    1986-01-01

    A compact power source that combines an organic Rankine cycle (ORC) electric generator with a nuclear reactor heat source is being designed and fabricated. Incorporating existing ORC technology with proven reactor technology, the compact reactor/ORC power source offers high reliability while minimizing the need for componenet development. Thermal power at 125 kWt is removed from the coated particle fueled, graphite moderated reactor by heat pipes operating at 500 0 C. Outside the reactor vessel and connected to the heat pipes are vaporizers in which the toluene ORC working fluid is heated to 370 0 C. In the turbine-alternator-pump (TAP) combined-rotating unit, the thermal energy of the toluene is converted to 25 kWe of electric power. Lumped parameter systems analyses combined with a finite element thermal analyses combined with a finite element thermal analysis have aided in the power source design. The analysis have provided assurance of reliable multiyear normal operation as well as full power operation with upset conditions, such as failed heat pipes and inoperative ORC vaporizers. Because of inherent high reliability, long life, and insensitivity to upset conditions, this power source is especially suited for use in remote, inaccessible locations where fuel delivery and maintenance costs are high

  3. A university contribution to reactor safety

    International Nuclear Information System (INIS)

    Hall, W.B.

    1980-01-01

    The total UK university effort available for research specifically directed towards reactor safety is certainly small in comparison with that in industry. To be worth while, the work should complement that in the industry, and ways in which this can, and in some cases does, happen, will be discussed. There is, however, another reason for university involvement: the need for an informed body of opinion on matters of reactor safety outside the industry. Without this it is difficult for the public and its representatives to assure themselves that the depth and scope of safety analysis is commensurate with the seriousness of the problem, and that the best available data and techniques are being used. An independent inspectorate is an essential element in this philosophy, but in addition there is much to be said for exposing the arguments to scrutiny by the widest possible range of informed critics. Such people will be much more effective if they are themselves involved in real problems in the field. In a university, this involvement is probably best achieved through research; as mentioned above, the type of research should preferably complement that being carried out in the industry. The current situation, and the future, are discussed. (author)

  4. Thorium utilization in power reactors

    International Nuclear Information System (INIS)

    Saraceno; Marcos.

    1978-10-01

    In this work the recent (prior to Aug, 1976) literature on thorium utilization is reviewed briefly and the available information is updated. After reviewing the nuclear properties relevant to the thorium fuel cycle we describe briefly the reactor systems that have been proposed using thorium as a fertile material. (author) [es

  5. Bottom reflector for power reactors

    International Nuclear Information System (INIS)

    Elter, C.; Kissel, K.F.; Schoening, J.; Schwiers, H.G.

    1982-01-01

    In pebble bed reactors erosion and damage due fuel elements movement on the surface of the bottom reflector should be minimized. This can be achieved by chamfering and/or rounding the cover edges of the graphite blocks and the edges between the drilled holes and the surface of the graphite block. (orig.) [de

  6. Power supply with nuclear reactor

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

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

  7. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  8. Parliament votes against building fifth power reactor

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    After a heated three-day debate, Finland's parliament voted on September 24 to reject the proposal to build the country's fifth nuclear power reactor. As predicted, the vote was close: 107 voted against more nuclear power, 90 were in favor, two members of the 200-seat parliament were not present, and the speaker did not vote

  9. Nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-06-01

    The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Lifetimes of 7 to 10 yr at full power, at converter operating temperatures of 1275 to 1675 0 K, are being studied. The systems are being designed such that no single-failure modes exist that will cause a complete loss of power. In fact, to meet the long lifetimes, highly redundant design features are being emphasized. Questions have been raised about safety since the COSMOS 954 incident. ''Fail-safe'' means to prevent exposure of the population to radioactive material, meeting the environmental guidelines established by the U.S. Government have been and continue to be a necessary requirement for any space reactor program. The major safety feature to prevent prelaunch and launch radioactive material hazards is not operating the reactor before achieving the prescribed orbit. Design features in the reactor ensure that accidental criticality cannot occur. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit, where orbital lifetimes are practically indefinite, the safety considerations are negligible. Orbits below 400 to 500 nautical miles are the ones where a safety issue is involved in case of satellite malfunction. The potential missions, the question of why reactors are being considered as a prime power candidate, reactor features, and safety considerations will be discussed

  10. TerraPower, Bill Gates' reactor

    International Nuclear Information System (INIS)

    Guidez, J.

    2016-01-01

    TerraPower is a traveling wave reactor, it means that the reactor gradually converts non fissile material into the fuel it needs and the active part of the core progressively moves through the core leaving spent fuel behind. The last design of the TerraPower shows that it will use depleted uranium as fuel and that its core will need reloading every 10 years. Re-arrangement of the nuclear fuel will have to be made every 18 months to keep the core reactive. Metallic nuclear fuels will be used as they allow the highest breeding rates. It appears that apart from the very specific configuration of the core, the TerraPower is a reactor very similar to sodium-cooled fast reactors. Neutron transport inside traveling wave reactor core is complex and simulations show that the piling-up of fission product tends to kill the chain reaction and a continuous neutron addition may be necessary to keep the reactor going. A large part of the TerraPower feasibility studies concerns neutron transport inside its core. (A.C.)

  11. Education for university students, high school teachers and the general public using the Kinki University Reactor

    International Nuclear Information System (INIS)

    Tsuruta, T.

    2007-01-01

    Atomic Energy Research Institute of Kinki University is equipped with a nuclear reactor which is called UTR-KINKI. UTR is the abbreviation for University Teaching and Research Reactor. The reactor is the first one installed in Japanese universities. Though the reactor is owned and operated by Kinki University, its use is widely open to scientists and students from other universities and research institutions. The reactor is made the best of teaching instrument for the training of high school teachers. In addition, the reactor is utilized for general public education concerning atomic energy. (author)

  12. Gaseous fuel reactors for power systems

    Science.gov (United States)

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

    1977-01-01

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

  13. DOE University Reactor Sharing Program. Final technical report for 1996--1997

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1998-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor and radiation facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Radiation Effects Laboratory, and the UMCP 2 x 4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments. The Maryland University Training Reactor (MUTR) and the associated laboratories are made available to any interested institution six days a week on a scheduled basis. Most institutions are scheduled at the time of their first request--a reflection of their commitment to the Reactor Sharing Program. The success of the past years by no means guarantees future success. Therefore, the reactor staff is more aggressively pursuing its outreach program, especially with junior colleges and universities without reactor or radiation facilities; more aggressively developing demonstration and training programs for students interested in careers in nuclear power and radiation technology; and more aggressively up-grading the reactor facilities--not only to provide a better training facility but to prepare for relicensing in the year 2000

  14. Power control device in nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To enable smooth power changes in power conditioning systems by calculating forecast values for the neutron flux distribution and power distribution and by controlling the driving speed of control rods so as to correspond the forecast values with aimed values. Constitution: Control rod position is detected by a position detector and sent to a control computer as the position information. At the same time, the neutron flux distribution information is obtained by the neutron monitors, the power distribution information is obtained by a reactor power computer and they are outputted to the control computer. The control computer calculates the forecast values for the neutron flux distribution and the reactor power distribution from the information, and compares them with the aimed values from a setter and then outputs control signals so as to correspond the forecast values with the aimed values. The control rods can be inserted in appropriate velocity by the control signals. (Horiuchi, T.)

  15. Enhancement of aerobic granulation by zero-valent iron in sequencing batch airlift reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kong, Qiang, E-mail: kongqiang0531@hotmail.com [College of Life Science, Shandong Normal University, 88 Wenhua Donglu, Jinan 250014, Shandong (China); Ngo, Huu Hao [School of Civil and Environmental Engineering, University of Technology Sydney, Broadway, NSW 2007 (Australia); Shu, Li [School of Engineering, Faculty of Science, Engineering and Built Environment, Deakin University, Geelong, Victoria 3216 (Australia); Fu, Rong-shu; Jiang, Chun-hui [College of Life Science, Shandong Normal University, 88 Wenhua Donglu, Jinan 250014, Shandong (China); Miao, Ming-sheng, E-mail: mingshengmiao@163.com [College of Life Science, Shandong Normal University, 88 Wenhua Donglu, Jinan 250014, Shandong (China)

    2014-08-30

    Highlights: • Zero-valent iron (ZVI) was used firstly to enhance the aerobic granulation. • ZVI significantly decreased the start-up time of the aerobic granulation. • ZVI had the function of enhancing organic material diversity identified by 3-D EEM. • ZVI could enhance the diversity of microbial community. - Abstract: This study elucidates the enhancement of aerobic granulation by zero-valent iron (ZVI). A reactor augmented with ZVI had a start-up time of aerobic granulation (43 days) that was notably less than that for a reactor without augmentation (64 days). The former reactor also had better removal efficiencies for chemical oxygen demand and ammonium. Moreover, the mature granules augmented with ZVI had better physical characteristics and produced more extracellular polymeric substances (especially of protein). Three-dimensional-excitation emission matrix fluorescence showed that ZVI enhanced organic material diversity. Additionally, ZVI enhanced the diversity of the microbial community. Fe{sup 2+} dissolution from ZVI helped reduce the start-up time of aerobic granulation and increased the extracellular polymeric substance content. Conclusively, the use of ZVI effectively enhanced aerobic granulation.

  16. Enhancement of aerobic granulation by zero-valent iron in sequencing batch airlift reactor

    International Nuclear Information System (INIS)

    Kong, Qiang; Ngo, Huu Hao; Shu, Li; Fu, Rong-shu; Jiang, Chun-hui; Miao, Ming-sheng

    2014-01-01

    Highlights: • Zero-valent iron (ZVI) was used firstly to enhance the aerobic granulation. • ZVI significantly decreased the start-up time of the aerobic granulation. • ZVI had the function of enhancing organic material diversity identified by 3-D EEM. • ZVI could enhance the diversity of microbial community. - Abstract: This study elucidates the enhancement of aerobic granulation by zero-valent iron (ZVI). A reactor augmented with ZVI had a start-up time of aerobic granulation (43 days) that was notably less than that for a reactor without augmentation (64 days). The former reactor also had better removal efficiencies for chemical oxygen demand and ammonium. Moreover, the mature granules augmented with ZVI had better physical characteristics and produced more extracellular polymeric substances (especially of protein). Three-dimensional-excitation emission matrix fluorescence showed that ZVI enhanced organic material diversity. Additionally, ZVI enhanced the diversity of the microbial community. Fe 2+ dissolution from ZVI helped reduce the start-up time of aerobic granulation and increased the extracellular polymeric substance content. Conclusively, the use of ZVI effectively enhanced aerobic granulation

  17. Advances in ICF power reactor design

    International Nuclear Information System (INIS)

    Hogan, W.J.; Kulcinski, G.L.

    1985-01-01

    Fifteen ICF power reactor design studies published since 1980 are reviewed to illuminate the design trends they represent. There is a clear, continuing trend toward making ICF reactors inherently safer and environmentally benign. Since this trend accentuates inherent advantages of ICF reactors, we expect it to be further emphasized in the future. An emphasis on economic competitiveness appears to be a somewhat newer trend. Lower cost of electricity, smaller initial size (and capital cost), and more affordable development paths are three of the issues being addressed with new studies

  18. Containment and surveillance techniques at power reactors

    International Nuclear Information System (INIS)

    Stirling, A.J.

    1982-01-01

    This session will provide participants with an understanding of the functions of safeguards equipment at power reactors, including equipment for fuel accounting, video and film surveillance, diversion monitoring, and containment and surveillance of irradiated fuel in storage. In addition, some appreciation of the impact that reactor safeguards have on the plant operator will be gained. From this, participants will be able to ensure that a reactor safeguards system meets their nation's international and national nonproliferation objectives with a minimum of interference to plant operations

  19. Different types of power reactors and provenness

    International Nuclear Information System (INIS)

    Goodman, E.I.

    1977-01-01

    The lecture guides the potential buyer in the selection of a reactor type. Recommended criteria regarding provenness, licensability, and contractual arrangements are defined and discussed. Tabular data summarizing operating experience and commercial availability of units are presented and discussed. The status of small and medium power reactors which are of interest to many developing countries is presented. It is stressed that each prospective buyer will have to establish his own criteria based on specific conditions which will be applied to reactor selection. In all cases it will be found that selection, either pre-selection of bidders or final selection of supplier, will be a fairly complex evaluation. (orig.) [de

  20. Safety of next generation power reactors

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

  1. The Cornell electron-positron storage ring - CESR

    International Nuclear Information System (INIS)

    DeWire, J.W.

    1977-01-01

    At the Laboratory of Nuclear Studies of Cornell University we are working on a project to convert the present 12 GeV electron synchrotron complex into the Cornell Electron-Positron Storage Ring - CESR. The design studies for this new device were begun in early 1975. During the past eighteen months the National Science Foundation has supported a program of research and development on CESR and funds to begin construction are included in the NSF budget now before the U.S. Congress. Our goal is to have CESR in operation in the fall of 1979. (orig.) [de

  2. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Buden, D.; Ranken, W.A.; Koenig, D.R.

    1980-01-01

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  3. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Malkoske, G.R.; Norton, J.L.; Slack, J.

    2002-01-01

    MDS Nordion has been supplying cobalt-60 sources to industry for industrial and medical purposes since 1946. These cobalt-60 sources are used in many market and product segments, but are primarily used to sterilize single-use medical products including; surgical kits, gloves, gowns, drapes, and cotton swabs. Other applications include sanitization of cosmetics, microbial reduction of pharmaceutical raw materials, and food irradiation. The technology for producing the cobalt-60 isotope was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) almost 55 years ago using research reactors at the AECL Chalk River Laboratories in Ontario, Canada. The first cobalt-60 source produced for medical applications was manufactured by MDS Nordion and used in cancer therapy. The benefits of cobalt-60 as applied to medical product manufacturing, were quickly realized and the demand for this radioisotope quickly grew. The same technology for producing cobalt-60 in research reactors was then designed and packaged such that it could be conveniently transferred to a utility/power reactor. In the early 1970's, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production for industrial irradiation applications was initiated in the four Pickering A CANDU reactors. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology for producing cobalt-60 in additional CANDU reactors. CANDU is unique among the power reactors of the world, being heavy water moderated and fuelled with natural uranium. They are also designed and supplied with stainless steel adjusters, the primary function of which is to shape the neutron flux to optimize reactor power and fuel bum-up, and to provide excess reactivity needed to overcome xenon-135 poisoning following a reduction of power. The reactor is designed to develop full power output with all of the adjuster

  4. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2009-01-01

    The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view

  5. Invasion Ecology. Student Edition. Cornell Scientific Inquiry Series.

    Science.gov (United States)

    Krasny, Marianne E.; Trautmann, Nancy; Carlsen, William; Cunningham, Christine

    This book contains the student edition of the Environmental Inquiry curriculum series developed at Cornell University. It is designed to teach learning skills for investigating the behaviors of non-native and native species and demonstrate how to apply scientific knowledge to solve real-life problems. This book focuses on strange intruders…

  6. Deflecting cavity for beam diagnostics at Cornell ERL injector

    International Nuclear Information System (INIS)

    Belomestnykh, Sergey; Bazarov, Ivan; Shemelin, Valery; Sikora, John; Smolenski, Karl; Veshcherevich, Vadim

    2010-01-01

    A single-cell, 1300-MHz, TM110-like mode vertically deflecting cavity is designed and built for beam slice emittance measurements, and to study the temporal response of negative electron affinity photocathodes in the ERL injector at Cornell University. We describe the cavity shape optimization procedure, RF and mechanical design, its performance with beam.

  7. An inherently safe power reactor module

    International Nuclear Information System (INIS)

    Salerno, L.N.

    1985-01-01

    General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

  8. Comparison of zero-dimensional and one-dimensional thermonuclear burn computations for the reversed-field pinch reactor (RFPR)

    International Nuclear Information System (INIS)

    Nebel, R.A.; Hagenson, R.L.; Moses, R.W.; Krakowski, R.A.

    1980-01-01

    Conceptual fusion reactor designs of the Reversed-Field Pinch Reactor (RFPR) have been based on profile-averaged zero-dimensional (point) plasma models. The plasma response/performance that has been predicted by the point plasma model is re-examined by a comprehensive one-dimensional (radial) burn code that has been developed and parametrically evaluated for the RFPR. Agreement is good between the zero-dimensional and one-dimensional models, giving more confidence in the RFPR design point reported previously from the zero-dimensional analysis

  9. Corrosion control in CANDU nuclear power reactors

    International Nuclear Information System (INIS)

    Lesurf, J.E.

    1974-01-01

    Corrosion control in CANDU reactors which use pressurized heavy water (PHW) and boiling light water (BLW) coolants is discussed. Discussions are included on pressure tubes, primary water chemistry, fuel sheath oxidation and hydriding, and crud transport. It is noted that corrosion has not been a significant problem in CANDU nuclear power reactors which is a tribute to design, material selection, and chemistry control. This is particularly notable at the Pickering Nuclear Generating Station which will have four CANDU-PHW reactors of 540 MWe each. The net capacity factor for Pickering-I from first full power (May 1971) to March 1972 was 79.5 percent, and for Pickering II (first full power November 1971) to March 1972 was 83.5 percent. Pickering III has just reached full power operation (May 1972) and Pickering IV is still under construction. Gentilly CANDU-BLW reached full power operation in May 1972 after extensive commissioning tests at lower power levels with no major corrosion or chemistry problems appearing. Experience and operating data confirm that the value of careful attention to all aspects of corrosion control and augur well for future CANDU reactors. (U.S.)

  10. Reactors of different types in the world nuclear power

    International Nuclear Information System (INIS)

    Simonov, K.V.

    1991-01-01

    The status of the world nuclear power is briefly reviewed. It is noted that PWR reactors have decisive significance in the world power. The second place is related to gas-cooled graphite-moderated reactors. Channel-type heavy water moderated reactors are relatively important. Nuclear power future is associated with fast liquid-metal cooled breeder reactors

  11. Neutrino mass models and the implications of a non-zero reactor angle

    International Nuclear Information System (INIS)

    King, S.F.

    2009-01-01

    In this talk we survey some of the recent promising developments in the search for the theory behind neutrino mass and mixing, and indeed all fermion masses and mixing. The talk is organized in terms of a neutrino mass models decision tree according to which the answers to experimental questions provide sign posts to guide through the maze of theoretical models eventually towards a complete theory of flavour and unification. It is also discussed the theoretical implications of the measurement of non-zero reactor angle, as hinted at by recent experimental measurements.

  12. Power reactor events, May-June 1986

    International Nuclear Information System (INIS)

    Massaro, S.A.

    1986-12-01

    Power Reactor Events is a bi-monthly newsletter that compiles operating experience information about commercial nuclear power plants. This includes summaries of noteworthy events and listings and/or abstracts of USNRC and other documents that discuss safety-related or possible generic issues. It is intended to feed back some of the lessons learned from operational experience to the various plant personnel, i.e., managers, licensed reactor operators, training coordinators, and support personnel. Events at the following plants are reported: McGuire Unit 1; Susquehanna Units 1 and 2; Browns Ferry Units 1, 2, and 3; and River Bend Unit 1

  13. Time-optimal control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1987-01-01

    Control laws that permit adjustments in reactor power to be made in minimum time and without overshoot have been formulated and demonstrated. These control laws which are derived from the standard and alternate dynamic period equations, are closed-form expressions of general applicability. These laws were deduced by noting that if a system is subject to one or more operating constraints, then the time-optimal response is to move the system along these constraints. Given that nuclear reactors are subject to limitations on the allowed reactor period, a time-optimal control law would step the period from infinity to the minimum allowed value, hold the period at that value for the duration of the transient, and then step the period back to infinity. The change in reactor would therefore be accomplished in minimum time. The resulting control laws are superior to other forms of time-optimal control because they are general-purpose, closed-form expressions that are both mathematically tractable and readily implanted. Moreover, these laws include provisions for the use of feedback. The results of simulation studies and actual experiments on the 5 MWt MIT Research Reactor in which these time-optimal control laws were used successfully to adjust the reactor power are presented

  14. Nuclear Power Reactor simulator - based training program

    International Nuclear Information System (INIS)

    Abdelwahab, S.A.S.

    2009-01-01

    nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude

  15. Reactor power system deployment and startup

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  16. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

    International Nuclear Information System (INIS)

    Dhami, P.S; Yadav, J.S; Agarwal, K.

    2017-01-01

    Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

  17. Power control device for nuclear reactors

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo

    1984-01-01

    Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

  18. Radiation streaming in power reactors. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Lahti, G.P.; Lee, R.R.; Courtney, J.C. (eds.)

    1979-02-01

    Separate abstracts are included for each of the 14 papers given at a special session on Radiation Streaming in Power Reactors held on November 15 at the American Nuclear Society 1978 Winter Meeting in Washington, D.C. The papers describe the methods of calculation, the engineering of shields, and the measurement of radiation environments within the containments of light water power reactors. Comparisons of measured and calculated data are used to determine the accuracy of computer predictions of the radiation environment. Specific computational and measurement techniques are described and evaluated. Emphasis is on radiation streaming in the annular region between the reactor vesel and the primary shield and its resultant environment within the primary containment.

  19. Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

    International Nuclear Information System (INIS)

    Ganguly, Chaitanyamoy

    2015-01-01

    Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

  20. High Flux Isotope Reactor power upgrade status

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Hale, R.E.; Cheverton, R.D.

    1997-01-01

    A return to 100-MW operation is being planned for the High Flux Isotope Reactor (HFIR). Recent improvements in fuel element manufacturing procedures and inspection equipment will be exploited to reduce hot spot and hot streak factors sufficiently to permit the power upgrade without an increase in primary coolant pressure. Fresh fuel elements already fabricated for future use are being evaluated individually for power upgrade potential based on their measured coolant channel dimensions

  1. The IAEA power reactor information system - PRIS

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  3. Recalculating the steady state conditions of the V-1000 zero-power facility at Kurchatov Institute using Monte Carlo and nodal diffusion codes

    Energy Technology Data Exchange (ETDEWEB)

    Sahlberg, Ville [VTT Technical Research Centre of Finland Ltd, VTT (Finland)

    2017-09-15

    Continuous-energy Monte Carlo reactor physics code Serpent 2 was used to model the critical steady state conditions measured in V-1000 zero-power critical facility at Kurchatov Institute (KI), Moscow in 1990-1992. The Serpent 2 results were compared to measurements and Serpent 2 was used to generate group constants for reactor dynamics code HEXTRAN. The results of a HEXTRAN calculation of the steady state were compared to Serpent 2. The relative power density distribution of the SERPENT2 calculations compared with the measurements was within the statistical accuracy. The comparison of HEXTRAN and Serpent 2 node-wise relative power density distributions showed an accuracy of ±10%.

  4. Power balance in an Ohmically heated fusion reactor

    International Nuclear Information System (INIS)

    Christiansen, J.P.; Roberts, K.V.

    1982-01-01

    A simplified power-balance equation (zero-dimensional model) is used to study the performance of an Ohmically heated fusion reactor with emphasis on a pulsed reversed-field pinch concept (RFP). The energy confinement time tausub(E) is treated as an adjustable function, and empirical tokamak scaling laws are employed in the numerical estimates, which are supplemented by 1-D ATHENE code calculations. The known heating rates and energy losses are represented by the net energy replacement time tausub(W), which is exhibited as a surface in density (n) and temperature (T) space with a saddle point (nsub(*), Tsub(*)), the optimum ignition point. It is concluded that i) ignition by Ohmic heating is more practicable for the RFP reactor than for a tokamak reactor with the same tausub(E), (ii) if at fixed current the minor radius can be reduced or at fixed minor radius the current can be increased, then it is found that Ohmic ignition becomes more likely when present tokamak scaling laws are used. More definitive estimates require, however, a knowledge of tausub(E), which can only be obtained by establishing a reliable set of experimental RFP scaling laws and, in particular, by extending RFP experiments closer to the reactor regime. (author)

  5. The program of reactors and nuclear power plants

    International Nuclear Information System (INIS)

    Calabrese, Carlos R.

    2001-01-01

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

  6. Zero-emission fuel-fired power plants with ion transport membrane

    NARCIS (Netherlands)

    Yantovski, E.; Gorski, J.; Smyth, B.; ten Elshof, Johan E.

    2004-01-01

    Firstly, some points in relation to the history of zero-emissions power cycles are highlighted. Amongst the many schemes, only one which deals with the combustion of a fuel in “artificial air” (i.e. a mixture of oxygen and re-circulated carbon dioxide), is selected. This paper describes the zero

  7. Correlations between power and test reactor data bases

    International Nuclear Information System (INIS)

    Guthrie, G.L.; Simonen, E.P.

    1989-02-01

    Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

  8. Monitoring device for the reactor power distribution

    International Nuclear Information System (INIS)

    Uematsu, Hitoshi; Tsuiki, Makoto

    1982-01-01

    Purpose: To enable accurate monitoring for the power distribution in a short time, as well as independent detection for in-core neutron flux detectors in abnormal operation due to failures or like other causes to thereby surely provide reliable substitute values. Constitution: Counted values are inputted from a reactor core present status data detector by a power distribution calculation device to calculate the in-core neutron flux density and the power distribution based on previously stored physical models. While on the other hand, counted value from the in-core neutron detectors and the neutron flux distribution and the power distribution calculated from the power distribution calculation device are inputted from a BCF calculation device to compensate the counting errors incorporated in the counted value from the in-core neutron flux detectors and the calculation errors incorporated in the power distribution calculated in the power distribution calculation device respectively and thereby calculate the power distribution in the reactor core. Further, necessary data are inputted to the power distribution calculation device by an input/output device and the results calculated in the BCF calculation device are displayed. (Aizawa, K.)

  9. Assessment of the thorium fuel cycle in power reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Homan, F.J.; Allen, E.J.

    1977-01-01

    A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

  10. Self-powered detectors for power reactors: an overview

    International Nuclear Information System (INIS)

    Ma, J.

    2006-01-01

    In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

  11. Self-powered detectors for power reactors: an overview

    Energy Technology Data Exchange (ETDEWEB)

    Ma, J. [Univ. of Western Ontario, Dept. of Mechanical and Materials Engineering, London, Ontario (Canada)]. E-mail: jma64@uwo.ca

    2006-07-01

    In this paper, Self-Powered Detectors (SPDs) for applications in nuclear power reactors have been reviewed. Based on their responses to radiation, these detectors can be divided into delayed response Self-Powered Neutron Detector (SPND), prompt response SPND and Self-Powered Gamma Detector (SPGD). The operational principles of these detectors are presented and their distinctive characteristics are examined accordingly. The analytical models and Monte Carlo method to calculate the responses of these detectors to neutron flux and external gamma rays are reviewed. The paper has also considered some related signal processing techniques, such as detector calibrations and detector signal compensations. Furthermore, a couple of failure modes have also been analyzed. Finally, applications of SPD in nuclear power reactors are summarized. (author)

  12. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the

  13. Utilization of a university reactor for public acceptance

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun [Kyung Hee Univ., Gyeonggi-do (Korea, Republic of). Reactor Research and Education Center

    2015-06-15

    AGN-201K is a university reactor in Kyung Hee University (KHU) mainly used for student education as training short course as well as academic course for senior-level. After the Fukushima accident, public concern on radiation hazard has been increased beyond rational level at a neighboring country. It was found that AGN-201K be the perfect tool for interaction with general public. It is very safe to operate with general participants because it is adapted to the very low power. However, radiation level is reasonably high to detect and shield for practice. KHU has a Regional Radiation Monitoring Post where environmental radiation level at Suwon city is continuously measured. In this facility, radiation level at soil, rain, and local agricultural products were measured and reported to the national monitoring headquarter. A new mission of reactor research and education center of KHU has been tried from last summer. Facilities were opened for high school students and teachers for their science camps during summer and winter. A special public acceptance program named as experience camp for understanding the nuclear power and radiation was held 6 times for the last one-year period. Even though number of attendee was limited and small, feedback from participants was hot and positive enough to make professors be ready to sacrifice their personal time.

  14. Utilization of a university reactor for public acceptance

    International Nuclear Information System (INIS)

    Kim, Myung Hyun

    2015-01-01

    AGN-201K is a university reactor in Kyung Hee University (KHU) mainly used for student education as training short course as well as academic course for senior-level. After the Fukushima accident, public concern on radiation hazard has been increased beyond rational level at a neighboring country. It was found that AGN-201K be the perfect tool for interaction with general public. It is very safe to operate with general participants because it is adapted to the very low power. However, radiation level is reasonably high to detect and shield for practice. KHU has a Regional Radiation Monitoring Post where environmental radiation level at Suwon city is continuously measured. In this facility, radiation level at soil, rain, and local agricultural products were measured and reported to the national monitoring headquarter. A new mission of reactor research and education center of KHU has been tried from last summer. Facilities were opened for high school students and teachers for their science camps during summer and winter. A special public acceptance program named as experience camp for understanding the nuclear power and radiation was held 6 times for the last one-year period. Even though number of attendee was limited and small, feedback from participants was hot and positive enough to make professors be ready to sacrifice their personal time.

  15. Universally composable zero-knowledge arguments and commitments from signature cards

    NARCIS (Netherlands)

    D. Hofheinz (Dennis); J. Müller-Quade; D. Unruh

    2007-01-01

    htmlabstractIn the Universal Composability framework many cryptographic tasks cannot be built from scratch. Additional \\helping" functionalities are needed to realise zero-knowledge or bit commitment. However, all the additional functionalities presented in the literature so far have to be

  16. Method of estimating the reactor power distribution

    International Nuclear Information System (INIS)

    Mitsuta, Toru; Fukuzaki, Takaharu; Doi, Kazuyori; Kiguchi, Takashi.

    1984-01-01

    Purpose: To improve the calculation accuracy for the power distribution thereby improve the reliability of power distribution monitor. Constitution: In detector containing strings disposed within a reactor core, movable type neutron flux monitors are provided in addition to position fixed type neutron monitors conventionally disposed so far. Upon periodical monitoring, a power distribution X1 is calculated from a physical reactor core model. Then, a higher power position X2 is detected by position detectors and value X2 is sent to a neutron flux monitor driving device to displace the movable type monitors to a higher power position in each of the strings. After displacement, the value X1 is amended by an amending device using measured values from the movable type and fixed type monitors and the amended value is sent to a reactor core monitor device. Upon failure of the fixed type monitors, the position is sent to the monitor driving device and the movable monitors are displaced to that position for measurement. (Sekiya, K.)

  17. Reactor power distribution pattern judging device

    International Nuclear Information System (INIS)

    Ikehara, Tadashi.

    1992-01-01

    The judging device of the present invention comprises a power distribution readout system for intaking a power value from a fuel segment, a neural network having an experience learning function for receiving a power distribution value as an input variant, mapping it into a desirable property and self-organizing the map, and a learning date base storing a plurality of learnt samples. The read power distribution is classified depending on the similarity thereof with any one of representative learnt power distribution, and the corresponding state of the reactor core is outputted as a result of the judgement. When an error is found in the classified judging operation, erroneous cases are additionally learnt by using the experience and learning function, thereby improving the accuracy of the reactor core characteristic estimation operation. Since the device is mainly based on the neural network having a self-learning function and a pattern classification and judging function, a judging device having a human's intuitive pattern recognition performance and a pattern experience and learning performance is obtainable, thereby enabling to judge the state of the reactor core accurately. (N.H.)

  18. [The power of religion in the public sphere] / Alar Kilp

    Index Scriptorium Estoniae

    Kilp, Alar, 1969-

    2012-01-01

    Arvustus: Buthler, Judith, Habermas, Jürgen, Taylor, Charles, West, Cornel. The power of religion in the public sphere. (Eduardo Mendieta, Jonathan VanAntwerpen (eds.) Afterword by Craig Calhoun.) New York ; Chichester : Columbia University Press, 2011

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  20. Tandem mirror reactor power balance studies

    International Nuclear Information System (INIS)

    Gorker, G.E.; Perkins, L.J.

    1985-01-01

    A tandem mirror reactor (TMR) power plant balance model has been developed and is now being used as a computer aid for performing parametric studies. End-cell power injection into the plasma and the physics thermal Q are used to determine the fusion power. About 80% of the fusion power is transferred by high-energy neutrons to the blanket modules and structures. The other 20% of the fusion power in the high-energy alpha particles is used to heat the deuterium-tritium (D-T) plasma. Most of the plasma-ionized particles transfer their energy to the halo dumps and direct converters. The plant efficiency is calculated for three different system cycles: (1) the pressurized water/saturated steam cycle; (2) the superheated steam cycle; and (3) the more complex superheat/reheat cycle. There is a signficiant improvement in plant efficiency as the electrical power multiplication factor and steam cycle efficiency increases

  1. TRIGA research reactors with higher power density

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    1994-01-01

    The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (≅48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D 2 O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 10 14 n/cm 2 ·s and fast fluxes (>0.1 MeV) of 2 x 10 14 n/cm 2 ·s. The core centerline thermal neutron flux in the D 2 O reflector is about 2 x 10 14 n/cm 2 ·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D 2 O reflector of 3 x 10 14 n/cm 2 ·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

  2. In core system mapping reactor power distribution

    International Nuclear Information System (INIS)

    Yoriyaz, H.; Moreira, J.M.L.

    1989-01-01

    Based on the signals of SPND'S (Self Powered Neutron Detectors) distributed inside of a core, the spatial power distribution is obtained using the MAP program, developed in this work. The methodology applied in MAP program uses a least mean square technique to calculate expansion coefficients that depend on the SPND'S signals. The final power or neutron flux distribution is obtained by a combination of certains functions or expansion modes that are provided from diffusion calculation with the CITATION code. The MAP program is written in PASCAL language and will be used in IEA-R1 reactor for assisting its operation. (author) [pt

  3. DOE University Reactor Sharing Program. Renewal for 1994--1995

    International Nuclear Information System (INIS)

    Chappas, W.J.; Adams, V.G.

    1994-01-01

    The Department of Energy University Reactor Sharing Program at University of Maryland, College Park (UMCP) has, once again, stimulated a broad use of the reactor facilities by undergraduate and graduate students, visitors, and professionals. Participants are exposed to topics such as nuclear engineering, radiation safety, and nuclear reactor operations. This information is presented through various means including tours, slide presentations, experiments, and discussions. Student research using the MUTR is also encouraged. In addition, the Reactor Sharing Program here at the University of Maryland does not limit itself to the confines of the TRIGA reactor facility. Incorporated in the program are the Maryland University Neutron Activation Analysis Laboratory, the Maryland University Radiation Effects Laboratory, and the UMCP 2x4 Thermal Hydraulic Loop. These facilities enhance and give an added dimension to the tours and experiments

  4. Neutron fluctuations in accelerator driven and power reactors via backward master equations

    International Nuclear Information System (INIS)

    Zhifeng Kuang

    2000-05-01

    The transport of neutrons in a reactor is a random process, and thus the number of neutrons in a reactor is a random variable. Fluctuations in the number of neutrons in a reactor can be divided into two categories, namely zero noise and power reactor noise. As the name indicates, they dominate (i.e. are observable) at different power levels. The reasons for their occurrences and utilization are also different. In addition, they are described via different mathematical tools, namely master equations and the Langevin equation, respectively. Zero noise carries information about some nuclear properties such as reactor reactivity. Hence methods such as Feynman- and Rossi-alpha methods have been established to determine the subcritical reactivity of a subcritical system. Such methods received a renewed interest recently with the advent of the so-called accelerator driven systems (ADS). Such systems, intended to be used either for energy production or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those of the traditionally used radioactive sources which were also assumed in the derivation of the Feynman- and Rossi-alpha formulae. Therefore it is necessary to re-derive the Feynman- and Rossi-alpha formulae. Such formulae for ADS have been derived recently but in simpler neutronic models. One subject of this thesis is the extension of such formulae to a more general case in which six groups of delayed neutron precursors are taken into account, and the full joint statistics of the prompt and all delayed groups is included. The involved complexity problems are solved with a combination of effective analytical techniques and symbolic algebra codes. Power reactor noise carries information about parametric perturbation of the system. Langevin technique has been used to extract such information. In such a treatment, zero noise has been neglected. This is a pragmatic

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

    International Nuclear Information System (INIS)

    Chenais, Jacques

    2016-01-01

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

  6. Basic training of nuclear power reactor personnel

    International Nuclear Information System (INIS)

    Palabrica, R.J.

    1981-01-01

    The basic training of nuclear power reactor personnel should be given very close attention since it constitutes the foundation of their knowledge of nuclear technology. Emphasis should be given on the thorough understanding of basic nuclear concepts in order to have reasonable assurance of successful assimilation by those personnel of more specialized and advanced concepts to which they will be later exposed. Basic training will also provide a means for screening to ensure that those will be sent for further spezialized training will perform well. Finally, it is during the basic training phase when nuclear reactor operators will start to acquire and develop attitudes regarding reactor operation and it is important that these be properly founded. (orig.)

  7. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

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

  8. Systems and methods for tracking a device in zero-infrastructure and zero-power conditions, and a tracking device therefor

    KAUST Repository

    Shamim, Atif

    2017-03-23

    Disclosed are embodiments for a tracking device having multiple layers of localization and communication capabilities, and particularly having the ability to operate in zero-infrastructure or zero-power conditions. Also disclosed are methods and systems that enhance location determination in zero-infrastructure and zero-power conditions. In one example, a device, system and/or method includes an infrastructure-based localization module, an infrastructure-less localization module and a passive module that can utilize at least two of the modules to determine a location of the tracking device.

  9. Rotational perturbations of Friedmann universes in Einstein zero mass scalar theory

    International Nuclear Information System (INIS)

    Krori, K.D.; Sarmah, J.C.; Goswami, D.

    1983-01-01

    The authors find that except in the case of 'perfect dragging', the zero mass scalar field has a damping effect on the rotation of matter. One of the three cases studied here shows that the scalar field may exist only during a certain interval in the course of the evolution of the universe. The zero mass scaler field has acquired particular importance recently because of a suggestion by Weinberg and Wilezek that there should exist a pseudo-scalar boson, the so-called axion, of negligible mass

  10. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  11. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  12. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-09

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

  13. Source driven breeding thermal power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba

    1978-03-01

    The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that the DT device. The efficienct of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed. (aurthor)

  14. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    Mohanakrishnan, P.; Rastogi, B.P.

    1982-01-01

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  15. Source driven breeding thermal power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

    1978-03-01

    Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available - including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (author)

  16. RF Energy Harvesting for Ubiquitous, Zero Power Wireless Sensors

    Directory of Open Access Journals (Sweden)

    Warda Saeed

    2018-01-01

    Full Text Available This paper presents a review of wireless power transfer (WPT followed by a comparison between ambient energy sources and an overview of different components of rectennas that are used for RF energy harvesting. Being less costly and environment friendly, rectennas are used to provide potentially inexhaustible energy for powering up low power sensors and portable devices that are installed in inaccessible areas where frequent battery replacement is difficult, if not impossible. The current challenges in rectenna design and a detailed comparison of state-of-the-art rectennas are also presented.

  17. Power conditioning for space nuclear reactor systems

    Science.gov (United States)

    Berman, Baruch

    1987-01-01

    This paper addresses the power conditioning subsystem for both Stirling and Brayton conversion of space nuclear reactor systems. Included are the requirements summary, trade results related to subsystem implementation, subsystem description, voltage level versus weight, efficiency and operational integrity, components selection, and shielding considerations. The discussion is supported by pertinent circuit and block diagrams. Summary conclusions and recommendations derived from the above studies are included.

  18. Compact approach to fusion power reactors

    International Nuclear Information System (INIS)

    Hagenson, R.L.; Krakowski, R.A.; Bathke, C.G.; Miller, R.L.

    1984-01-01

    The potential of the Reversed-Field Pinch (RFP) for development into an efficient, compact, copper-coil fusion reactor has been quantified by comprehensive parametric tradeoff studies. These compact systems promise to be competitive in size, power density, and cost to alternative energy sources. Conceptual engineering designs that largely substantiate these promising results have since been completed. This 1000-MWe(net) design is described along with a detailed rationale and physics/technology assessment for the compact approach to fusion

  19. Small and medium power reactors 1985

    International Nuclear Information System (INIS)

    1986-05-01

    This report is intended for designers and planners concerned with Small and Medium Power Reactors. It provides a record of the presentations during the meetings held on this subject at the Agency's General Conference in September 1985. This information should be useful as it indicates the principal findings and main conclusions and recommendations resulting from these meetings. A separate abstract was prepared for each of the 10 presentations in this report

  20. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1986-01-01

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author) [pt

  1. Reactor console replacement at Washington State University

    International Nuclear Information System (INIS)

    Lovas, Thomas A.

    1978-01-01

    A replacement reactor console was installed in 1977 at the W.S.U. 1 MW TRIGA-fueled reactor as the final step in an instrumentation upgrade program. The program was begun circa 1972 with the design, construction and installation of various systems and equipment. Major instruments were installed in the existing console and tested in the course of reactor operation. The culmination of the program was the installation of a cubicle designed and constructed to house the updated instrumentation. (author)

  2. Exergoeconomic estimates for a novel zero-emission process generating hydrogen and electric power

    International Nuclear Information System (INIS)

    Tsatsaronis, George; Kapanke, Kerstin; Maria Blanco Marigorta, Ana

    2008-01-01

    This paper presents the exergoeconomic analysis of a novel process generating electric energy and hydrogen. Coal and high-temperature heat are used as input energy to the process. The process is a true 'zero-emission process' because (a) no NO X is formed during coal combustion with sulfuric acid, and (b) the combustion products CO 2 and SO 2 are removed separately as compressed liquids from the overall process. The process cycle is based on two chemical reactions. The first reaction takes place in an electrolytic cell and delivers the hydrogen product. In the second step, coal reacts with sulfuric acid in a high-pressure combustion reactor. The combustion gas is expanded in a gas turbine to produce electric power. The combustion products are compressed and separated so that almost pure CO 2 can be removed from the cycle. The overall process is characterized by very high energetic and exergetic efficiencies. However, the overall process is very capital intensive. The electrolytic cell dominates the costs associated with the overall process. Detailed results of the thermodynamic simulation, the economic and the exergoeconomic analyses of the process including estimates of the product costs are presented

  3. Critical Power Response to Power Oscillations in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Farawila, Yousef M.; Pruitt, Douglas W.

    2003-01-01

    The response of the critical power ratio to boiling water reactor (BWR) power oscillations is essential to the methods and practice of mitigating the effects of unstable density waves. Previous methods for calculating generic critical power response utilized direct time-domain simulations of unstable reactors. In this paper, advances in understanding the nature of the BWR oscillations and critical power phenomena are combined to develop a new method for calculating the critical power response. As the constraint of the reactor state - being at or slightly beyond the instability threshold - is removed, the new method allows the calculation of sensitivities to different operation and design parameters separately, and thus allows tighter safety margins to be used. The sensitivity to flow rate and the resulting oscillation frequency change are given special attention to evaluate the extension of the oscillation 'detect-and-suppress' methods to internal pump plants where the flow rate at natural circulation and oscillation frequency are much lower than jet pump plants

  4. Nuclear Education and Training Courses as a Commercial Product of a Low Power Research Reactor

    International Nuclear Information System (INIS)

    Böck, H.; Villa, M.; Steinhauser, G.

    2013-01-01

    The Vienna University of Technology (VUT) operates a 250 kW TRIGA Mark II research reactor at the Atominstitut (ATI) since March 1962. This reactor is uniquely devoted to nuclear education and training with the aim to offer an instrument to perform academic research and training. During the past decade a number of requests to the Atominstitut asked for the possibility to offer this reactor for external training courses. Over the years, such courses have been developed as regular courses for students during their academic curricula at the VUT/ATI. The courses cover such subjects as “Reactor physics and kinetics”, and “Reactor instrumentation and control”, in total about 20 practical exercises. Textbooks have been developed in English language for both courses. Target groups for commercial courses are other universities without an access to research reactors (i.e., the Technical University of Bratislava, Slovak Republic, or the University of Manchester, UK), international organisations (i.e., IAEA Dept of Safeguards, training section), research centres (ie. Mol, Belgium) for retraining of their reactor staff or nuclear power plants for staff retraining. These courses have been very successful during the past five years in such a manner that the Atominstitut has now to decline new course applications as the reactor is also used for Masters thesis and PhD work which requires full power operation while courses require low power operation. The paper describes typical training programs, target groups and possible transfers of these courses to other reactors. (author)

  5. Validation of finite element code DELFIN by means of the zero power experiences at the nuclear power plant of Atucha I

    International Nuclear Information System (INIS)

    Grant, C.R.

    1996-01-01

    Code DELFIN, developed in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and currents among elements and a more realistic representation of the hexagonal lattice of the reactor. It can be used for fuel management calculation, Xenon oscillation and spatial kinetics. Using the HUEMUL code for cell calculation (which uses a generalized two dimensional collision probability theory and has the WIMS library incorporated in a data base), the zero power experiences performed in 1974 were calculated. (author). 8 refs., 9 figs., 3 tabs

  6. SP-100 space reactor power system readiness

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  7. Nuclear power reactors and hydrogen storage systems

    International Nuclear Information System (INIS)

    Ibrahim Aly Mahmoud El Osery.

    1980-01-01

    Among conclusions and results come by, a nuclear-electric-hydrogen integrated power system was suggested as a way to prevent the energy crisis. It was shown that the hydrogen power system using nuclear power as a leading energy resource would hold an advantage in the current international situation as well as for the long-term future. Results reported provide designers of integrated nuclear-electric-hydrogen systems with computation models and routines which will allow them to explore the optimal solution in coupling power reactors to hydrogen producing systems, taking into account the specific characters of hydrogen storage systems. The models were meant for average computers of a type easily available in developing countries. (author)

  8. Program of RA reactor start-up to nominal power; Program dizanja reaktora 'RA' na nominalnu snagu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1959-07-01

    The zero start-up program is followed by the program of RA reactor start-up to nominal power. This program is desed in detail and includes the following measurements: radiation characteristics at the exit of the channels; gamma and fast neutron dose distribution in the reactor; influence of absorbers on the reactivity; temperature effect; absolute flux and calibration of ionization chambers; xenon effect; thermal and hydraulics; dosimetry around the reactor; neutron flux in the reactor core and in the reactor hall; heavy water level; thermal characteristics after shutdown. A list of measuring devices and instrumentation is included with the detailed action plan and list of responsible staff members.

  9. Power reactor noise studies and applications

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, V

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  10. Power reactor noise studies and applications

    International Nuclear Information System (INIS)

    Arzhanov, V.

    2002-03-01

    The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

  11. Systems and methods for tracking a device in zero-infrastructure and zero-power conditions, and a tracking device therefor

    KAUST Repository

    Shamim, Atif; Bilal, Rana Muhammad

    2017-01-01

    and systems that enhance location determination in zero-infrastructure and zero-power conditions. In one example, a device, system and/or method includes an infrastructure-based localization module, an infrastructure-less localization module and a passive

  12. Measurement of nuclear reactor noise at low power; Merenje nuklearnog reaktorskog suma na malim snagama

    Energy Technology Data Exchange (ETDEWEB)

    Velickovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1968-07-01

    Theoretical interpretation of reactor noise experiments is based on stochastic model developed and described in this paper. Ratio l/{beta} as well as subcriticality level can be determined bu measuring transfer function. In this paper the ratio l/{beta} was determined directly from auto-correlation functions for different critical configurations of the RB zero power reactor core and not by transfer function. This simplified the procedure significantly. It was found that the 0.5 W power level is most suitable for experimental study of neutron fluctuations. In this case fluctuations are intense compared to noise of the detector and electronic devices used.

  13. Boiler systems for nuclear powered reactors

    International Nuclear Information System (INIS)

    Cook, R.K.; George, B.V.

    1979-01-01

    A power generating plant which comprises a heat source, at least one main steam turbine and at least one main boiler heated by heat from the heat source and providing the steam to drive the turbine, comprises additionally at least one further steam turbine, smaller than the main turbine, and at least one further boiler, of lower capacity than the main boiler, and heated from the same heat source and providing steam for the further turbine. Particularly advantageous in nuclear power stations, where the heat source is a nuclear reactor, the invention enables peak loads, above the normal continuous rating of the main generators driven by the main turbines, to be met by the further turbine(s) and one or more further generators driven thereby. This enables the main turbines to be freed from the thermal stresses of rapid load changes, which stresses are more easily accommodated by the smaller and thus more tolerant further turbine(s). Thus auxiliary diesel-driven or other independent power plant may be made partly or wholly unnecessary. Further, low-load running which would be inefficient if achieved by means of the main turbine(s), can be more efficiently effected by shutting them down and using the smaller further turbine(s) instead. These latter may also be used to provide independent power for servicing the generating plant during normal operation or during emergency or other shutdown, and in this latter case may also serve as a heat sink for the shutdown reactor

  14. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  15. Reactor power control systems in nuclear power plants

    International Nuclear Information System (INIS)

    Nakajima, Kazuo.

    1980-01-01

    Purpose: To enable power control by automatic control rod operation based on the calculated amounts of operation for the control rods determined depending on a power set value from reactor operators or on power variation amounts from other devices. Constitution: When an operator designates an automatic selection by way of a control rod operation panel, automatic signals are applied to a manual-automatic switching circuit and the mode judging circuit of a rod pattern control device. Then, mode signals such as for single operation, load setting, load following and the like produced by the operator are judged in a circuit, wherein a control rod pattern operation circuit calculates the designation for the control rods and the operation amounts for the control rods depending on the designated modes and automatic control is conducted for the control rods by a rod position control circuit, a rod drive control device and the like connected at a rod position monitor device. The reactor power is thus controlled automatically to reduce the operator's labours. The automatic power control can also be conducted in the same manner by the amount of power variations applied to the device from the external device. (Yoshino, Y.)

  16. System for forecasting a reactor power distribution

    International Nuclear Information System (INIS)

    Motoda, Hiroshi; Nishizawa, Yasuo.

    1976-01-01

    Purpose: To dispense with frequent running of detector in a BWR type reactor and permit calculation of the prevailing value and forecast value of power distribution in a specified region in an on-line basis. Constitution: The prevailing power distribution P sub(OZ) (where Z indicates a position in the axial direction) at a given position is estimated by prevailing power distribution estimating means, and the average prevailing power distribution Q sub(OZ) in the core is estimated while making correction of a primary neutron distribution model by core average characteristic measuring means. Then, the estimated core average power distribution Q sub(Z) after alteration of the core flow rate or alteration of Xe concentration is estimated by core average power distribution estimating means. At this time, a forecast power distribution P sub(Z) in a specified region after alteration of the flow rate or alteration of the Xe concentration is calculated on the basis of a relation P sub(Z) = (Q sub(Z)/Q sub(OZ)) by using P sub(OZ), Q sub(OZ) and Q sub(Z). The above calculations are carried out in a short period of time by using a process computer. (Ikeda, J.)

  17. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

    Simulation for safety analysis of actinide recycled fast power reactor has been performed. The objective is to know reactor response about ULOF and ULOF and UTOP simultaneous accident. From parameter result such reactivity feedback, power, temperature, and cooled flow rate can conclusion that reactor have inherent safety system, which can back to new Equilibrium State

  18. The universe in zero words the story of mathematics as told through equations

    CERN Document Server

    Mackenzie, Dana

    2012-01-01

    Most popular books about science, and even about mathematics, tiptoe around equations as if they were something to be hidden from the reader's tender eyes. Dana Mackenzie starts from the opposite premise: He celebrates equations. No history of art would be complete without pictures. Why, then, should a history of mathematics--the universal language of science--keep the masterpieces of the subject hidden behind a veil? The Universe in Zero Words tells the history of twenty-four great and beautiful equations that have shaped mathematics, science, and society--from the elementary

  19. Power distribution monitor for nuclear reactor

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo; Kiguchi, Takashi.

    1974-01-01

    Object: To compare the measured local power region monitor (LPRM) index with the result of a primary calculation to correct the threshold condition for the primary calculation thereby to rapidly grasp and monitor the existing power distribution. Structure: The index of an LPRM disposed in a nuclear reactor is processed in a data processor to remove therefrom a noise, and transmitted to a threshold condition processor to be stored therein. The LPRM index measured by the threshold condition processor is compared with the calculated LPRM value transmitted from the primary processor, whereby the threshold condition is corrected and transmitted to the primary processor. After the completion of calculation, the traversing incore probe (TIP) indexing value is converted to a thermal output distribution or a linear output density distribution and transmitted to an output indicator or an output typewriter. The operator may monitor the existing power distribution by monitoring the output indicator. (Kamimura, M.)

  20. Power supply trip control for nuclear reactor

    International Nuclear Information System (INIS)

    Hager, R.E.; Gutman, Jerzy.

    1987-01-01

    A control system for a trip coil in a switchgear mechanism controls the supply of electrical power to a process control device and ensures de-energization of the trip coil shortly after the trip coil is energized. The trip coil is energized not by an independent dc source as in prior art, but from rectified power from a step down transformer supplied from the switchgear output side. The transformer feeds a rectifier which is connected to the trip coil via a trip activation device. The output of the rectifier can be monitored using an optical converter to determine the ability of the control system to activate the trip coil and the condition of the power supplied to the process control device. The control device may be a rod positioner in a pressurised water nuclear reactor. (author)

  1. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  2. Nuclear safety as applied to space power reactor systems

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1987-01-01

    Current space nuclear power reactor safety issues are discussed with respect to the unique characteristics of these reactors. An approach to achieving adequate safety and a perception of safety is outlined. This approach calls for a carefully conceived safety program which makes uses of lessons learned from previous terrestrial power reactor development programs. This approach includes use of risk analyses, passive safety design features, and analyses/experiments to understand and control off-design conditions. The point is made that some recent accidents concerning terrestrial power reactors do not imply that space power reactors cannot be operated safety

  3. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    Slack, J.; Norton, J.L.; Malkoske, G.R.

    2003-01-01

    therapy machines. Today the majority of the cancer therapy cobalt-60 sources used in the world are manufactured using material from the NRU reactor in Chalk River. The same technology that was used for producing cobalt-60 in a research reactor was then adapted and transferred for use in a CANDU power reactor. In the early 1970s, in co-operation with Ontario Power Generation (formerly Ontario Hydro), bulk cobalt-60 production was initiated in the four Pickering A CANDU reactors located east of Toronto. This was the first full scale production of millions of curies of cobalt-60 per year. As the demand and acceptance of sterilization of medical products grew, MDS Nordion expanded its bulk supply by installing the proprietary Canadian technology in additional CANDUs. Over the years MDS Nordion has partnered with CANDU reactor owners to produce cobalt-60 at various sites. CANDU reactors that have, or are still producing cobalt-60, include Pickering A, Pickering B, Gentilly 2, Embalse in Argentina, and Bruce B. In conclusion, the technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and Atomic Energy of Canada, has been safely, economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world. MDS Nordion is presently adding three more CANDU power reactors to its supply chain. These three additional cobalt producing CANDU's will help supplement the ability of the health care industry to provide safe, sterile, medical disposable products to people around the world. As new applications for cobalt-60 are identified, and the demand for bulk cobalt-60 increases, MDS Nordion and AECL

  4. Nuclear power reactors in the world. April 1990 ed.

    International Nuclear Information System (INIS)

    1990-01-01

    This is the tenth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 1989 on power reactors operating or under construction, and shut down; Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's power reactor information system (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States

  5. Nuclear power reactors in the world. Apr 1991 ed.

    International Nuclear Information System (INIS)

    1991-01-01

    This is the eleventh edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: General information as of the end of 1990, on power reactors operating or under construction, and shut down; performance data on reactors operating in the Agency's Member States, as reported to the IAEA. This information is collected by the Agency by circulating questionnaires to the Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States. 5 figs, 19 tabs

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  8. After LS1: Magnet powering with zero downtime - A dream?

    International Nuclear Information System (INIS)

    Zerlauth, M.

    2012-01-01

    Despite a number of improvements already applied in the course of the year, the magnet powering system of the LHC still accounts for around 50% of the premature beam dumps. This number might even further increase when moving to higher beam energies in the next years. With mitigations of radiation effects and the prospects for beam induced magnet quenches being discussed elsewhere, we aim at identifying possible mid- and long-term improvements within the various equipment systems to further reduce the number of equipment failures leading to a loss of the particle beams. Amongst others, this includes the sensitivity of equipment to external causes such as electromagnetic perturbations or perturbations on the electrical network. To conclude, the gain of the identified mitigations will have to be balanced against the potential impact on schedule and cost. (author)

  9. Superconducting RF activities at Cornell University

    International Nuclear Information System (INIS)

    Kirchgessner, J.; Barnes, P.; Belomestnykh, S.; Chojnacki, E.; Ehrlich, R.; Flynn, G.; Graber, J.; Hartung, W.; Hays, T.; Kaplan, R.; and others.

    1996-01-01

    The decision was made to try to increase the luminosity of CESR as high as possible with a single magnet ring. This, the Phase-III plan, can yield a luminosity of 1 x 10 33 with 45 bunches in each beam for a total current in each beam of 0.5 amperes. This plan utilizes four superconducting, single cell cavities. The use of only four SC accelerating cells as compared to the present twenty normal conducting cells decreases both the broad band and narrow band impedances sufficiently to allow stable operation at these half ampere current levels. (R.P.)

  10. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  11. Decommissioning of Swedish nuclear power reactors. Technology and costs

    International Nuclear Information System (INIS)

    1994-06-01

    The main topics discussed are planning, technology and costs of decommissioning nuclear power reactors. Oskarshamn-3 (BWR) and Ringhals-4 (PWR) have been used as reference reactors. 29 refs, figs, tabs

  12. Conversion and standardization of university reactor fuels using low-enrichment uranium - Options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The U.S. Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the U.S. Department of Energy. (author)

  13. Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs

    International Nuclear Information System (INIS)

    Harris, D.R.; Matos, J.E.; Young, H.H.

    1985-01-01

    The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab

  14. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    Full text: The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance

  15. To question of NPP power reactor choice for Kazakhstan

    International Nuclear Information System (INIS)

    Batyrbekov, G.A.; Makhanov, Y.M.; Reznikova, R.A.; Sidorenco, A.V.

    2004-01-01

    The requirements to NPP power reactors that will be under construction in Kazakhstan are proved and given in the report. A comparative analysis of the most advanced projects of power reactors with light and heavy water under pressure of large, medium and low power is carried out. Different reactors have been considered as follows: 1. Reactors with high-power (700 MW(el) and up) such as EPR, French - German reactor; CANDU-9, Canadian heavy-water reactor; System 80+, developed by ABB Combustion Engineering company, USA; KNGR, Korean reactor of the next generation; APWR, Japanese advanced reactor; WWER-1000 (V-392) - development of Atomenergoproect /Gydropress, Russian Federation; EP 1000, European passive reactor. 2. Reactors with medium power (300 MW (el) - 700 MW (el): AP-600, passive PWR of the Westinghouse company; CANDU-6, Canadian heavy-water reactor; AC-600, Chinese passive PWR; WWER-640, Russian passive reactor; MS-600 Japanese reactor of Mitsubishi Company; KSNP-600, South Korean reactor. 3. Reactors with low power (a few MW(el)- 300 MW(el)): IRIS, reactor of IV generation, developed by the International Corporation of 13 organizations from 7 countries, SMART, South Korean integrated reactor; CAREM, Argentina integrated reactor; MRX, Japanese integrated reactor; 'UNITERM', Russian NPP with integrated reactor, development of NIKIET; AHEC-80, Russian NPP, developed by OKBM. A comparison of the projects of the above-mentioned power reactors was carried out with respect to 15 criteria of nuclear, radiating, ecological safety and economic competitiveness, developed especially for this case. Data on a condition and prospects of power production and power consumption, stations and networks in Kazakhstan necessary for the choice of projects of NPP reactors for Kazakhstan are given. According to the data a balance of power production and power consumption as a whole in the country was received at the level of 59 milliard kw/h. However, strong dis balance in the

  16. power system reliability in supplying nuclear reactors

    International Nuclear Information System (INIS)

    Gad, M.M.M.

    2007-01-01

    this thesis presents a simple technique for deducing minimal cut set (MCS) from the defined minimal path set (MPS) of generic distribution system and this technique have been used to evaluate the basic reliability indices of Egypt's second research reactor (ETRR-2) electrical distribution network. the alternative system configurations are then studied to evaluate their impact on service reliability. the proposed MCS approach considers both sustained and temporary outage. the temporary outage constitutes an important parameter in characterizing the system reliability indices for critical load point in distribution system. it is also consider the power quality impact on the reliability indices

  17. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1987-01-01

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10 -8 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt

  18. Obituary: Cornell H. Mayer, 1921-2005

    Science.gov (United States)

    Radhakrishnan, Venkataraman

    2006-12-01

    Cornell (Connie) H. Mayer, a pioneer of radio astronomy, died on 19 November 2005 of congestive heart failure at his home in Mt. Vernon, Virginia. He was eighty-three. Cornell Mayer was born in Ossian, Iowa on 10 December 1921. After graduating from the University of Iowa in 1943, he joined the Navy during World War II and was stationed at the Naval Research Laboratory (NRL) in Washington, DC. There he assisted Fred T. Haddock in the development of the first radar antenna inside a submarine periscope. This device has been credited with shortening the war in the Pacific because of the number of Japanese ships that were sunk with its aid. With Haddock, Connie also discovered centimeter-wave radio bursts from the sun coincident with solar flares. They made the first detection of thermal radio emission from the Orion nebula and other galactic HII regions. They also detected extragalactic objects and thus initiated the important field of centimeter-wave astronomy. Their observations were made with a 50-foot parabolic reflector on a gun mount located on the roof of one of the NRL buildings. This telescope had the world's highest radio resolving power for many years. With Haddock's departure to the University of Michigan in 1956 to create a new radio observatory there, Connie became head of a group in the Radio Astronomy Branch at NRL, where he remained until his retirement in 1980. Much of his work involved the measurement of planetary temperatures by analysis of radio emissions. By making technical innovations in instrumentation--such as replacing disc choppers with a ferrite switch to compare the sky and reference load, or using argon gas tubes for calibration--Connie greatly improved the performance of his equipment. This resulted in the discovery of an astonishing, 600oC surface temperature of Venus, which contradicted the widespread notion that Venus was similar to the Earth and potentially habitable. In spite of the extraordinarily careful and systematic way that

  19. Study on reactor power transient characteristics (reactor training experiments). Control rod reactivity calibration by positive period method and other experiment

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Sunagawa, Takeyoshi

    2014-01-01

    In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity calibration experiments using positive Period method were carried out for shim safety rod and regulating rod, respectively. The results were obtained as reasonable values in comparison with the nominal value of the UTR-KINKI. The measurement of reactor power change after reactor scram was performed, and the presence of the delayed neutron precursor was confirmed by calculating the half-life. The spatial dose rate measurement experiment of neutrons and γ-rays in the reactor room in a reactor power 1W operating conditions were also performed. (author)

  20. Introduction to nuclear power reactors and their health physics systems

    International Nuclear Information System (INIS)

    Brtis, J.S.

    1982-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  2. Cascade: a high-efficiency ICF power reactor

    International Nuclear Information System (INIS)

    Pitts, J.H.

    1985-01-01

    Cascade attains a net power-plant efficiency of 49% and its cost is competitive with high-temperature gas-cooled reactor, pressurized-water reactor, and coal-fired power plants. The Cascade reactor and blanket are made of ceramic materials and activation is 6 times less than that of the MARS Tandem Mirror Reactor operating at comparable power. Hands-on maintenance of the heat exchangers is possible one day after shutdown. Essentially all tritium is recovered in the vacuum system, with the remainder recovered from the helium power conversion loop. Tritium leakage external to the vacuum system and power conversion loop is only 0.03 Ci/d

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  4. Proceedings of first SWCR-KURRI academic seminar on research reactors and related research topics

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Cong, Zhebao

    1986-01-01

    These are the proceedings of an academic seminar on research reactors and related research topics held at the Southwest Centre for Reactor Engineering Research and Design in Chengdu, Sichuan, People's Republic of China in September 24-26 in 1985. Included are the chairmen's addresses and 10 papers presented at the seminar in English. The titles of these papers are: (1) Nuclear Safety and Safeguards, (2) General Review of Thorium Research in Japanese Universities, (3) Comprehensive Utilization and Economic Analysis of the High Flux Engineering Test Reactor, (4) Present States of Applied Health Physics in Japan, (5) Neutron Radiography with Kyoto University Reactor, (6) Topics of Experimental Works with Kyoto University Reactor, (7) Integral Check of Nuclear Data for Reactor Structural Materials, (8) The Reactor Core, Physical Experiments and the Operation Safety Regulation of the Zero Energy Thermal Reactor for PWR Nuclear Power Plant, (9) HFETR Core Physical Parameters at Power, (10) Physical Consideration for Loads of Operated Ten Cycles in HFETR. (author)

  5. Revision of the second basic plans of power reactor development in Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    1978-01-01

    Revision of the second basic plans concerning power reactor development in PNC (Power Reactor and Nuclear Fuel Development Corporation) is presented. (1) Fast breeder reactors: As for the experimental fast breeder reactor, after reaching the criticality, the power is raised to 50 MW thermal output within fiscal 1978. The prototype fast breeder reactor is intended for the electric output of 200 MW -- 300 MW, using mixed plutonium/uranium oxide fuel. Along the above lines, research and development will be carried out on reactor physics, sodium technology, machinery and parts, nuclear fuel, etc. (2) Advanced thermal reactor: The prototype advanced thermal reactor, with initial fuel primarily of slightly enriched uranium and heavy water moderation and boiling water cooling, of 165 MW electric output, is brought to its normal operation by the end of fiscal 1978. Along the above lines, research and development will be carried out on reactor physics, machinery and parts, nuclear fuel, etc. (Mori, K

  6. Nuclear reactor power for an electrically powered orbital transfer vehicle

    Science.gov (United States)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Kia, T.; Nesmith, B.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low earth orbit (LEO) and geosynchronous earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant.

  7. Nuclear reactor power for an electrically powered orbital transfer vehicle

    International Nuclear Information System (INIS)

    Jaffe, L.; Beatty, R.; Bhandari, P.

    1987-01-01

    To help determine the systems requirements for a 300-kWe space nuclear reactor power system, a mission and spacecraft have been examined which utilize electric propulsion and this nuclear reactor power for multiple transfers of cargo between low Earth orbit (LEO) and geosynchronous Earth orbit (GEO). A propulsion system employing ion thrusters and xenon propellant was selected. Propellant and thrusters are replaced after each sortie to GEO. The mass of the Orbital Transfer Vehicle (OTV), empty and dry, is 11,000 kg; nominal propellant load is 5000 kg. The OTV operates between a circular orbit at 925 km altitude, 28.5 deg inclination, and GEO. Cargo is brought to the OTV by Shuttle and an Orbital Maneuvering Vehicle (OMV); the OTV then takes it to GEO. The OTV can also bring cargo back from GEO, for transfer by OMV to the Shuttle. OTV propellant is resupplied and the ion thrusters are replaced by the OMV before each trip to GEO. At the end of mission life, the OTV's electric propulsion is used to place it in a heliocentric orbit so that the reactor will not return to Earth. The nominal cargo capability to GEO is 6000 kg with a transit time of 120 days; 1350 kg can be transferred in 90 days, and 14,300 kg in 240 days. These capabilities can be considerably increased by using separate Shuttle launches to bring up propellant and cargo, or by changing to mercury propellant

  8. ATMEA and medium power reactors. The ATMEA joint venture and the ATMEA1 medium power reactor

    International Nuclear Information System (INIS)

    Mathet, Eric; Castello, Gerard

    2012-01-01

    This Power Point presentation presents the ATMEA company (a joint venture of Areva and Mitsubishi), the main features of its medium power reactor (ATMEA1) and its building arrangement, indicates the general safety objectives. It outlines the features of its robust design which aim at protecting, cooling down and containing. It indicates the regulatory and safety frameworks, comments the review of the safety options by the ASN and the results of this assessment

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  10. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  11. The control of emissions from nuclear power reactors in Canada

    International Nuclear Information System (INIS)

    Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

    1988-01-01

    Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

  12. CHESS-the Cornell High Energy Synchrotron Source

    International Nuclear Information System (INIS)

    Batterman, B.W.; Cornell Univ., Ithaca, NY

    1980-01-01

    The Wilson Laboratory at Cornell University has done pioneering work on development of high energy synchrotrons. In the last decade, the 12 GeV synchrotron has been the most energetic electron synchrotron in the world. In 1975 plans were formulated to build a 4-8 GeV storage ring in the same tunnel as the synchrotron and to use the latter as the injector for the storage ring. This small radius (the normal bend magnets have R = 87 m), coupled with the relatively high electron energy of the storage ring, makes these magnets potent sources of synchrotron radiation. In June of 1978 the National Science Foundation funded a project to create CHESS, the Cornell High Energy Synchrotron Source. (orig./FKS)

  13. Biography of Professor Cornel Tiberiu Opriş. Professional maturity.

    Science.gov (United States)

    Rotaru, Alexandru; Rotaru, Horatiu

    2017-01-01

    Professor Cornel Tiberiu Opris was the founder and Chair of the Clinic and University Department of Oral and Maxillofacial Surgery in Cluj, after the Education Reform of 1948. The article illustrates how the founder of these institutions led a valiant struggle for obtaining and arranging a location for the newly established Faculty of Dentistry, within the Institute of Medicine and Pharmacy. Professor Cornel Tiberiu Opriş established himself as the most prodigious researcher at the Faculty for over a quarter-century, until his retirement, introducing his original conception in the therapeutic and surgical field. He created in Cluj-Napoca a specialist medical school by imposing national prestige for the institution that he led.

  14. The Cornell Kitchen: Housing and Design Research in Postwar America.

    Science.gov (United States)

    Penner, Barbara

    2018-01-01

    The Cornell Kitchen (1950-55) was produced at Cornell University by a multidisciplinary team with expertise in home economics, engineering, architecture, and psychology. It promised to deliver rational design, functional principles, aesthetic appeal, and emotional satisfaction in one prefabricated, easy-to-install package. This article sets out the kitchen's history from its design to its field-testing phase to its impact on postwar kitchens. It argues that the kitchen represents an important effort to approach housing in a more scientific way; scientific methods were deployed to understand both the physical and socio-psychological problems of dwelling. The project also sought to introduce a specific model for leveraging housing research into the real world, partnering with industry to mass produce scientific designs. Social scientific methods were hence used to create not only more livable but also more saleable products in an effort to appeal to manufacturers and consumers alike.

  15. Study on treatment of coking wastewater by biofilm reactors combined with zero-valent iron process

    International Nuclear Information System (INIS)

    Lai Peng; Zhao Huazhang; Zeng Ming; Ni Jinren

    2009-01-01

    Experiments were conducted to investigate the behavior of the integrated system with biofilm reactors and zero-valent iron (ZVI) process for coking wastewater treatment. Particular attention was paid to the performance of the integrated system for removal of organic and inorganic nitrogen compounds. Maximal removal efficiencies of chemical oxygen demand (COD), ammonia nitrogen (NH 3 -N) and total inorganic nitrogen (TIN) were up to 96.1, 99.2 and 92.3%, respectively. Moreover, it was found that some phenolic compounds were effectively removed. The refractory organic compounds were primarily removed in ZVI process of the integrated system. These compounds, with molecular weights either ranged 10,000-30,000 Da or 0-2000 Da, were mainly the humic acid (HA) and hydrophilic (HyI) compounds. Oxidation-reduction and coagulation were the main removal mechanisms in ZVI process, which could enhance the biodegradability of the system effluent. Furthermore, the integrated system showed a rapid recovery performance against the sudden loading shock and remained high efficiencies for pollutants removal. Overall, the integrated system was proved feasible for coking wastewater treatment in practical applications

  16. Utilization of the research reactors for the power reactor control instrumentation development

    International Nuclear Information System (INIS)

    Duchene, J.; Verdant, R.; Gilbert, J.

    1977-01-01

    Studies on characteristics and reliability of control instruments lead to testing with various radiations of various intensities and energy spectra. Osiris and Triton reactors present this great variety of radiations and a flexibility of use better than power reactors [fr

  17. Power Excursion Accident Analysis of Research Water Reactor

    International Nuclear Information System (INIS)

    Khaled, S.M.; Doaa, G.M.

    2009-01-01

    A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

  18. UF6 breeder reactor power plants for electric power generation

    International Nuclear Information System (INIS)

    Rust, J.H.; Clement, J.D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

  19. Computer modeling of the dynamic processes in the Maryland University Training Reactor - (MUTR)

    International Nuclear Information System (INIS)

    White, Bernard H. IV; Ebert, David

    1988-01-01

    The simulator described in this paper models the behaviour of the Maryland University Training Reactor (MUTR). The reactor is a 250 kW, TRIGA reactor. The computer model is based on a system of five primary equations and eight auxiliary equations. The primary equations consist of the prompt jump approximation, a heat balance equation for the fuel and the moderator, and iodine and xenon buildup equations. For the comparison with the computer program, data from the reactor was acquired by using a personal computer (pc) which contained a Strawberry Tree data acquisition Card, connected to the reactor. The systems monitored by the pc were: two neutron detectors, fuel temperature, water temperature, three control rod positions and the period meter. The time differenced equations were programmed in the basic language. It has been shown by this paper, that the MUTR power rise from low power critical to high power, can be modelled by a relatively simple computer program. The program yields accurate agreement considering the simplicity of the program. The steady state error between the reactor and computer power is 4.4%. The difference in steady state temperatures, 112 deg. C and 117 deg. C, of the reactor and computer program, respectively, also yields a 4.5% error. Further fine tuning of the coefficients will yield higher accuracies

  20. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power...

  1. Thermionic reactor power conditioner design for nuclear electric propulsion.

    Science.gov (United States)

    Jacobsen, A. S.; Tasca, D. M.

    1971-01-01

    Consideration of the effects of various thermionic reactor parameters and requirements upon spacecraft power conditioning design. A basic spacecraft is defined using nuclear electric propulsion, requiring approximately 120 kWe. The interrelationships of reactor operating characteristics and power conditioning requirements are discussed and evaluated, and the effects on power conditioner design and performance are presented.

  2. Statistic method of research reactors maximum permissible power calculation

    International Nuclear Information System (INIS)

    Grosheva, N.A.; Kirsanov, G.A.; Konoplev, K.A.; Chmshkyan, D.V.

    1998-01-01

    The technique for calculating maximum permissible power of a research reactor at which the probability of the thermal-process accident does not exceed the specified value, is presented. The statistical method is used for the calculations. It is regarded that the determining function related to the reactor safety is the known function of the reactor power and many statistically independent values which list includes the reactor process parameters, geometrical characteristics of the reactor core and fuel elements, as well as random factors connected with the reactor specific features. Heat flux density or temperature is taken as a limiting factor. The program realization of the method discussed is briefly described. The results of calculating the PIK reactor margin coefficients for different probabilities of the thermal-process accident are considered as an example. It is shown that the probability of an accident with fuel element melting in hot zone is lower than 10 -8 1 per year for the reactor rated power [ru

  3. Power generation costs for alternate reactor fuel cycles

    International Nuclear Information System (INIS)

    Smolen, G.R.; Delene, J.G.

    1980-09-01

    The total electric generating costs at the power plant busbar are estimated for various nuclear reactor fuel cycles which may be considered for power generation in the future. The reactor systems include pressurized water reactors (PWR), heavy-water reactors (HWR), high-temperature gas cooled reactors (HTGR), liquid-metal fast breeder reactors (LMFBR), light-water pre-breeder and breeder reactors (LWPR, LWBR), and a fast mixed spectrum reactor (FMSR). Fuel cycles include once-through, uranium-only recycle, and full recycle of the uranium and plutonium in the spent fuel assemblies. The U 3 O 8 price for economic transition from once-through LWR fuel cycles to both PWR recycle and LMFBR systems is estimated. Electric power generation costs were determined both for a reference set of unit cost parameters and for a range of uncertainty in these parameters. In addition, cost sensitivity parameters are provided so that independent estimations can be made for alternate cost assumptions

  4. Nuclear science and engineering education at a university research reactor

    International Nuclear Information System (INIS)

    Loveland, W.

    1993-01-01

    The role of an on-site irradiation facility in nuclear science and engineering education is examined. Using the example of a university research reactor, the use of such devices in laboratory instruction, public outreach programs, special instructional programs, research, etc. is discussed. Examples from the Oregon State University curriculum in nuclear chemistry, nuclear engineering and radiation health are given. (author) 1 tab

  5. The reconstruction of the training reactor of the Budapest Technical University

    International Nuclear Information System (INIS)

    Viragh, E.

    1981-01-01

    The reconstruction of the training reactor between 1978 and 1981 did not hinder the education and training activities of the University. Dosimetric measurements during the test run revealed no additional hazard from the elevation of power from 10 to 100 kW. (author)

  6. Gas-cooled reactor application for a university campus

    International Nuclear Information System (INIS)

    Colak, Ue.; Kadiroghlu, O.K.; Soekmen, C.N.; Schmitt, H.

    1991-01-01

    Large urban areas with unfavourable topographic and meteorological conditions suffer severe air pollution during the winter months. Use of low grade lignites, imported higher quality coal or imported fuel oil are the sources of air pollution in the form of sulphur dioxide, fly ash and soot. Large housing complexes or old and historical locations within the city are in need of pollution free centralized district heating systems. Natural gas imported from the Soviet Union is a solution for this problem. Lack of gas distribution network for high pressure gas within the city is the main bottle-neck for the heating systems utilizing natural gas. Concern of the safety of flammable high pressure gas circulating within the city is another drawback for the natural gas heating systems. Nuclear district heating is an environmentally viable option worth looking into it. Localized urban nuclear heating is an interesting solution for large urban areas with old and historical character. The results of a feasibility study on the HGR application for the Hacettepe University presented here, summarizes the concept of gas-cooled heating reactors specially designed for urban centers. The inherently safe characteristics of the pebble bed heating reactor makes localized urban nuclear heating a viable alternative to other heat sources. An economical analysis of various heat sources with equal power levels is done for the Beytepe campus of Hacettepe University in Ankara. Under special boundary conditions, the price for heat generation can be much lower for nuclear heating with GHR 20 than for hard coal or fuel oil. It is also possible that if the price escalation rate for natural gas exceeds 3%, then nuclear heating with GHR can be more competitive. It is concluded that the nuclear heating of Beytepe campus with a GHR 20 is feasible and economical. (author) 3 figs., 5 refs

  7. Future plans on the Kyoto University Research Reactor (KUR)

    International Nuclear Information System (INIS)

    Shibata, Seiichi

    2000-01-01

    The Research Reactor Institute (RRI), Kyoto University, for aiming at performing the 'Experiments using a reactor and its related research', was established in Showa 38 (1963) as a cooperative research institute for universities and so on in allover Japan. Operation using KUR of one of main facilities in RRI was started by 1 MW of its rated output in 1964, and converted to 5 MW in 1968, after which through development , addition and modification of various research apparatus it has been proposed to the cooperative application researches with universities and so on in allover Japan, hitherto. Among these periods, its research organization is improved to six departments containing twenty divisions and two attached research facilities to progress some investigations on future plans at RRI for response to new researching trends. Here were described on present state of research on use of low concentrated uranium fuels at research reactor, and future plans on neutron factory and hybrid reactor. The former aims at establishment of a new research facility capable of alternating to KUR for future academic research on research reactor containing high quality and high degree application of neutron field and safety management and feature upgrading of nuclear energy. And, the latter aims at development on an accelerator drive uncritical reactor combined an accelerator neutron source and an uncritical reactor. (G.K.)

  8. How power is generated in a nuclear reactor

    International Nuclear Information System (INIS)

    Swaminathan, V.

    1978-01-01

    Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

  9. The concept of the innovative power reactor

    Directory of Open Access Journals (Sweden)

    Sang Won Lee

    2017-10-01

    Full Text Available The Fukushima accident reveals the vulnerability of existing active nuclear power plant (NPP design against prolonged loss of external electricity events. The passive safety system is considered an attractive alternative to cope with this kind of disaster. Also, the passive safety system enhances both the safety and the economics of NPPs. The adoption of a passive safety system reduces the number of active components and can minimize the construction cost of NPPs. In this paper, reflecting on the experience during the development of the APR+ design in Korea, we propose the concept of an innovative Power Reactor (iPower, which is a kind of passive NPP, to enhance safety in a revolutionary manner. The ultimate goal of iPower is to confirm the feasibility of practically eliminating radioactive material release to the environment in all accident conditions. The representative safety grade passive system includes a passive emergency core cooling system, a passive containment cooling system, and a passive auxiliary feedwater system. Preliminary analysis results show that these concepts are feasible with respect to preventing and/or mitigating the consequences of design base accidents and severe accidents.

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

    International Nuclear Information System (INIS)

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

    1972-01-01

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

  11. University of Missouri research reactor exhaust ventilation/laboratory fume hood upgrade

    International Nuclear Information System (INIS)

    Edwards, C.B. Jr.; McKibben, J.C.; McCracken, C.B.

    1989-01-01

    The University of Missouri research reactor (MURR) facility is located in Research Park, 1 mile south of the Columbia campus. The reactor is a 10-MW pressurized loop, in-pool-type, light-water-moderated, beryllium-and-graphite-reflected core, serviced by six radial beam tubes for research, and has sample irradiation facilities in both a flux trap and in the graphite region. The reactor operates at full power 150 h/week, 52 week/yr, making it one of the best operating schedules and the most extensively used of any university research reactor. This extensive utilization includes many programs, such as radioisotope applications, neutron activation analysis, etc., that depend heavily on fume hoods, glove boxes, and hot cells that put a tremendous demand on the exhaust system. The exhaust system is required to be operable whenever the reactor is operating and must have the capability of being operated from an emergency electrical generator on loss of site electrical power. The originally installed exhaust ventilation system was below needed capacity and, with increased program requirements and system age, the necessity to upgrade the system was paramount. The challenge was to complete the upgrade construction while continuing to operate the reactor and maintain all the other ongoing programs, rather than take the easy way of an extended shutdown. This paper discusses how MURR met this challenge and solved these problems, problems that are similarly experienced by almost all research reactors to some degree when major work is required on critical systems

  12. Nuclear Power: Outlook for New U.S. Reactors

    National Research Council Canada - National Science Library

    Parker, Larry; Holt, Mark

    2007-01-01

    .... The renewed interest in nuclear power has resulted primarily from higher prices for natural gas, improved operation of existing reactors, and uncertainty about future restrictions on coal emissions...

  13. Management of radioactive wastes at power reactor sites in India

    International Nuclear Information System (INIS)

    Amalraj, R.V.; Balu, K.

    Indian nuclear power programme, at the present stage, is based on natural uranium fuelled heavy water moderated CANDU type reactors except for the first nuclear power station consisting of two units of enriched uranium fuelled, light water moderated, BWR type of reactors. Some of the salient aspects of radioactive waste management at power reactor sites in India are discussed. Brief reviews are presented on treatment of wastes, their disposal and environmental aspects. Indian experience in power reactor waste management is also summarised identifying some of the areas needing further work. (auth.)

  14. Instrumentation and control for reactor power setback in PFBR

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Vasal, Tanmay; Nagaraj, C.P.; Madhusoodanan, K.

    2013-01-01

    In Prototype Fast Breeder Reactor (PFBR), a 500 MWe plant, Reactor Power Setback is a special operation envisaged for bulk power reduction on occurrence of certain events in Balance of Plant. The bulk power reduction requires a large negative reactivity perturbation if reactor is operating on nominal power. This necessitates a reliable monitoring system with fault tolerant I and C architecture in order to inhibit reactor SCRAM on negative reactivity trip signal. The impact of above events on the process is described. Design of a functional prototype module to carry out RPSB logic operation and its interface with other instruments has been discussed. (author)

  15. CORNELL: CLEO discovers B meson penguins

    International Nuclear Information System (INIS)

    Anon.

    1993-01-01

    The CLEO collaboration at Cornell's CESR electron-positron storage ring has discovered a rare type of B meson decay in which only a high energy photon and a K* meson are produced. These decays provide the first unambiguous evidence for an alternative route for heavy quark decay that has been given the whimsical name ''penguin diagram''. In the mid-1970s penguin diagrams were proposed to explain the puzzling strangeness quantum number selection rules in the decay of K mesons. At the same time it was realized that penguin diagrams could also be important in the CP violation seen in neutral K meson decay. CP violation, an asymmetry between matter and antimatter, is an essential ingredient in understanding why there is much more matter than antimatter in the universe. CP violation introduces a definite direction to the arrow of time, which could otherwise point equally forwards or backwards. In addition, penguin decays are very sensitive to some extensions of the Standard Model of weak decay. Although penguin diagrams were first proposed to explain an effect in K meson decay, the K system gives no unique signature for them, and verification of penguin processes meant looking elsewhere. In the Standard Model, quarks decay under the influence of the weak force, emitting a W boson. Since the W is charged, the charge of the initial quark differs from that of the final quark, so the charge of the quark changes as well as its flavour

  16. CORNELL: CLEO discovers B meson penguins

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1993-06-15

    The CLEO collaboration at Cornell's CESR electron-positron storage ring has discovered a rare type of B meson decay in which only a high energy photon and a K* meson are produced. These decays provide the first unambiguous evidence for an alternative route for heavy quark decay that has been given the whimsical name ''penguin diagram''. In the mid-1970s penguin diagrams were proposed to explain the puzzling strangeness quantum number selection rules in the decay of K mesons. At the same time it was realized that penguin diagrams could also be important in the CP violation seen in neutral K meson decay. CP violation, an asymmetry between matter and antimatter, is an essential ingredient in understanding why there is much more matter than antimatter in the universe. CP violation introduces a definite direction to the arrow of time, which could otherwise point equally forwards or backwards. In addition, penguin decays are very sensitive to some extensions of the Standard Model of weak decay. Although penguin diagrams were first proposed to explain an effect in K meson decay, the K system gives no unique signature for them, and verification of penguin processes meant looking elsewhere. In the Standard Model, quarks decay under the influence of the weak force, emitting a W boson. Since the W is charged, the charge of the initial quark differs from that of the final quark, so the charge of the quark changes as well as its flavour.

  17. Pellet bed reactor for multi-modal space power

    International Nuclear Information System (INIS)

    Buden, D.; Williams, K.; Mast, P.; Mims, J.

    1987-01-01

    A review of forthcoming space power needs for both civil and military missions indicates that power requirements will be in the tens of megawatts. The electrical power requirements are envisioned to be twofold: long-duration lower power levels will be needed for station keeping, communications, and/or surveillance; short-duration higher power levels will be required for pulsed power devices. These power characteristics led to the proposal of a multi-modal space power reactor using a pellet bed design. Characteristics desired for such a multimegawatt reactor power source are standby, alert, and pulsed power modes; high-thermal output heat source (approximately 1000 MWt peak power); long lifetime station keeping power (10 to 30 years); high temperature output (1500 K to 1800 K); rapid-burst power transition; high reliability (above 95 percent); and stringent safety standards compliance. The proposed pellet bed reactor is designed to satisfy these characteristics

  18. Lens Systems Incorporating A Zero Power Corrector Objectives And Magnifiers For Night Vision Applications

    Science.gov (United States)

    McDowell, M. W.; Klee, H. W.

    1986-02-01

    The use of the zero power corrector concept has been extended to the design of objective lenses and magnifiers suitable for use in night vision goggles. A novel design which can be used as either an f/1.2 objective or an f/2 magnifier is also described.

  19. Gas separation membranes for zero-emission fossil power plants: MEM-BRAIN

    NARCIS (Netherlands)

    Czyperek, M.; Zapp, P.; Bouwmeester, Henricus J.M.; Modigell, M.; Ebert, K.; Voigt, I.; Meulenberg, W.A.; Singheiser, L.; Stöver, D.

    2010-01-01

    The objective of the “MEM-BRAIN” project is the development and integration of ceramic and polymeric gas separation membranes for zero-emission fossil power plants. This will be achieved using membranes with a high permeability and selectivity for either CO2, O2 or H2, for the three CO2 capture

  20. An approach of raising the low power reactor trip block (P-7) in Maanshan Power Plant

    International Nuclear Information System (INIS)

    Wang, L.C.

    1984-01-01

    The technical specification for the Maanshan Nuclear Power Station (FSAR Table 16.2.2-3) requires that with an increasing reactor power level above the setpoint of low power reactor trip block (P-7), a turbine trip shall initiate a reactor trip. This anticipatory reactor trip on turbine trip prevents the pressurizer PORV from openning during turbine trip event. In order to reduce unnecessary reactor trip due to turbine trip on low reactor power level during Maanshan start-up stage, Taiwan Power Company performed a transient analysis for turbine trip event by using RETRAN code. The highest reactor power level at which a turbine trip will not open the pressurizer PORV is searched. The results demonstrated that this power level can be increased from the original value-10% of the rated thermal power-to about 48% of the rated thermal power

  1. Universal Fast Breeder Reactor Subassembly Counter manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications

  2. Universal Fast Breeder Reactor Subassembly Counter manual

    Energy Technology Data Exchange (ETDEWEB)

    Menlove, H.O.; Eccleston, G.W.; Swansen, J.E.; Goris, P.; Abedin-Zadeh, R.; Ramalho, A.

    1984-08-01

    A neutron coincidence counter has been designed for the measurement of fast breeder reactor fuel assemblies. This assay system can accommodate the full range of geometries and masses found in fast breeder subassemblies under IAEA safeguards. The system's high-performance capability accommodates high plutonium loadings of up to 16 kg. This manual describes the system and its operation and gives performance and calibration parameters for typical applications.

  3. Design and analysis on super-critical water cooled power reactors

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki

    2005-01-01

    The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

  4. Reactor G1: high power experiments

    International Nuclear Information System (INIS)

    Laage, F. de; Teste du Baillet, A.; Veyssiere, A.; Wanner, G.

    1957-01-01

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

  5. Calculation of research reactor RA power at uncontrolled reactivity changes

    International Nuclear Information System (INIS)

    Cupac, S.

    1978-01-01

    The safety analysis of research reactor RA involves also the calculation of reactor power at uncontrolled reactivity changes. The corresponding computer code, based on Point Kinetics Model has been made. The short review of method applied for solving kinetic equations is given and several examples illustrating the reactor behaviour at various reactivity changes are presented. The results already obtained are giving rather rough picture of reactor behaviour in considered situations. This is the consequence of using simplified feed back and reactor cooling models, as well as temperature reactivity coefficients, which do not correspond to the actual reactor RA structure (which is now only partly fulfilled with 80% enriched uranium fuel). (author) [sr

  6. Nuclear power reactors in the world. Apr 1985 ed.

    International Nuclear Information System (INIS)

    1985-01-01

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

  7. Design studies of Tokamak power reactor in JAERI

    International Nuclear Information System (INIS)

    Tone, T.; Nishikawa, M.; Tanaka, Y.

    1985-01-01

    Recent design studies of tokamak power reactor and related activities conducted in JAERI are presented. A design study of the SPTR (Swimming-Pool Type Reactor) concept was carried out in FY81 and FY82. The reactor design studies in the last two years focus on nuclear components, heat transport and energy conversion systems. In parallel of design studies, tokamak systems analysis code is under development to evaluate reactor performances, cost and net energy balance

  8. Calculation of power density with MCNP in TRIGA reactor

    International Nuclear Information System (INIS)

    Snoj, L.; Ravnik, M.

    2006-01-01

    Modern Monte Carlo codes (e.g. MCNP) allow calculation of power density distribution in 3-D geometry assuming detailed geometry without unit-cell homogenization. To normalize MCNP calculation by the steady-state thermal power of a reactor, one must use appropriate scaling factors. The description of the scaling factors is not adequately described in the MCNP manual and requires detailed knowledge of the code model. As the application of MCNP for power density calculation in TRIGA reactors has not been reported in open literature, the procedure of calculating power density with MCNP and its normalization to the power level of a reactor is described in the paper. (author)

  9. Power limit and quality limit of natural circulation reactor

    International Nuclear Information System (INIS)

    Zhao Guochang; Ma Changwen

    1997-01-01

    The circulation characteristics of natural circulation reactor in boiling regime are researched. It is found that, the circulation mass flow rate and the power have a peak value at a mass quality respectively. Therefore, the natural circulation reactor has a power limit under certain technological condition. It can not be increased steadily by continually increasing the mass quality. Corresponding to this, the mass quality of natural circulation reactor has a reasonable limit. The relations between the maximum power and the reactor parameters, such as the resistance coefficient, the working pressure and so on, are analyzed. It is pointed out that the power limit of natural circulation reactor is about 1000 MW at present technological condition. Taking the above result and low quality stability experimental result into account, the authors recommend that the reasonable mass quality of natural circulation reactor working in boiling regime is from 2% to 3% under the researched working pressure

  10. Assessment of nuclear reactor concepts for low power space applications

    Science.gov (United States)

    Klein, Andrew C.; Gedeon, Stephen R.; Morey, Dennis C.

    1988-01-01

    The results of a preliminary small reactor concepts feasibility and safety evaluation designed to provide a first order validation of the nuclear feasibility and safety of six small reactor concepts are given. These small reactor concepts have potential space applications for missions in the 1 to 20 kWe power output range. It was concluded that low power concepts are available from the U.S. nuclear industry that have the potential for meeting both the operational and launch safety space mission requirements. However, each design has its uncertainties, and further work is required. The reactor concepts must be mated to a power conversion technology that can offer safe and reliable operation.

  11. Calibration of RB reactor power; Kalibrisanje snage reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

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

    1976-09-15

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

  12. Calculations for accidents in water reactors during operation at power

    International Nuclear Information System (INIS)

    Blanc, H.; Dutraive, P.; Fabrega, S.; Millot, J.P.

    1976-07-01

    The behaviour of a water reactor on an accident occurring as the reactor is normally operated at power may be calculated through the computer code detailed in this article. Reactivity accidents, loss of coolant ones and power over-running ones are reviewed. (author)

  13. Comparison of Pickering NGS performance with world power reactors, 1977

    International Nuclear Information System (INIS)

    Buhay, S.

    Pickering NGS performance is compared, in highly graphic form, with the perfomance of other nuclear power plants around the world. The four Pickering reactors score in the top six, rated by gross capacity factor. Major system suppliers for world power reactors above 500 MW are cataloged. (E.C.B.)

  14. VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

    International Nuclear Information System (INIS)

    Matejka, K.; Sklenka, L.

    2003-01-01

    The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

  15. Zero-dimensional model of a reversed field pinch fusion reactor

    International Nuclear Information System (INIS)

    Veerasingam, R.

    1987-12-01

    A zero-dimensional model for the energy balance for electrons and ions of a Reversed Field Pinch (RFP) device has been developed. The model can be used as a tool for parametric studies and has been applied to simulate some ZT-40M experiments. In the model multiplicative coefficients C 1 , C 2 , C 3 and C 4 are introduced to treat anomalous resistivity, electron and ion energy confinement times and to account for the instability driven ion heating that is observed in RFP experiments. Parametric studies were performed to determine the sensitivity of the model to changes in multiplicative coefficients. A set of coefficients which can simulate a number of ZT-40M experiments have been obtained and the sensitivity of T/sub e/ and T/sub i/ to these coefficients was examined. Both flat top and ramp current waveforms were studied. The effects of different levels of impurities were also examined. The results showed that while all the three impurities used, viz., carbon, oxygen and nickel contributed to Z/sub eff/, nickel dominated the impurity radiation power. The results were then applied to study the behavior of the ZT-H device which is being built in the Confinement Physics Research Facility at the Los Alamos National Laboratory. 30 refs., 39 figs., 12 tabs

  16. Education and training activities at North Carolina State University's PULSTAR reactor

    International Nuclear Information System (INIS)

    Mayo, C.W.

    1992-01-01

    Research reactor utilization has been an integral part of the North Carolina State University's (NCSU's) nuclear engineering program since its inception. The undergraduate curriculum has a strong teaching laboratory component. Graduate classes use the reactor for selected demonstrations, experiments, and projects. The reactor is also used for commercial power reactor operator training programs, neutron radiography, neutron activation analysis (NAA), and sample and tracer activation for industrial short courses and services as part of the university's land grant mission. The PULSTAR reactor is a 1-MW pool-type reactor that uses 4% enriched UO 2 pellet fuel in Zircaloy II cladding. Standard irradiation facilities include wet exposure ports, a graphite thermal column, and a pneumatic transfer system. In the near term, general facility upgrades include the installation of signal isolation and computer data acquisition and display functions to improve the teaching and research interface with the reactor. In the longer term, the authors foresee studies of new core designs and the development of beam experiment design tools. These would be used to study modifications that may be desired at the end of the current core life and to undertake the development of new research instruments

  17. Axial power monitoring uncertainty in the Savannah River Reactors

    International Nuclear Information System (INIS)

    Losey, D.C.; Revolinski, S.M.

    1990-01-01

    The results of this analysis quantified the uncertainty associated with monitoring the Axial Power Shape (APS) in the Savannah River Reactors. Thermocouples at each assembly flow exit map the radial power distribution and are the primary means of monitoring power in these reactors. The remaining uncertainty in power monitoring is associated with the relative axial power distribution. The APS is monitored by seven sensors that respond to power on each of nine vertical Axial Power Monitor (APM) rods. Computation of the APS uncertainty, for the reactor power limits analysis, started with a large database of APM rod measurements spanning several years of reactor operation. A computer algorithm was used to randomly select a sample of APSs which were input to a code. This code modeled the thermal-hydraulic performance of a single fuel assembly during a design basis Loss-of Coolant Accident. The assembly power limit at Onset of Significant Voiding was computed for each APS. The output was a distribution of expected assembly power limits that was adjusted to account for the biases caused by instrumentation error and by measuring 7 points rather than a continuous APS. Statistical analysis of the final assembly power limit distribution showed that reducing reactor power by approximately 3% was sufficient to account for APS variation. This data confirmed expectations that the assembly exit thermocouples provide all information needed for monitoring core power. The computational analysis results also quantified the contribution to power limits of the various uncertainties such as instrumentation error

  18. Liquid-poison type power controlling device for nuclear reactor

    International Nuclear Information System (INIS)

    Horiuchi, Tetsuo; Yamanari, Shozo; Sugisaki, Toshihiko; Goto, Hiroshi.

    1981-01-01

    Purpose: To improve the safety and the operability of a nuclear reactor by adjusting the density of liquid poison. Constitution: The thermal expansion follow-up failure between cladding and a pellet upon abrupt and local variations of the power is avoided by adjusting the density of liquid poison during ordinary operation in combination with a high density liquid poison tank and a filter and smoothly controlling the reactor power through a pipe installed in the reactor core. The high density liquid poison is abruptly charged in to the reactor core under relatively low pressure through the tube installed in the reactor core at the time of control rod insertion failure in an accident, thereby effectively shutting down the reactor and improving the safety and the operability of the reactor. (Yoshihara, H.)

  19. Nuclear science and engineering education at a university research reactor

    International Nuclear Information System (INIS)

    Loveland, W.

    1990-01-01

    The research and teaching operations of the Nuclear Chemistry Division of the Dept. of Chemistry and the Dept. of Nuclear Engineering are housed at the Oregon State University Radiation Center. This facility which includes a 1.1 MW TRIGA reactor was used for 53 classes from a number of different academic departments last year. About one-half of these classes used the reactor and ∼25% of the reactor's 45 hour week was devoted to teaching. Descriptions will be given of reactor-oriented instructional programs in nuclear engineering, radiation health and nuclear chemistry. In nuclear chemistry, classes in (a) nuclear chemistry for nuclear engineers, (b) radiotracer methods, (c) elementary and advanced activation analysis, and (d) advanced nuclear instrumentation will be described in detail. The use of the facility to promote general nuclear literacy among college students, high school and grade school students and the general population will also be covered

  20. Reconstruction calculation of pin power for ship reactor core

    International Nuclear Information System (INIS)

    Li Haofeng; Shang Xueli; Chen Wenzhen; Wang Qiao

    2010-01-01

    Aiming at the limitation of the software that pin power distribution for ship reactor core was unavailable, the calculation model and method of the axial and radial pin power distribution were proposed. Reconstruction calculations of pin power along axis and radius was carried out by bicubic and bilinear interpolation and cubic spline interpolation, respectively. The results were compared with those obtained by professional reactor physical soft with fine mesh difference. It is shown that our reconstruction calculation of pin power is simple and reliable as well as accurate, which provides an important theoretic base for the safety analysis and operating administration of the ship nuclear reactor. (authors)

  1. Zero-discharge wastewater treatment facility for a 900-MWe GCC power plant

    International Nuclear Information System (INIS)

    Rosain, R.M.; Dalan, J.A.

    1992-05-01

    Florida Power and Light desires to examine the prospect of achieving zero liquid discharge from the gasification area of their proposed 900-MW coal gasification-combined cycle (GCC) power plant expansion at the Martin station. This report provides information about the technologies available, cost, and process selection methods, and recommends a preferred system for achieving zero liquid discharge from the gasification block. The recommended system consists of primary clarification and vapor compression evaporation, followed by carbon adsorption post-treatment of the evaporator distillate. Dry solids are produced from the evaporator concentrate with a crystallizer/centrifuge combination. The system recovers 99 percent of the wastewater as pure distillate vater. The predicted capital cost for the 265-gpm system is $12.5 million; the predicted operating costs are $18.60/1000 gallons. Both costs are in 1990 dollars. Promising treatment technologies to examine for future designs are cooling tower treatment and freeze crystallization

  2. Optimization of Nuclear Reactor power Distribution using Genetic Algorithm

    International Nuclear Information System (INIS)

    Kim, Hyu Chan

    1996-02-01

    The main purpose of study is to develop a computer code named as 'MGA-SCOUPE' which can determine an optimal fuel-loading pattern for the nuclear reactor. The developed code, MGA-SCOUPE, automatically lots of searches for the globally optimum solutions based upon the modified Genetic Algorithm(MGA). The optimization goal of the MGA-SCOUPE is (1) the minimization of the deviations in the power peaking factors both at BOC and EOC, and (2) the maximization of the average burnup ration at EOC of the total fuel assemblies. For the reactor core calculation module in the MGA-SCOUPE, the SCOUPE code was partially modified and used. It had been developed originally in MIT and has been used currently in Kyung Hee University. The application of the MGA-SCOUPE to KORI 4-4 Cycle Model show several satisfactory results. Among them, two dominant improvements compared with the SCOUPE code can be summarized as follow: - The MGA-SCOUPE removes the user-dependency problem of the SCOUPE in the optimal loading pattern searches. Therefore, the searching process in the MGA-SCOUPE can be easily automated. - The final fuel loading pattern obtained by the MGA-SCOUPE shows 25.8%, 18.7% reduced standard deviations of the power peaking factors both at BOC and EOC, and 45% increased avg. burnup ratio at EOC compare with those of the SCOUPE

  3. Fractal characterization for noise signal validation in power reactors

    International Nuclear Information System (INIS)

    Aguilar Martinez, Omar

    2003-01-01

    Up to now, a great variety of methods is used for the dynamical characterization of different components of Nuclear Power Plants (NPPs). With this aim, time and spectral analysis are usually considered, and different tools of non-stationary and non-gaussian analysis are also presented. When applying non-lineal dynamics theory for noise signal validation purposes in power reactors, the extraction of fractal echoes plays a main role. Fractal characterization for noise signal validation purposes can be integrated to the task of processing and acquisition of time signals in noise (fluctuation parameters) analysis systems. The possibility of discrimination between deterministic chaotic signals and pure noise signals has been incorporated, as a complement; to noise signals analysis in normal and anomalous operational conditions in NPPs using a fractal approach. In this work the detailed analysis of a neutronic sensor response is considered and the fractal characterization of its dynamics state (i.e. sensor line) for noise signal classification, it is presented. The experiment from where the time series (signals) were obtained, was carried out at the Research Reactor of the Technical University of Budapest, Hungary, during a model experiment for ageing process study of in-core neutron detectors (author)

  4. Application of Cherenkov light observation to reactor measurements (1). Estimation of reactor power from Cherenkov light intensity

    International Nuclear Information System (INIS)

    Yamamoto, Keiichi; Takeuchi, Tomoaki; Kimura, Nobuaki; Ohtsuka, Noriaki; Tsuchiya, Kunihiko; Sano, Tadafumi; Nakajima, Ken; Homma, Ryohei; Kosuge, Fumiaki

    2015-01-01

    Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. The calculation and the observation were performed to obtain the quantity of the Cherenkov light caused by the gamma and beta rays emitted from the fuels in the core of Kyoto University Research Reactor. The results indicate that the real-time reactor power can be estimated from the brightness of the Cherenkov light observed by a CCD camera. This method can also work for the estimation of the burn-up of spent fuels at commercial reactors. Since the observed brightness value of the Cherenkov light was influenced by the camera position, the optical observation method should be improved to achieve high accuracy observation. (author)

  5. Nuclear reactor power control system based on flexibility model

    International Nuclear Information System (INIS)

    Li Gang; Zhao Fuyu; Li Chong; Tai Yun

    2011-01-01

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

  6. Simulation of decreasing reactor power level with BWR simulator

    International Nuclear Information System (INIS)

    Suwoto; Zuhair; Rivai, Abu Khalid

    2002-01-01

    Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods

  7. Music from Ground Zero. The coloniality of theory and musical analysis at the university.

    Directory of Open Access Journals (Sweden)

    Pilar Jovanna Holguín

    2018-01-01

    Full Text Available This article seeks to answer the influence of the coloniality of knowledge, the cleansing of blood and the hybris of the zero ground in the conception of the theory and musical analysis which is taught in our Latin American universities. The aim is to examine the survival of coloniality in the imaginary of musical theory and analysis in order to recognize the predominant hegemonic discourses and propose some options to decolonize these disciplines of professional music education. The rationale is based on the presentation of some concepts of decolonial studies, research on the conservatory model and the ideology of theory and analysis.The article is divided into three parts. In the first one, an approximation is made to the concept of coloniality and its categories. The second takes the categories to review the ideology of the chosen fields of musical knowledge and the third proposes some options to decolonize our conceptions in higher education.

  8. Minority and female training programs at the Ford Nuclear Reactor, University of Michigan

    International Nuclear Information System (INIS)

    Burn, R.R.

    1992-01-01

    Nuclear power industry operations staffs are composed predominantly of white males because most of the personnel come from the nuclear submarine and surface branches of the U.S. Navy. The purpose of the minority and female training programs sponsored by the Ford Nuclear Reactor at the University of Michigan is to provide a path for minorities and women to enter the nuclear industry as operators, technicians, and, in the long term, as graduate engineers. The training programs are aimed at high school students, preferably juniors. While the training is directed toward operation of a nuclear reactor, it is equally applicable to careers in most other technical fields. It is hoped that some of the participants will remain at the Ford Nuclear Reactor as reactor operators, enter college, and obtain college degrees, after which they will enter the nuclear industry as graduate engineers

  9. Power start up of the Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

    1994-01-01

    After accomplishing the physical start-up of the reactor, the power start-up was carried out in February 1984. The power of the reactor has reached: 10 KW on 6/2/1984, 100 KW on 7/2/1984, 200 KW and 300 KW on 8/2/1984; 400 KW and nominal power 500 KW on 9/2/1984. The reactivity temperature coefficient and the xenon poisoning were determined. 3 figs., 12 tabs

  10. Nonlinear analysis on power reactor dynamics

    International Nuclear Information System (INIS)

    Konno, H.; Hayashi, K.

    1997-01-01

    We have shown that the origin of intermittent oscillation observed in a BWR can be ascribed to the couplings among the spatial modes starting from a non-linear center manifold equation with a delay-time and a spatial diffusion. We can reduce the problem to the stochastic coupled van der Pol oscillators with non-linear coupling term. This non-linear coupling term plays an important role to break the symmetry of the system and the non-linear damping of the system. The phenomenological generalization of van der Pol oscillator coupled by the linear diffusion term is not appropriate for describing the nuclear power reactors. However, one must start from the coupled partial differential equations by taking into account the two energy group neutrons, the thermo-hydraulic equations including two-phase flow. In this case, the diffusion constant must be a complex number as is demonstrated in a previous paper. The results will be reported in the near future. (J.P.N.)

  11. Nuclear Power Reactors in the World. 2013 Ed

    International Nuclear Information System (INIS)

    2013-01-01

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

  12. Nuclear Power Reactors in the World. 2014 Ed

    International Nuclear Information System (INIS)

    2014-01-01

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

  13. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    International Nuclear Information System (INIS)

    Mesquita, Henrique F.A.; Ferreira, Andrea V.

    2015-01-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  14. Power reactor noise measurements in Hungary

    International Nuclear Information System (INIS)

    Pallagi, D.; Horanyi, S.; Hargitai, T.

    1975-01-01

    An outline is given of the history of reactor noise research in Hungary. A brief description is given of studies in the WWR-SM reactor, a modified version of the original WWR-S thermal reactor, for the detection of in-core simulated boiling by analysis of the noise of out-of-core ionization chambers. Coolant velocity measurements by transit time analysis of temperature fluctuations are described. (U.K.)

  15. Design and construction of small power reactors

    International Nuclear Information System (INIS)

    Tachi, Yasuo

    1992-01-01

    Small size reactors are considered to have many advantages over large-sized reactors. But at the same time, small size reactors show eventual disadvantages in economy. In this paper one of the possibilities to improve its basic disadvantage will be discussed from a manufacturer's point of view. The stress will be placed on the possibility and possible effects of adoption of Computer Aided Engineering. (author). 2 figs

  16. MEASUREMENT ERROR EFFECT ON THE POWER OF CONTROL CHART FOR ZERO-TRUNCATED POISSON DISTRIBUTION

    Directory of Open Access Journals (Sweden)

    Ashit Chakraborty

    2013-09-01

    Full Text Available Measurement error is the difference between the true value and the measured value of a quantity that exists in practice and may considerably affect the performance of control charts in some cases. Measurement error variability has uncertainty which can be from several sources. In this paper, we have studied the effect of these sources of variability on the power characteristics of control chart and obtained the values of average run length (ARL for zero-truncated Poisson distribution (ZTPD. Expression of the power of control chart for variable sample size under standardized normal variate for ZTPD is also derived.

  17. Safety studies concerning nuclear power reactors

    International Nuclear Information System (INIS)

    Bailly, Jean; Pelce, Jacques

    1980-01-01

    The safety of nuclear installations poses different technical problems, whether concerning pressurized water reactors or fast reactors. But investigating methods are closely related and concern, on the one hand, the behavior of shields placed between fuel and outside and, on the other, analysis of accidents. The article is therefore in two parts based on the same plan. Concerning light water reactors, the programme of studies undertaken in France accounts for the research carried out in countries where collaboration agreements exist. Concerning fast reactors, France has the initiative of their studies owing to her technical advance, which explains the great importance of the programmes under way [fr

  18. Utilization of research reactors in universities and their medical applications

    International Nuclear Information System (INIS)

    Kanda, Keiji.

    1983-01-01

    In Japan, five research reactors and a critical assembly are operated by the universities. They are opened to all university researchers, the system of which is financially supported by the Ministry of Education, Culture and Science of the Japanese government. Usually KUR is operated eight cycles per year. One cycle consists of the following four week operation: 1. Mainly for researchers from other universities; 2. Mainly for researchers in the institute; 3. Mainly for beam experiment; 4. Sort time experiment. In the weeks of 1 ∼ 3 the KUR is operated continously from Tuesday morning to Friday evening. The experiment include studies on physics, chemistry, biology, medicine, engineering etc. Recently the medical application of research reactors has become popular in Japan. The new technique of the boron neutron capture thereby has been successfully applied to brain tumors and will be to melanoma (skin cancer) in near future. (author)

  19. A method of reactor power decrease by 2DOF control system during BWR power oscillation

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Suzuki, Katsuo

    1998-09-01

    Occurrence of power oscillation events caused by void feedback effects in BWRs operated at low-flow and high-power condition has been reported. After thoroughly examining these events, BWRs have been equipped with the SRI (Selected Rod Insertion) system to avoid the power oscillation by decreasing the power under such reactor condition. This report presents a power control method for decreasing the reactor power stably by a two degree of freedom (2DOF) control. Performing a numerical simulation by utilizing a simple reactor dynamics model, it is found that the control system designed attains a satisfactory control performance of power decrease from a viewpoint of setting time and oscillation. (author)

  20. Use of university research reactors to teach control engineering

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1991-01-01

    University research reactors (URRs) have provided generations of students with the opportunity to receive instruction and do hands-on work in reactor dynamics, neutron scattering, health physics, and neutron activation analysis. Given that many URRs are currently converting to programmable control systems, the opportunity now exists to provide a similar learning experience to those studying systems control engineering. That possibility is examined here with emphasis on the need for the inclusion of experiment in control engineering curricula, the type of activities that could be performed, and safety considerations

  1. Status of the University of Missouri-Columbia Research Reactor upgrade

    Energy Technology Data Exchange (ETDEWEB)

    McKibben, J C; Edwards, Jr, C B; Meyer, Jr, W A [MU Research Reactor, Columbia, MO (United States); Kim, S S [Idaho Nuclear Engineering Laboratory, Idaho Falls, ID (United States)

    1990-05-01

    The University of Missouri-Columbia (MU) Research Reactor Facility staff is in the process of upgrading the operational and research capabilities of the reactor and associated facilities. The upgrades include an extended life aluminide fuel element, a power increase, improved instrumentation and control equipment, a cold neutron source, a building addition, and improved research instrumentation and equipment. These upgrades will greatly enhance the capabilities of the facility and the research programs. This paper discusses the parts of the upgrade and current status of implementation. (author)

  2. Status of the University of Missouri-Columbia Research Reactor upgrade

    International Nuclear Information System (INIS)

    McKibben, J.C.; Edwards, C.B. Jr.; Meyer, W.A. Jr.; Kim, S.S.

    1990-01-01

    The University of Missouri-Columbia (MU) Research Reactor Facility staff is in the process of upgrading the operational and research capabilities of the reactor and associated facilities. The upgrades include an extended life aluminide fuel element, a power increase, improved instrumentation and control equipment, a cold neutron source, a building addition, and improved research instrumentation and equipment. These upgrades will greatly enhance the capabilities of the facility and the research programs. This paper discusses the parts of the upgrade and current status of implementation. (author)

  3. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    Vathaire, F. de; Vernier, Ph.; Pascouet, A.

    1964-01-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

  4. Damage of fuel assembly premature changing in a power reactor

    International Nuclear Information System (INIS)

    Rudik, A.P.

    1987-01-01

    Material balance, including energy recovery and nuclear fuel flow rate, under conditions of premature FA extraction from power reactor is considered. It is shown that in cases when before and after FA extraction reactor operates not under optimal conditions damage of FA premature changing is proportional to the first degree of fuel incomplete burning. If normal operating conditions of reactor or its operation after FA changing is optimal, the damage is proportional to the square of fuel incomplete burning

  5. The next generation of power reactors - safety characteristics

    International Nuclear Information System (INIS)

    Modro, S.M.

    1995-01-01

    The next generation of commercial nuclear power reactors is characterized by a new approach to achieving reliability of their safety systems. In contrast to current generation reactors, these designs apply passive safety features that rely on gravity-driven transfer processes or stored energy, such as gas-pressurized accumulators or electric batteries. This paper discusses the passive safety system of the AP600 and Simplified Boiling Water Reactor (SBWR) designs

  6. Radiation embrittlement in pressure vessels of power reactors

    International Nuclear Information System (INIS)

    Kempf, Rodolfo; Fortis, Ana M.

    2007-01-01

    It is presented the project to study the effect of lead factors on the mechanical behavior of Reactor Pressure Vessel steels. It is described the facility designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. The objective is to obtain the fracture behavior of irradiated specimens with different lead factors and to know their dependence with the diffusion of alloy elements. (author) [es

  7. Nuclear power reactors in the world. April 2005 ed

    International Nuclear Information System (INIS)

    2005-01-01

    This is the twenty-fifth edition of Reference Data Series No. 2, Nuclear Power Reactors in the World, which is published once per year, to present the most recent reactor data available to the Agency. It contains the following summarized information: - General information as of the end of 2004 on power reactors operating or under construction, and shut down; - Performance data on reactors operating in the Agency's Member States, as reported to the IAEA. The information is collected by the Agency by circulating questionnaires to Member States through the designated national correspondents. The replies are used to maintain computerized files on general and design data of, and operating experience with, power reactors. The Agency's Power Reactor Information System (PRIS) comprising the above files provides all the information and data previously published in the Agency's Power Reactors in Member States and currently published in the Agency's Operating Experience with Nuclear Power Stations in Member States and available at the Internet address http://www.iaea.org/programmes/a2

  8. Status report on nuclear reactors for space electric power

    International Nuclear Information System (INIS)

    Buden, D.

    1978-01-01

    The Los Alamos Scientific Laboratory is studying reactor power plants for space applications in the late 1980s and 1990s. The study is concentrating on high-temperature, compact, fast reactors that can be coupled with various radiation shielding systems and thermoelectric, dynamic, or thermionic electric power conversion systems, depending on the mission. Increased questions have been raised about safety since the COSMOS 954 incident. High orbits (above 400 to 500 nautical miles) have sufficient lifetimes to allow radioactive elements to decay to safe levels. The major proposed applications for satellites with reactors in Earth orbit are in geosynchronous orbit (19,400 nautical miles). In missions at geosynchronous orbit where orbital lifetimes are practically indefinite, the safety considerations are negligible. The potential missions, why reactors are being considered as a prime power candidate, reactor features, and safety considerations are discussed

  9. Calculation device for fuel power history in BWR type reactors

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To enable calculations for power history and various variants of power change in the power history of fuels in a BWR type reactor or the like. Constitution: The outputs of the process computation for the nuclear reactor by a process computer are stored and the reactor core power distribution is judged from the calculated values for the reactor core power distribution based on the stored data. Data such as for thermal power, core flow rate, control rod position and power distribution are recorded where the changes in the power distribution exceed a predetermined amount, and data such as for thermal power and core flow rate are recorded where the changes are within the level of the predetermined amount, as effective data excluding unnecessary data. Accordingly, the recorded data are taken out as required and the fuel power history and the various variants in the fuel power are calculated and determined in a calculation device for fuel power history and variants for fuel power fluctuation. (Furukawa, Y.)

  10. Containment vessel construction for nuclear power reactors

    International Nuclear Information System (INIS)

    Sulzer, H.D.; Coletti, J.L.

    1975-01-01

    A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

  11. Nuclear Power Reactors in the World. 2016 Ed

    International Nuclear Information System (INIS)

    2016-01-01

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

  12. Ultrasonic level and temperature sensor for power reactor applications

    International Nuclear Information System (INIS)

    Dress, W.B.; Miller, G.N.

    1983-01-01

    An ultrasonic waveguide employing torsional and extensional acoustic waves has been developed for use as a level and temperature sensor in pressurized and boiling water nuclear power reactors. Features of the device include continuous measurement of level, density, and temperature producing a real-time profile of these parameters along a chosen path through the reactor vessel

  13. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  14. Small size modular fast reactors in large scale nuclear power

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  15. Conceptual Study for development of a low power research reactor

    International Nuclear Information System (INIS)

    Park, C.; Kim, H. S.; Park, J. H.; Chae, H. T.; Lee, B. C.

    2013-01-01

    Even though the nuclear society is again facing with difficult situations after Fukusima accident, some countries still continues to consider nuclear power as one option of national energy sources and to introduce nuclear energy. As a research reactor has been regarded as a step-stone to establish infrastructures for the nuclear power development program, some countries that have plan to introduce the nuclear power energy are considering to construct a research reactor. Particularly, a low power research reactor whose main purpose is basic researches on the nuclear technology and education/training would be of interest to developing countries when taking the economy and level of science and technology into consideration. And many low power research reactors at operation are obsolescent and their numbers are decreasing. Hence, some concepts on a low power research reactor are being studied for the future needs. This paper presents the conceptual study on the basic requirements and the preliminary design features of a low power research reactor

  16. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  17. Power from plutonium: fast reactor fuel

    International Nuclear Information System (INIS)

    Bishop, J.F.W.

    1981-01-01

    Points of similarity and of difference between fast reactor fuel and fuels for AGR and PWR plants are established. The flow of uranium and plutonium in fast and thermal systems is also mentioned, establishing the role of the fast reactor as a plutonium burner. A historical perspective of fast reactors is given in which the substantial experience accumulated in test and prototype is indicated and it is noted that fast reactors have now entered the commercial phase. The relevance of the data obtained in the test and prototype reactors to the behaviour of commercial fast reactor fuel is considered. The design concepts employed in fuel are reviewed, including sections on core support styles, pin support and pin detail. This is followed by a discussion of current issues under the headings of manufacture, performance and reprocessing. This section includes a consideration of gel fuel, achievable burn-up, irradiation induced distortions and material choices, fuel form, and fuel failure mechanisms. Future development possibilities are also discussed and the Paper concludes with a view on the logic of a UK fast reactor strategy. (U.K.)

  18. Maintainability considerations for the central cell in WITAMIR-I, a conceptual design of a tandem mirror fusion power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.N.

    1980-10-01

    The concepts for maintaining the central cell reactor components for WITAMIR-I are described. WITAMIR-I is a conceptual tandem mirror fusion power reactor utilizing thermal barriers designed by the University of Wisconsin-Madison. Unique solutions to the difficult problems of routine blanket replacement and maintenance are proposed. Solutions are also proposed for maintaining the central cell coils and the shield

  19. Reactor shutdown: nuclear power plant performance

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The article essentially looks at the performance of nine of Sweden's nuclear reactors. A table lists the percentage of time for the first three quarters of 1981 that the reactors were operating, and the number of hours out of service for planned or other reasons. In particular, one station - Ringhals 3 - was out of action because of a damaged tube in the associated steam generator. The same fault occurred with another reactor - Ringhals 4 - before this was brought into service. The reasons for the failure and its importance are briefly discussed. (G.P.)

  20. Regulations of local choices for installation of power reactors

    International Nuclear Information System (INIS)

    1969-09-01

    The present regulations specify the criteria under which the Comissao Nacional de Energia Nuclear will approve the local proposed for the installation of power reactors, according to his attributions established in the Law 4118, dated of August 27, 1962

  1. Nuclear Power: Outlook for New U.S. Reactors

    National Research Council Canada - National Science Library

    Parker, Larry; Holt, Mark

    2007-01-01

    Nearly three decades after the most recent order was placed for a new nuclear power plant in the United States, several utilities are now expressing interest in building a total of up to 30 new reactors...

  2. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  3. Thermophotovoltaic Energy Conversion in Space Nuclear Reactor Power Systems

    National Research Council Canada - National Science Library

    Presby, Andrew L

    2004-01-01

    .... This has potential benefits for space nuclear reactor power systems currently in development. The primary obstacle to space operation of thermophotovoltaic devices appears to be the low heat rejection temperatures which necessitate large radiator areas...

  4. Preliminary Study of 20 MWth Experiment Power Reactor based on Pebble Bed Reactor

    Science.gov (United States)

    Irwanto, Dwi; Permana, Sidik; Pramuditya, Syeilendra

    2017-07-01

    In this study, preliminary design calculations for experimental small power reactor (20 MWt) based on Pebble Bed Reactor (PBR) are performed. PBR technology chosen due to its advantages in neutronic and safety aspects. Several important parameters, such as fissile enrichment, number of fuel passes, burnup and effective multiplication factor are taken into account in the calculation to find neutronic characteristics of the present reactor design.

  5. Research reactors spent fuel management in the Nuclear Research Institute Rez

    International Nuclear Information System (INIS)

    Rychecky, J.

    2001-01-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15

  6. Research reactors spent fuel management in the Nuclear Research Institute Rez

    Energy Technology Data Exchange (ETDEWEB)

    Rychecky, J. [Nuclear Research Institute, 25068 Rez (Czech Republic)

    2001-07-01

    In Czech Republic 3 research and testing nuclear reactors are operated at present time, with the biggest one being the Nuclear Research Institute (NRI) reactor LVR-15, operated with maximum power 10 MW. This reactor serves as a radiation source for material testing, producing of ionizing radiation sources, theoretical studies, and, most recently, for boron neutron capture therapy. Another NRI reactor LR-0 is a reactor of zero power used mainly for the studies of WWER 1000 spent fuel criticality. For training of students the reactor called VRABEC (VR-1), operated also with very low power, serves since 1990 at the Faculty of Nuclear Engineering, of Czech Technical University. The similar testing type reactor (SR-0), already decommissioned, was also used since 1974 to 1989 in Skoda, Nuclear Machinery, Plzen. This contribution summarizes the present state of the spent fuel (SF) management of these nuclear reactors. As the SF management is different for very low or zero power reactors and power reactors, the first type will be only briefly discussed, and then the main attention will be devoted to SF management of the NRI experimental reactor LVR-15.

  7. Nuclear Power Engineering Education Program, University of Illinois

    International Nuclear Information System (INIS)

    Jones, B.G.

    1993-01-01

    The DOE/CECo Nuclear Power Engineering Education Program at the University of Illinois in its first year has significantly impacted the quality of the power education which our students receive. It has contributed to: the recently completed upgrade of the console of our Advanced TRIGA reactor which increases the reactor's utility for training, the procurement of new equipment to upgrade and refurbish several of the undergraduate laboratory set-ups, and the procurement of computational workstations in support of the instructional computing laboratory. In addition, smaller amounts of funds were used for the recruitment and retention of top quality graduate students, the support of faculty to visit other institutions to attract top students into the discipline, and to provide funds for faculty to participate in short courses to improve their skills and background in the power area. These items and activities have helped elevate in the student's perspective the role of nuclear power in the discipline. We feel this is having a favorable impact on student career selection and on ensuring the continued supply of well educated nuclear engineering graduates

  8. Cermet-fueled reactors for multimegawatt space power applications

    International Nuclear Information System (INIS)

    Cowan, C.L.; Armijo, J.S.; Kruger, G.B.; Palmer, R.S.; Van Hoomisson, J.E.

    1988-01-01

    The cermet-fueled reactor has evolved as a potential power source for a broad range of multimegawatt space applications. In particular, the fast spectrum reactor concept can be used to deliver 10s of megawatts of electric power for continuous, long term, unattended operation, and 100s of megawatts of electric power for times exceeding several hundred seconds. The system can also be utilized with either a gas coolant in a Brayton power conversion cycle, or a liquid metal coolant in a Rankine power conversion cycle. Extensive testing of the cermet fuel element has demonstrated that the fuel is capable of operating at very high temperatures under repeated thermal cycling conditions, including transient conditions which approach the multimegawatt burst power requirements. The cermet fuel test performance is reviewed and an advanced cermet-fueled multimegawatt nuclear reactor is described in this paper

  9. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  10. Heat pipe nuclear reactor for space power

    Science.gov (United States)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  11. Specific energy released in power reactors

    International Nuclear Information System (INIS)

    Zaritskaya, T.S.; Kiselev, G.V.; Rudik, A.P.; Tsenter, Eh.M.

    1986-01-01

    Technique of determination are described and analysis of specific energy for different methods of critically maintance of RBMK and WWER-440 reactors are conducted. Characteristics of the optimal mode of critically maintanance are determined

  12. Methods and equipments used in power reactors

    International Nuclear Information System (INIS)

    Beraha, R.; Delevallee, A.

    1976-01-01

    The various reactor γ fuel scanning facilities presently operating around the world are reviewed. Both equipments proposed by FRAMATOME are described: one is intended for scanning removable fuel pencils, and the other one for fuel assembly scanning [fr

  13. Assessment of tritium breeding requirements for fusion power reactors

    International Nuclear Information System (INIS)

    Jung, J.

    1983-12-01

    This report presents an assessment of tritium-breeding requirements for fusion power reactors. The analysis is based on an evaluation of time-dependent tritium inventories in the reactor system. The method presented can be applied to any fusion systems in operation on a steady-state mode as well as on a pulsed mode. As an example, the UWMAK-I design was analyzed and it has been found that the startup inventory requirement calculated by the present method significantly differs from those previously calculated. The effect of reactor-parameter changes on the required tritium breeding ratio is also analyzed for a variety of reactor operation scenarios

  14. Optimal Multiuser Zero Forcing with Per-Antenna Power Constraints for Network MIMO Coordination

    Directory of Open Access Journals (Sweden)

    Kaviani Saeed

    2011-01-01

    Full Text Available We consider a multicell multiple-input multiple-output (MIMO coordinated downlink transmission, also known as network MIMO, under per-antenna power constraints. We investigate a simple multiuser zero-forcing (ZF linear precoding technique known as block diagonalization (BD for network MIMO. The optimal form of BD with per-antenna power constraints is proposed. It involves a novel approach of optimizing the precoding matrices over the entire null space of other users' transmissions. An iterative gradient descent method is derived by solving the dual of the throughput maximization problem, which finds the optimal precoding matrices globally and efficiently. The comprehensive simulations illustrate several network MIMO coordination advantages when the optimal BD scheme is used. Its achievable throughput is compared with the capacity region obtained through the recently established duality concept under per-antenna power constraints.

  15. Preliminary nuclear power reactor technology qualitative assessment for Malaysia

    International Nuclear Information System (INIS)

    Shamsul Amri Sulaiman

    2011-01-01

    Since the worlds first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-mile island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many countries, both developed and developing, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of tnb-kepco preliminary feasibility study for nuclear power program in peninsular malaysia and other published presentations and/or papers by multiple experts. The results suggested that pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of generation IV reactors is remote (e.g. Not until 2030), generation III/ III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall lead to deeper and full analyses of the recommended reactor technologies for a comprehensive feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also implore the best way to select systematically the first civilian nuclear power reactor. (Author)

  16. Cornell's LEPP, CHESS research labs expected to get $124 million in NSF funding for elementary particle and X-ray research

    CERN Multimedia

    2003-01-01

    "Cornell University will be awarded up to $124 million over the next five years by the National Science Foundation (NSF) to support research at the Laboratory for Elementary-Particle Physics (LEPP) and the Cornell High Energy Synchrotron Source (CHESS), a national user facility" (1 page).

  17. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    Notley, M.J.F.

    1983-07-01

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  18. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  19. Self-operation type power control device for nuclear reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru.

    1993-01-01

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

  20. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    Wati, Nurokhim

    2008-01-01

    Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

  1. The value and cost of university research reactors

    International Nuclear Information System (INIS)

    Harling, O.K.; Bernard, J.A.

    1989-01-01

    In this paper the authors provide a brief overview of the value and costs of U.S. university research reactors (URRs). More than three dozen URRs are currently operating in an approximately equal number of states. These URRs are an important part of the U.S. capabilities in nuclear science and technology. These multipurpose research facilities are located on the campuses of universities and colleges and therefore are easily accessible to university staff and students as well as to the high-technology industries, which often are located near universities. The close proximity, i.e., convenient location, to a diverse user base is a major reason for the multifaceted applications of URRs, including basic and applied science, technology, education, and industrial applications. The URRs have an extraordinarily broad range of applicability, including medicine and the life sciences, materials science, environmental sciences, earth and planetary sciences, and nuclear energy

  2. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

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

  3. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  4. Fuel management of mixed reactor type power plant systems

    International Nuclear Information System (INIS)

    Csom, Gyula

    1988-01-01

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

  5. Approach to developing reliable space reactor power systems

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  6. Safety and licensing for small and medium power reactors

    International Nuclear Information System (INIS)

    Trauger, D.B.

    1987-01-01

    Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

  7. National nuclear power planning of China and advanced reactor

    International Nuclear Information System (INIS)

    Qian Jihui

    1990-01-01

    The necessity of investigation on the trends of advanced reactor technology all over the world is elabrated while China is going to set up its long-term national nuclear power programme. In author's opinion, thermal reactor power plants will have a quite long period development in the next century and a new trend of second generation NPPs might emerge in the beginning of next century. These new generation advanced reactors are characterized with new design concepts based on the inherent or passive safety features. Among them, most promising ones are those of AP-600 and MHTGR. Chinese experts are paying special attention to and closely following these two directions

  8. An aqueous lithium salt blanket option for fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Steiner, D.; Varsamis, G. (Rensselaer Polytechnic Inst., Troy, NY (USA). Dept. of Nuclear Engineering and Engineering Physics); Deutsch, L.; Rathke, J. (Grumman Corp., Bethpage, NY (USA). Advanced Energy Systems); Gierszewski, P. (Canadian Fusion Fuels Technology Project (CFFTP), Mississauga, ON (Canada))

    1989-04-01

    An aqueous lithium salt blanket (ALSB) concept is proposed which could be the basis for either a power reactor blanket or a test module in an engineering test reactor. The design is based on an austenitic stainless steel structure, a beryllium multiplier, and a salt breeder concentration of about 32 g LiNO/sub 3/ per 100 cm/sup 3/ of H/sub 2/O. To limit tritium release rates, the salt breeder solution is separated from the water coolant circuit. The overall tritium system cost for a 2400 MW (fusion power) reactor is estimated to be 180 million Dollar US87 installed. (orig.).

  9. Comparison of LOFT zero power physics testing measurement results with predicted values

    International Nuclear Information System (INIS)

    Rushton, B.L.; Howe, T.M.

    1978-01-01

    The results of zero power physics testing measurements in LOFT have been evaluated to assess the adequacy of the physics data used in the safety analyses performed for the LOFT FSAR and Technical Specifications. Comparisons of measured data with computed data were made for control rod worths, temperature coefficients, boron worths, and pressure coefficients. Measured boron concentrations at exact critical points were compared with predicted concentrations. Based on these comparisons, the reactivity parameter values used in the LOFT safety analyses were assessed for conservatism

  10. State-plane analysis of zero-voltage-switching resonant dc/dc power converters

    Science.gov (United States)

    Kazimierczuk, Marian K.; Morse, William D.

    The state-plane analysis technique for the zero-voltage-switching resonant dc/dc power converter family of topologies, namely the buck, boost, buck-boost, and Cuk converters is established. The state plane provides a compression of information that allows the designer to uniquely examine the nonlinear dynamics of resonant converter operation. Utilizing the state plane, resonant converter modes of operation are examined and the switching frequencies are derived for the boundaries between these modes, including the boundary of energy conversion.

  11. Station Blackout Analysis of HTGR-Type Experimental Power Reactor

    Science.gov (United States)

    Syarip; Zuhdi, Aliq; Falah, Sabilul

    2018-01-01

    The National Nuclear Energy Agency of Indonesia has decided to build an experimental power reactor of high-temperature gas-cooled reactor (HTGR) type located at Puspiptek Complex. The purpose of this project is to demonstrate a small modular nuclear power plant that can be operated safely. One of the reactor safety characteristics is the reliability of the reactor to the station blackout (SBO) event. The event was observed due to relatively high disturbance frequency of electricity network in Indonesia. The PCTRAN-HTR functional simulator code was used to observe fuel and coolant temperature, and coolant pressure during the SBO event. The reactor simulated at 10 MW for 7200 s then the SBO occurred for 1-3 minutes. The analysis result shows that the reactor power decreases automatically as the temperature increase during SBO accident without operator’s active action. The fuel temperature increased by 36.57 °C every minute during SBO and the power decreased by 0.069 MW every °C fuel temperature rise at the condition of anticipated transient without reactor scram. Whilst, the maximum coolant (helium) temperature and pressure are 1004 °C and 9.2 MPa respectively. The maximum fuel temperature is 1282 °C, this value still far below the fuel temperature limiting condition i.e. 1600 °C, its mean that the HTGR has a very good inherent safety system.

  12. The system of the measurement of reactor power and the monitoring of core power distribution

    International Nuclear Information System (INIS)

    Li Xianfeng

    1999-01-01

    The author mainly describes the measurement of the reactor power and the monitoring of the core power distribution in DAYA BAY nuclear power plant, introduces the calibration for the measurement system. Ex-core nuclear instrumentation system (RPN) and LOCA surveillance system (LSS) are the most important system for the object. they perform the measurement of the reactor power and the monitoring of the core power distribution on-line and timely. They also play the important roles in the reactor control and the reactor protection. For the same purpose there are test instrumentation system (KME) and in-core instrumentation system (RIC). All of them work together ensuring the exact measurement and effective monitoring, ensuring the safety of the reactor power plant

  13. Building America Case Study: New Town Builders' Power of Zero Energy Center, Denver, Colorado (Brochure)

    Energy Technology Data Exchange (ETDEWEB)

    2014-10-01

    New Town Builders, a builder of energy efficient homes in Denver, Colorado, offers a zero energy option for all the homes it builds. To attract a wide range of potential homebuyers to its energy efficient homes, New Town Builders created a 'Power of Zero Energy Center' linked to its model home in the Stapleton community of Denver. This case study presents New Town Builders' marketing approach, which is targeted to appeal to homebuyers' emotions rather than overwhelming homebuyers with scientific details about the technology. The exhibits in the Power of Zero Energy Center focus on reduced energy expenses for the homeowner, improved occupant comfort, the reputation of the builder, and the lack of sacrificing the homebuyers' desired design features to achieve zero net energy in the home. The case study also contains customer and realtor testimonials related to the effectiveness of the Center in influencing homebuyers to purchase a zero energy home.

  14. New Whole-House Solutions Case Study: New Town Builders' Power of Zero Energy Center - Denver, Colorado

    Energy Technology Data Exchange (ETDEWEB)

    None

    2014-10-01

    New Town Builders, a builder of energy efficient homes in Denver, Colorado, offers a zero energy option for all the homes it builds. To attract a wide range of potential homebuyers to its energy efficient homes, New Town Builders created a "Power of Zero Energy Center" linked to its model home in the Stapleton community. This case study presents New Town Builders' marketing approach, which is targeted to appeal to homebuyers' emotions rather than overwhelming homebuyers with scientific details about the technology. The exhibits in the Power of Zero Energy Center focus on reduced energy expenses for the homeowner, improved occupant comfort, the reputation of the builder, and the lack of sacrificing the homebuyers' desired design features to achieve zero net energy in the home. This case study also contains customer and realtor testimonials related to the effectiveness of the Center in influencing homebuyers to purchase a zero energy home.

  15. An overview of future sustainable nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Poullikkas, Andreas [Electricity Authority of Cyprus, P.O. Box 24506, 1399 Nicosia (Cyprus)

    2013-07-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are an evolution of generation II reactors, incorporate passive or inherent safety features that require no active controls or operational intervention to avoid accidents in the event of malfunction, and may rely on gravity, natural convection or resistance to high temperatures. Today, partly due to the high capital cost of large power reactors generating electricity and partly due to the consideration of public perception, there is a shift towards the development of smaller units. These may be built independently or as modules in a larger complex, with capacity added incrementally as required. Small reactors most importantly benefit from reduced capital costs, simpler units and the ability to produce power away from main grid systems. These factors combined with the ability of a nuclear power plant to use process heat for co-generation, make the small reactors an attractive option. Generally, modern small reactors for power generation are expected to have greater simplicity of design, economy of mass production and reduced installation costs. Many are also designed for a high level of passive or inherent safety in the event of malfunction. Generation III+ designs are generally extensions of the generation III concept, which include advanced passive safety features. These designs can maintain the safe state without the use of any active control components. Generation IV reactors, which are future designs that are currently under research and development, will tend to have closed

  16. THE PHASE REACTOR INDUCTANCE SELECTION TECHNIQUE FOR POWER ACTIVE FILTER

    Directory of Open Access Journals (Sweden)

    D. V. Tugay

    2016-12-01

    Full Text Available Purpose. The goal is to develop technique of the phase inductance power reactors selection for parallel active filter based on the account both low-frequency and high-frequency components of the electromagnetic processes in a power circuit. Methodology. We have applied concepts of the electrical circuits theory, vector analysis, mathematical simulation in Matlab package. Results. We have developed a new technique of the phase reactors inductance selection for parallel power active filter. It allows us to obtain the smallest possible value of THD network current. Originality. We have increased accuracy of methods of the phase reactor inductance selection for power active filter. Practical value. The proposed technique can be used in the design and manufacture of the active power filter for real objects of energy supply.

  17. Investigation of materials for fusion power reactors

    Science.gov (United States)

    Bouhaddane, A.; Slugeň, V.; Sojak, S.; Veterníková, J.; Petriska, M.; Bartošová, I.

    2014-06-01

    The possibility of application of nuclear-physical methods to observe radiation damage to structural materials of nuclear facilities is nowadays a very actual topic. The radiation damage to materials of advanced nuclear facilities, caused by extreme radiation stress, is a process, which significantly limits their operational life as well as their safety. In the centre of our interest is the study of the radiation degradation and activation of the metals and alloys for the new nuclear facilities (Generation IV fission reactors, fusion reactors ITER and DEMO). The observation of the microstructure changes in the reactor steels is based on experimental investigation using the method of positron annihilation spectroscopy (PAS). The experimental part of the work contains measurements focused on model reactor alloys and ODS steels. There were 12 model reactor steels and 3 ODS steels. We were investigating the influence of chemical composition on the production of defects in crystal lattice. With application of the LT 9 program, the spectra of specimen have been evaluated and the most convenient samples have been determined.

  18. Radioisotope research, production, and processing at the University of Missouri Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehrhardt, G.J.; Ketring, A.R.; Ja, Wei; Ma, D.; Zinn, K.; Lanigan, J.

    1995-12-31

    The University of Missouri Research Reactor (MURR) is a 10 MW, light-water-cooled and moderated research reactor which first achieved criticality in 1996 and is currently the highest powered university-owned research reactor in the U.S. For many years a major supplier of reactor-produced isotopes for research and commercial purposes, in the last 15 years MURR has concentrated on development of reactor-produced beta-particle emitters for experimental use in nuclear medicine therapy of cancer and rheumatoid arthritis. MURR has played a major role in the development of bone cancer pain palliation with the agents {sup 153}Sm EDTMP and {sup 186}Re/{sup 188}Re HEDP, as well as in the use of {sup 186}Re, {sup 177}Lu, {sup 166}Ho, and {sup 105}Rh for radioimmunotherapy and receptor-agent-guided radiotherapy. MURR is also responsible for the development of therapeutic, {sup 90}Y-labeled glass microspheres for the treatment of liver tumors, a product ({sup 90}Y Therasphere{trademark}) which is currently an approved drug in Canada. MURR has also pioneered the development of {sup 188}W/{sup 188}Re and {sup 99}Mo/{sup 99m}Tc gel generators, which make the use of low specific activity {sup 188}W and {sup 99}Mo practical for such isotope generators.

  19. Radioisotope research, production, and processing at the University of Missouri Research Reactor

    International Nuclear Information System (INIS)

    Ehrhardt, G.J.; Ketring, A.R.; Ja, Wei; Ma, D.; Zinn, K.; Lanigan, J.

    1995-01-01

    The University of Missouri Research Reactor (MURR) is a 10 MW, light-water-cooled and moderated research reactor which first achieved criticality in 1996 and is currently the highest powered university-owned research reactor in the U.S. For many years a major supplier of reactor-produced isotopes for research and commercial purposes, in the last 15 years MURR has concentrated on development of reactor-produced beta-particle emitters for experimental use in nuclear medicine therapy of cancer and rheumatoid arthritis. MURR has played a major role in the development of bone cancer pain palliation with the agents 153 Sm EDTMP and 186 Re/ 188 Re HEDP, as well as in the use of 186 Re, 177 Lu, 166 Ho, and 105 Rh for radioimmunotherapy and receptor-agent-guided radiotherapy. MURR is also responsible for the development of therapeutic, 90 Y-labeled glass microspheres for the treatment of liver tumors, a product ( 90 Y Therasphere trademark) which is currently an approved drug in Canada. MURR has also pioneered the development of 188 W/ 188 Re and 99 Mo/ 99m Tc gel generators, which make the use of low specific activity 188 W and 99 Mo practical for such isotope generators

  20. Operating history of U.S. nuclear power reactors

    International Nuclear Information System (INIS)

    1974-01-01

    The operating history of U. S. nuclear power plants through December 31, 1974 has been collected. Included are those nuclear reactor facilities which produce electricity, even if in token amounts, or which are part of a development program concerned with the generation of electricity through the use of a nuclear reactor as a heat source. The information is based on data furnished by facility operators. The charts are plotted in terms of cumulative thermal energy as a function of time. Since only those shutdowns of five days or more are shown, the charts do not give a detailed history of plant operation. They do, however, give an overview of the operating history of a variety of developmental and experimental nuclear power reactors. The data show the yearly gross generation of electricity for each U. S. nuclear plant and, for civilian power plants, information on reactor availability and plant capacity factor. (U.S.)