WorldWideScience

Sample records for nuclear power reactor

  1. 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.)

  2. 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.

  3. 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

  4. 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

  5. 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.

  6. 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)

  7. 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)

  8. 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

  9. 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.)

  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. 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.)

  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. 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

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. 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

  20. 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)

  1. 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.

  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. 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

  4. 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.)

  5. 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.)

  6. 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...

  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. 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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. Advanced nuclear reactor and nuclear fusion power generation

    International Nuclear Information System (INIS)

    2000-04-01

    This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

  14. 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.

  15. 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...

  16. 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.

  17. 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)

  18. 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

  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. Nuclear safety. Concerns about the nuclear power reactors in Cuba

    International Nuclear Information System (INIS)

    Wells, Jim; Aloise, Gene; Flaherty, Thomas J.; Fitzgerald, Duane; Zavala, Mario; Hayward, Mary Alice

    1992-09-01

    In 1976, the Soviet Union and Cuba concluded an agreement to construct two 440-megawatt nuclear power reactors near Cienfuegos on the south central coast of Cuba, about 180 miles south of Key West, Florida. The construction of these reactors, which began around 1983, was a high priority for Cuba because of its heavy dependence on imported oil. Cuba is estimated to need an electrical generation capacity of 3,000 megawatts by the end of the decade. When completed, the first reactor unit would provide a significant percentage (estimated at over 15 percent) of Cuba's need for electricity. It is uncertain when Cuba's nuclear power reactors will become operational. On September 5, 1992, Fidel Castro announced the suspension of construction at both of Cuba's reactors because Cuba could not meet the financial terms set by the Russian government to complete the reactors. Cuban officials had initially planned to start up the first of the two nuclear reactors by the end of 1993. However, before the September 5 announcement, it was estimated that this reactor would not be operational until late 1995 or early 1996. The civil construction (such as floors and walls) of the first reactor is currently estimated to be about 90 percent to 97 percent complete, but only about 37 percent of the reactor equipment (such as pipes, pumps, and motors) has been installed. The civil construction of the second reactor is about 20 percent to 30 percent complete. No information was available about the status of equipment for the second reactor. According to former Cuban nuclear power and electrical engineers and a technician, all of whom worked at the reactor site and have recently emigrated from Cuba, Cuba's nuclear power program suffers from poor construction practices and inadequate training for future reactor operators. One former official has alleged, for example, that the first reactor containment structure, which is designed to prevent the accidental release of radioactive material into

  2. 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

  3. 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...

  4. 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...

  5. 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

  6. 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

  7. 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.)

  8. 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

  9. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  10. Nuclear reactors for electric power generation

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1987-01-01

    In this article the operation of a nuclear power plant, the status quo about the application of nuclear energy in the world are explained, the subjects of discussion between supporters and adversaries nowadays and the prospects for prolonged usage of nuclear power are summarized, viewed from the actual technical possibilities. 2 refs.; 7 figs.; 2 tabs

  11. 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.)

  12. 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.)

  13. 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

  14. 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

  15. The Carem reactor: Bridging the gap to nuclear power generation

    International Nuclear Information System (INIS)

    Ordonez, J.P.

    1998-01-01

    An idea is presented as an alternative for the introduction of nuclear power in presently non-nuclear countries. This idea involves going through an intermediate step between the traditional research reactor and the first commercial nuclear power plant. This intermediate step would consist of a very small nuclear power plant, with the principal goal of gaining in experience in the country on all the processes involved in introducing commercial nuclear generation. (author)

  16. 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.

  17. 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)

  18. 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.)

  19. Prestressed reactor vessel for nuclear power plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1982-01-01

    With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

  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. 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

  2. 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

  3. 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.)

  4. 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.

  5. 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.

  6. Computerized nuclear material database management system for power reactors

    International Nuclear Information System (INIS)

    Cheng Binghao; Zhu Rongbao; Liu Daming; Cao Bin; Liu Ling; Tan Yajun; Jiang Jincai

    1994-01-01

    The software packages for nuclear material database management for power reactors are described. The database structure, data flow and model for management of the database are analysed. Also mentioned are the main functions and characterizations of the software packages, which are successfully installed and used at both the Daya Bay Nuclear Power Plant and the Qinshan Nuclear Power Plant for the purposed of handling nuclear material database automatically

  7. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  8. 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.)

  9. 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.

  10. 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

  11. 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)

  12. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

    Structure and functioning of the nuclear steam generator system developed by BBR and its components are described. Auxiliary systems, control and load following behaviour and fuel management are discussed and the main data of PWR given. The brochure closes with a perspective of the future of the Muelheim-Kaerlich nuclear power plant. (GL) [de

  13. 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

  14. 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.)

  15. 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...

  16. 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...

  17. 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)

  18. 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

  19. 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

  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. WWER type reactors used as multipurpose nuclear power sources

    International Nuclear Information System (INIS)

    Fiala, J.; Mulak, J.

    1976-01-01

    Safety aspects are assessed of the siting of nuclear power installations in the vicinity of large housing estates and in areas with a high population density, mainly the aspect of the liquidation of the consequences of the maximum credible accident, i.e., the transversal rupture of the primary coolant circuit. The application of WWER type reactors as multipurpose nuclear power sources in Czechoslovakia is justified. It is shown that such a multipurpose nuclear power source differs from a purely condensation nuclear power plant mainly in the design of the secondary stage. The possibilities of such projects are indicated with a view to power and heat operation. (F.M.)

  2. New generation nuclear power units of PWR type integral reactors

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

    1997-01-01

    Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

  3. 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)

  4. New advanced small and medium nuclear power reactors: possible nuclear power plants for Australia

    International Nuclear Information System (INIS)

    Dussol, R.J.

    2003-01-01

    In recent years interest has increased in small and medium sized nuclear power reactors for generating electricity and process heat. This interest has been driven by a desire to reduce capital costs, construction times and interest during construction, service remote sites and ease integration into small grids. The IAEA has recommended that the term 'small' be applied to reactors with a net electrical output less than 300 MWe and the term 'medium' to 300-700 MWe. A large amount of experience has been gained over 50 years in the design, construction and operation of small and medium nuclear power reactors. Historically, 100% of commercial reactors were in these categories in 1951-1960, reducing to 21% in 1991-2000. The technologies involved include pressurised water reactors, boiling water reactors, high temperature gas-cooled reactors, liquid metal reactors and molten salt reactors. Details will be provided of two of the most promising new designs, the South African Pebble Bed Modular Reactor (PBMR) of about 110 MWe, and the IRIS (International Reactor Innovative and Secure) reactor of about 335 MWe. Their construction costs are estimated to be about US$l,000/kWe with a generating cost for the PBMR of about US1.6c/kWh. These costs are lower than estimated for the latest designs of large reactors such as the European Pressurised Reactor (EPR) designed for 1,600 MWe for use in Europe in the next decade. It is concluded that a small or medium nuclear power reactor system built in modules to follow an increasing demand could be attractive for generating low cost electricity in many Australian states and reduce problems arising from air pollution and greenhouse gas emissions from burning fossil fuels

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. Fuel element concept for long life high power nuclear reactors

    Science.gov (United States)

    Mcdonald, G. E.; Rom, F. E.

    1969-01-01

    Nuclear reactor fuel elements have burnups that are an order of magnitude higher than can currently be achieved by conventional design practice. Elements have greater time integrated power producing capacity per unit volume. Element design concept capitalizes on known design principles and observed behavior of nuclear fuel.

  11. New reactor technology: safety improvements in nuclear power systems.

    Science.gov (United States)

    Corradini, M L

    2007-11-01

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

  12. Historical construction costs of global nuclear power reactors

    International Nuclear Information System (INIS)

    Lovering, Jessica R.; Yip, Arthur; Nordhaus, Ted

    2016-01-01

    The existing literature on the construction costs of nuclear power reactors has focused almost exclusively on trends in construction costs in only two countries, the United States and France, and during two decades, the 1970s and 1980s. These analyses, Koomey and Hultman (2007); Grubler (2010), and Escobar-Rangel and Lévêque (2015), study only 26% of reactors built globally between 1960 and 2010, providing an incomplete picture of the economic evolution of nuclear power construction. This study curates historical reactor-specific overnight construction cost (OCC) data that broaden the scope of study substantially, covering the full cost history for 349 reactors in the US, France, Canada, West Germany, Japan, India, and South Korea, encompassing 58% of all reactors built globally. We find that trends in costs have varied significantly in magnitude and in structure by era, country, and experience. In contrast to the rapid cost escalation that characterized nuclear construction in the United States, we find evidence of much milder cost escalation in many countries, including absolute cost declines in some countries and specific eras. Our new findings suggest that there is no inherent cost escalation trend associated with nuclear technology. - Highlights: •Comprehensive analysis of nuclear power construction cost experience. •Coverage for early and recent reactors in seven countries. •International comparisons and re-evaluation of learning. •Cost trends vary by country and era; some experience cost stability or decline.

  13. 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

  14. Economic risks of nuclear power reactor accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Aldrich, D.C.

    1984-04-01

    Models to be used for analyses of economic risks from events which occur during US LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models have been developed for potential use by both the nuclear power industry and regulatory agencies in cost/benefit analyses for decision-making purposes. The new onsite cost models estimate societal losses from power production cost increases, plant capital losses, plant decontamination costs, and plant repair costs which may be incurred after LWR operational events. Early decommissioning costs, plant worker health impact costs, electric utility business costs, nuclear power industry costs, and litigation costs are also addressed. The newly developed offsite economic consequence models estimate The costs of post-accident population protective measures and public health impacts. The costs of population evacuation and temporary relocation, agricultural product disposal, land and property decontamination, and land interdiction are included in the economic models for population protective measures. Costs of health impacts and medical care costs are also included in the models

  15. 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

  16. 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)

  17. 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)

  18. Nuclear power reactor safety research activities in CIAE

    International Nuclear Information System (INIS)

    Pu Shendi; Huang Yucai; Xu Hanming; Zhang Zhongyue

    1994-01-01

    The power reactor safety research activities in CIAE are briefly reviewed. The research work performed in 1980's and 1990's is mainly emphasised, which is closely related to the design, construction and licensing review of Qinshan Nuclear Power Plant and the safety review of Guangdong Nuclear Power Station. Major achievements in the area of thermohydraulics, nuclear fuel, probabilistic safety assessment and severe accident researches are summarized. The foreseeable research plan for the near future, relating to the design and construction of 600 MWe PWR NPP at Qinshan Site (phase II development) is outlined

  19. 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

  20. Potential of light water reactors for future nuclear power plants

    International Nuclear Information System (INIS)

    Gueldner, R.

    2003-01-01

    Energy consumption worldwide is going to increase further in the next few decades. Reliable supplies of electricity can be achieved only by centralized power plant structures. In this scenario, nuclear power plants are going to play a leading role as reliable and competitive plants, also under deregulated market conditions. Today, light water reactors have achieved a leading position, both technically and economically, contributing 85% to worldwide electricity generation in nuclear plants. They will continue to be a proven technology in power generation. In many countries, activities therefore are concentrated on extending the service life of plants beyond a period of forty years. New nuclear generating capacities are expected to be created and added from the end of this decade onward. Most of this capacity will be in light water reactors. The concepts of third-generation reactors will meet all economic and technical safety requirements of the 21st century and will offer considerable potential for further development. Probably some thirty years from now, fourth-generation nuclear power plants will be ready for commercial application. These plants will penetrate especially new sectors of the energy markets. Public acceptance of new nuclear power plants is not a matter of reactor lines, provided that safety requirements are met. The important issue is the management of radioactive waste. The construction of new nuclear power plants in Western Europe and North America mainly hinges on the ability to explain to the public that there is a need for new plants and that nuclear power is fundamental to assuring sustainable development. (orig.)

  1. 77 FR 8902 - Draft Regulatory Guide: Issuance, Availability Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2012-02-15

    ... Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... draft regulatory guide (DG) DG-1271 ``Decommissioning of Nuclear Power Reactors.'' This guide describes... Regulatory Guide 1.184, ``Decommissioning of Nuclear Power Reactors,'' dated July 2000. This proposed...

  2. 10 CFR 73.58 - Safety/security interface requirements for nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... requirements for nuclear power reactors. (a) Each operating nuclear power reactor licensee with a license... 10 Energy 2 2010-01-01 2010-01-01 false Safety/security interface requirements for nuclear power reactors. 73.58 Section 73.58 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF...

  3. Nuclear power reactor licensing and regulation in the United States

    International Nuclear Information System (INIS)

    Shapar, H.K.

    1979-01-01

    The report is devoted to four subjects: an explanation of the origins, statutory basis and development of the present regulatory system in the United States; a description of the various actions which must be taken by a license applicant and by the Nuclear Regulatory Commission before a nuclear power plant can be constructed and placed on-line, an account of the current regulatory practices followed by the US NRC in licensing nuclear power reactors; an identification of some of the 'lessons learned' from the Three Mile Island accident and some proposed regulatory and legislative solutions. (NEA) [fr

  4. 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)

  5. Nuclear power

    International Nuclear Information System (INIS)

    Abd Khalik Wood

    2005-01-01

    This chapter discussed the following topics related to the nuclear power: nuclear reactions, nuclear reactors and its components - reactor fuel, fuel assembly, moderator, control system, coolants. The topics titled nuclear fuel cycle following subtopics are covered: , mining and milling, tailings, enrichment, fuel fabrication, reactor operations, radioactive waste and fuel reprocessing. Special topic on types of nuclear reactor highlighted the reactors for research, training, production, material testing and quite detail on reactors for electricity generation. Other related topics are also discussed: sustainability of nuclear power, renewable nuclear fuel, human capital, environmental friendly, emission free, impacts on global warming and air pollution, conservation and preservation, and future prospect of nuclear power

  6. 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.).

  7. 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.)

  8. Training nuclear power plant personnel on SR-O reactor

    International Nuclear Information System (INIS)

    Cerny, K.; Boucek, F.; Kveton, M.; Prokopec, Z.; Fleischhans, J.

    1983-01-01

    The SR-O reactor is an experimental pool-type reactor with a maximum output of 1 MW and maximum thermal neutron flux density of 5.3x10 13 m -2 s -1 . The reactor is described in detail and its specifications are given. The protection and control systems of the reactor permit both manual and automatic operation. The reactor is used for training courses for nuclear power plant operators and for post-graduate study courses for other specialists. Intensive courses for 4 to 6 persons take 15 to 20 days. The course is adjusted to the results of introductory theoretical tests. An optimal teaching method has been developed based on the flowchart algorithmic method, dividing activities into operations (manipulations with controls, issuing commands, making records, etc.) and decision making (information reception and processing). (M.D.)

  9. Systems aspects of a space nuclear reactor power system

    Science.gov (United States)

    Jaffe, L.; Fujita, T.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    Various system aspects of a 300-kW nuclear reactor power system for spacecraft have been investigated. Special attention is given to the cases of a reusable OTV and a space-based radar. It is demonstrated that the stowed length of the power system is important to mission design, and that orbital storage for months to years may be needed for missions involving orbital assembly.

  10. Design study of ship based nuclear power reactor

    International Nuclear Information System (INIS)

    Su'ud, Zaki; Fitriyani, Dian

    2002-01-01

    Preliminary design study of ship based nuclear power reactors has been performed. In this study the results of thermohydraulics analysis is presented especially related to behaviour of ship motion in the sea. The reactors are basically lead-bismuth cooled fast power reactors using nitride fuels to enhance neutronics and safety performance. Some design modification are performed for feasibility of operation under sea wave movement. The system use loop type with relatively large coolant pipe above reactor core. The reactors does not use IHX, so that the heat from primary coolant system directly transferred to water-steam loop through steam generator. The reactors are capable to be operated in difference power level during night and noon. The reactors however can also be used totally or partially to produce clean water through desalination of sea water. Due to the influence of sea wave movement the analysis have to be performed in three dimensional analysis. The computation time for this analysis is speeded up using Parallel Virtual Machine (PVM) Based multi processor system

  11. Characterization of radioactive waste from nuclear power reactors

    International Nuclear Information System (INIS)

    Piumetti, Elsa H.; Medici, Marcela A.

    2007-01-01

    Different kinds of radioactive waste are generated as result of the operation of nuclear power reactors and in all cases the activity concentration of several radionuclides had to be determined in order to optimize resources, particularly when dealing with final disposal or long-term storage. This paper describes the three basic approaches usually employed for characterizing nuclear power reactor wastes, namely the direct methods, the semi-empirical methods and the analytical methods. For some radionuclides or kind of waste, the more suitable method or combination of methods applicable is indicated, stressing that these methods shall be developed and applied during the waste generation step, i.e. during the operation of the reactor. In addition, after remarking the long time span expected from waste generation to their final disposal, the importance of an appropriate record system is pointed out and some basic requirements that should be fulfilled for such system are presented. It is concluded that the tools for a proper characterization of nuclear reactor radioactive waste are available though such tools should be tailored to each specific reactor and their history. (author) [es

  12. Modular Lead-Bismuth Fast Reactors in Nuclear Power

    Directory of Open Access Journals (Sweden)

    Vladimir Petrochenko

    2012-09-01

    Full Text Available On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

  13. 10 CFR 50.36a - Technical specifications on effluents from nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ...; Ineligibility of Certain Applicants § 50.36a Technical specifications on effluents from nuclear power reactors..., including expected occurrences, as low as is reasonably achievable, each licensee of a nuclear power reactor... the design, construction, and operation of nuclear power reactors indicates that compliance with the...

  14. 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.)

  15. The future of nuclear power and fourth-generation reactors

    International Nuclear Information System (INIS)

    Carre, F.; Renault, C.

    2006-01-01

    Faced with the exhaustion of fossil fuel resources, the output of existing nuclear power must quadruple between now and 2050, and the Commissariat a l'Energie atomique (CEA) and its industrial partners are cooperating in a programme of R and D on future nuclear power. France strategy puts rapid neutron reactors (RNR) at the forefront, in view of their possible introduction by 2040. These reactors allow a more efficient use of uranium resources and minimise the production of long-life nuclear waste. Two technologies which use respectively, sodium and gas as their coolant are being studied. For the sodium RNR, which benefits from significant existing experience, the key is to first improve its economic performance. For the gas RNR, which draws on the principles and the generic assets of the RNR, for those using helium as the coolant, and those with applications at high temperature, it is important firstly to demonstrate the key technologies such as the fuel. The decision of President Chirac to launch the study of a prototype, fourth-generation reactor for 2020 is stimulating the research effort into France future nuclear power. (author)

  16. Occupational radiation exposure at commercial nuclear power reactors, 1978

    International Nuclear Information System (INIS)

    Brooks, B.G.

    1979-11-01

    An updated compilation is presented of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1978. Data received from the 64 light water cooled reactors (LWRs) that had completed at least one year of commercial operation as of December 31, 1978 are included. This represents an increase of seven reactors over the number contained in last year's report. The total number of personnel monitored at LWRs during 1978 increased by approximately 12% to 76,121. The number of workers that received measurable doses, however, increased only 8% to 45,978. The total collective dose for 1978 is estimated to be 31,806 man-rems, a small decrease from last year's value of 32,511, which results in the average dose per worker decreasing slightly to 0.69 rems. The average collective dose per reactor also decreased, by approximately 15%, to a value of 497 man-rems

  17. Fuzzy controller for real time supervision of nuclear power reactor

    International Nuclear Information System (INIS)

    Bala Subramanian, R.

    2012-01-01

    Generally nuclear energy provides about 60% of the whole electricity production. A modulation of the nuclear power plants must be able to respond to the demand on the network. The pressurized water nuclear reactor has to yield correctly a load set point. Fundamentally, two parameters are concerned in leading this task to a successful conclusion: the power axial-offset and the control rods position. The focus of this study is the automation of the control of the power axial-offset by adding soluble boron and by minimizing the volume flows through the water pump. It is also important to take into consideration the liquid waste volume. Water or boron is injected into the reactor primary circuit. At the present time this task is still performed manually by an operator, for all previous attempts to automate it failed. That device, sketchily described in the paper, gave rise to the development of a real-time fuzzy controller for the power axial-offset and the control rods insertion in a pressurized water reactor (PWR). The fuzzy controller, which is the main subject of the paper, expresses more naturally the human expertise, thus avoiding the previous issue of empirical tunings. It was implemented in simulation using Matlab-Simulink on a Sun workstation. Two realistic tests discussed show that the fuzzy controller runs as efficiently as an expert operator does

  18. Reactivity control of nuclear power reactors: new options

    International Nuclear Information System (INIS)

    Alcala, F.

    1984-01-01

    Some actual aspects (referring to economy, non-proliferation and environmental impact) of nuclear power reactors has been analyzed from the point of view of the reactivity control physics. Specially studied have been the physical mechanisms related with the spectral shift control method and their general positive effects on those aspects. The analysis carried out suggested the application of the above method of control to reactors with non-hydrogenous fuel cells, which are mainly characterized by their high moderator/fuel ratio. Finally three different types of such fuel cells are presented and some results about one of them (belonging to a PHWR controlled by graphite rods) are given. (author)

  19. Nuclear power plant laundry drain treatment using membrane bio reactor

    International Nuclear Information System (INIS)

    Tsukamoto, Masaaki; Kohanawa, Osamu; Kinugasa, Atsushi; Ogawa, Naoki; Murogaki, Kenta

    2012-01-01

    In nuclear power plant, the radioactive effluent generated by washing the clothes worn in controlled area and the hand and shower water used at the controlled area are treated in laundry drain treatment system. Although various systems which treat such liquid waste preexist, the traditional treatment system has disadvantages such as high running cost and a large amount of secondary waste generation. To solve these matters, we have considered application of an activated sludge system, membrane bio reactor, which has been practically used in general industry. For nuclear power plant, the activated sludge system has been developed, tested in its adaptability and the adequacy has been proved. Some preexisting treatment systems have been replaced with this activated sludge system for the first time in a domestic nuclear power plant, and the renewal system is now in operation. The result is reported. (author)

  20. Global impact of carbon-14 from nuclear power reactors

    International Nuclear Information System (INIS)

    Moghissi, A.A.; Carter, M.W.

    1977-01-01

    Carbon-14 is produced by nuclear power reactors, predominently as a result of the interaction of a neutron and nitrogen-14 both in the fuel and in the coolant. Several other reactions also contribute to the production of carbon-14. Present operational procedures, in general, for reactors and fuel reprocessing plants result in the release of carbon-14 into the environment. Combustion of fossil fuels and certain industrial operations contribute to the supply of CO 2 in the atmosphere and this contribution is essentially free of carbon-14. Future carbon-14 burdens by assuming a thorough mixing of all CO 2 in the atmosphere is predicted. Available data on electric power generation, fossil fuel combustion and certain other information are used to calculate the projected specific activity of carbon-14 by the year 2000 and the twenty-first century. According to these calculations, the global population dose from carbon-14 can be substantial. Also, carbon-14 in the vicinity of nuclear power reactors is considered. Because of the chemistry of carbon-14, it is shown that local problems may be more significant around BWR's as compared to PWR's. Based on environmental considerations of carbon-14, its increasing production and discharge into the atmosphere, and available control technology, it is recommended that nitrogen use and its presence be minimized in pertinent reactor components and operations

  1. Nuclear detectors for in-core power-reactors

    International Nuclear Information System (INIS)

    Duchene, Jean; Verdant, Robert.

    1979-12-01

    Nuclear reactor control is commonly obtained through neutronic measurements, ex-core and in-core. In large size reactors flux instabilities may take place. For a good monitoring of them, local in-core power measurements become particularly useful. This paper intends to review the questions about neutronic sensors with could be used in-core. A historical account about methods is given first, from early power reactors with brief description of each system. Sensors presently used (ionization fission chambers, self-powered detectors) are then considered and also those which could be developped such as gamma thermometers. Their physical basis, main characteristics and operation modes are detailed. Preliminary tests and works needed for an extension of their life-time are indicated. As an example present irradiation tests at the CEA are then proposed. Two tables will help comparing the characteristics of each type in terms of its precise purpose: fuel monitoring, safety or power control. Finally a table summarizes the kind of sensors mounted on working power reactors and another one is a review of characteristics for some detectors from obtainable commercial sheets [fr

  2. Reactor core cooling device for nuclear power plant

    International Nuclear Information System (INIS)

    Tsuda, Masahiko.

    1992-01-01

    The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

  3. Systems aspects of a space nuclear reactor power system

    International Nuclear Information System (INIS)

    Jaffe, L.; Fujita, T.; Beatty, R.

    1988-01-01

    Selected systems aspects of a 300 kW nuclear reactor power system for spacecraft have been studied. The approach included examination of two candidate missions and their associated spacecraft, and a number of special topics dealing with the power system design and operation. The missions considered were a reusable orbital transfer vehicle and a space-based radar. The special topics included: Power system configuration and scaling, launch vehicle integration, operating altitude, orbital storage, start-up, thawing, control, load following, procedures in case of malfunction, restart, thermal and nuclear radiation to other portions of the spacecraft, thermal stresses between subsystems, boom and cable designs, vibration modes, attitude control, reliability, and survivability. Among the findings are that the stowed length of the power system is important to mission design and that orbital storage for months to years may be needed for missions involving orbital assembly

  4. Demographic statistics pertaining to nuclear power reactor sites

    International Nuclear Information System (INIS)

    1979-10-01

    Population statistics are presented for 145 nuclear power plant sites. Summary tables and figures are included that were developed to aid in the evaluation of trends and general patterns associated with the various parameters of interest, such as the proximity of nuclear plant sites to centers of population. The primary reason for publishing this information at this time is to provide a factual basis for use in discussions on the subject of reactor siting policy. The report is a revised and updated version of a draft report published in December 1977. Errors in the population data base have been corrected and new data tabulations added

  5. Development of portable laser peening systems for nuclear power reactors

    International Nuclear Information System (INIS)

    Chida, Itaru; Uehara, Takuya; Yoda, Masaki; Miyasaka, Hiroyuki; Kato, Hiromi

    2009-01-01

    Stress corrosion cracking (SCC) is the major factor to reduce the reliability of aged reactor components. Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening technology was developed and applied to reactor components in operating BWRs and PWRs. Laser peening is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water without any surface preparations. Laser peening systems, which deliver laser pulses with mirrors or through an optical fiber, were developed and have been applied to preventive maintenance against SCC in nuclear power reactors since 1999. Each system was composed of laser oscillators, a beam delivery system, a laser irradiation head, remote handling equipment and a monitor/control system. Beam delivery with mirrors was accomplished through alignment/tracking functions with sufficient accuracy. Reliable fiber-delivery was attained by the development of a novel input coupling optics and an irradiation head with auto-focusing. Recently, we have developed portable laser peening (PLP) system which could employ both mirror- and fiber- delivery technologies. Size and weight of the PLP system for BWR bottom was almost 1/25 compared to the previous system. PLP system would be the applicable to both BWRs and PWRs as one of the maintenance technologies. (author)

  6. Study on reactor power change and ambiguous control of third Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Wang Gongzhan

    2006-01-01

    The phenomenon of the average power reduction during long term full power operating in Third Qinshan nuclear power plant is analyzed . According to the basic conclusions of reactor power fluctuating derived by probability statistic and calculation the corresponding ambiguous control project is proposed. The operating performance could be achieved by the present controlling project is predicted additionally. (authors)

  7. Nuclear piston engine and pulsed gaseous core reactor power systems

    International Nuclear Information System (INIS)

    Dugan, E.T.

    1976-01-01

    The investigated nuclear piston engines consist of a pulsed, gaseous core reactor enclosed by a moderating-reflecting cylinder and piston assembly and operate on a thermodynamic cycle similar to the internal combustion engine. The primary working fluid is a mixture of uranium hexafluoride, UF 6 , and helium, He, gases. Highly enriched UF 6 gas is the reactor fuel. The helium is added to enhance the thermodynamic and heat transfer characteristics of the primary working fluid and also to provide a neutron flux flattening effect in the cylindrical core. Two and four-stroke engines have been studied in which a neutron source is the counterpart of the sparkplug in the internal combustion engine. The piston motions which have been investigated include pure simple harmonic, simple harmonic with dwell periods, and simple harmonic in combination with non-simple harmonic motion. The results of the conducted investigations indicate good performance potential for the nuclear piston engine with overall efficiencies of as high as 50 percent for nuclear piston engine power generating units of from 10 to 50 Mw(e) capacity. Larger plants can be conceptually designed by increasing the number of pistons, with the mechanical complexity and physical size as the probable limiting factors. The primary uses for such power systems would be for small mobile and fixed ground-based power generation (especially for peaking units for electrical utilities) and also for nautical propulsion and ship power

  8. Nuclear safety as applied to space power reactor systems

    International Nuclear Information System (INIS)

    Cummings, G.E.

    1987-01-01

    To develop a strategy for incorporating and demonstrating safety, it is necessary to enumerate the unique aspects of space power reactor systems from a safety standpoint. These features must be differentiated from terrestrial nuclear power plants so that our experience can be applied properly. Some ideas can then be developed on how safe designs can be achieved so that they are safe and perceived to be safe by the public. These ideas include operating only after achieving a stable orbit, developing an inherently safe design, ''designing'' in safety from the start and managing the system development (design) so that it is perceived safe. These and other ideas are explored further in this paper

  9. Activities in the field of small nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baranaev, Yu.D.; Dolgov, V.V.; Sergeev, Yu.A. [Physics and Power Eng. Inst., Obninsk (Russian Federation). State Res. Centre

    1997-10-01

    Considerable efforts have been undertaken for development, design, construction and operation of small nuclear power plants (SNPP) in Russia. Systematic work in this area was started in the mid-1950s. The driving force for this activity was the awareness that the use of nuclear fuel would practically solve the problem of fuel transportation. As far as the remote northern regions are concerned, this provides the key advantage of nuclear over conventional energy sources. The activity in the field of SNPP has included pre-design analytical feasibility studies and experimental research including large-scale experiments on critical assemblies, thermal and hydraulic test facilities, research and development work, construction and operation of pilot and demonstration SNPPs, and finally, construction and more than 20 years of operation of the commercial SNPP, namely Bilibino nuclear co-generation plant (NCGP) located in Chukotka autonomous district, which is one of the most remote regions in the far north-east of Russia. In recent years, studies have been carried out on the development of several new SNPP designs using advanced reactors of the new generation. Among these are the second stage of Bilibino NCGP, floating NCGP VOLNOLOM-3, designated for siting in the Arctic sea coast area, and a nuclear district heating plant for the town of Apatity, in the Murmansk region. In this paper, the background and current status of the SNPPs are given, and the problems as well as prospects of small nuclear reactors development and implementation are considered. (orig.) 20 refs.

  10. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

    1997-01-01

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  11. Summary of space nuclear reactor power systems, 1983--1992

    Energy Technology Data Exchange (ETDEWEB)

    Buden, D.

    1993-08-11

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power.

  12. Summary of space nuclear reactor power systems, 1983--1992

    International Nuclear Information System (INIS)

    Buden, D.

    1993-01-01

    This report summarizes major developments in the last ten years which have greatly expanded the space nuclear reactor power systems technology base. In the SP-100 program, after a competition between liquid-metal, gas-cooled, thermionic, and heat pipe reactors integrated with various combinations of thermoelectric thermionic, Brayton, Rankine, and Stirling energy conversion systems, three concepts:were selected for further evaluation. In 1985, the high-temperature (1,350 K), lithium-cooled reactor with thermoelectric conversion was selected for full scale development. Since then, significant progress has been achieved including the demonstration of a 7-y-life uranium nitride fuel pin. Progress on the lithium-cooled reactor with thermoelectrics has progressed from a concept, through a generic flight system design, to the design, development, and testing of specific components. Meanwhile, the USSR in 1987--88 orbited a new generation of nuclear power systems beyond the, thermoelectric plants on the RORSAT satellites. The US has continued to advance its own thermionic fuel element development, concentrating on a multicell fuel element configuration. Experimental work has demonstrated a single cell operating time of about 1 1/2-y. Technology advances have also been made in the Stirling engine; an advanced engine that operates at 1,050 K is ready for testing. Additional concepts have been studied and experiments have been performed on a variety of systems to meet changing needs; such as powers of tens-to-hundreds of megawatts and highly survivable systems of tens-of-kilowatts power

  13. Threshold self-powered gamma detector for use as a nuclear reactor power monitor

    International Nuclear Information System (INIS)

    LeVert, F.E.

    1977-01-01

    A study of a threshold self-powered gamma detector for use as a nuclear reactor power monitor was conducted. Measurements were performed to ascertain whether certain detector material arrangements could be used to obtain significant discrimination against low energy gammas. Results indicating agreement between detector response and reactor power output are presented. Evidence of rejection of low energy gammas by the detector is presented. The simplicity of construction and ruggedness of the detector are also discussed

  14. Remerschen nuclear power station with BBR pressurized water reactor

    International Nuclear Information System (INIS)

    Hoffmann, J.P.

    1975-01-01

    On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

  15. Multivariable robust control of an integrated nuclear power reactor

    Directory of Open Access Journals (Sweden)

    A. Etchepareborda

    2002-12-01

    Full Text Available The design of the main control system of the CAREM nuclear power plant is presented. This plant is an inherently safe low-power nuclear reactor with natural convection on the primary coolant circuit and is self-pressurized with a steam dome on the top of the pressure vessel (PV. It is an integrated reactor as the whole primary coolant circuit is within the PV. The primary circuit transports the heat to the secondary circuit through once-through steam generators (SG. There is a feedwater valve at the inlet of the SG and a turbine valve at the outlet of the SG. The manipulated variables are the aperture of these valves and the reactivity of the control rods. The control target is to regulate the primary and secondary pressures and to monitor steam flow reference ramps on a range of nominal flow from 100% to 40%. The requirements for the control system are robust stability, low-order simple controllers and transient/permanent error bounding. The controller design is based on a detailed RETRAN plant model, from which linear perturbed open-loop dynamic models at different powers are identified. Two low-order nominal models with their associated uncertainties are chosen for two different power ranges. Robust controllers with acceptable performances are designed for each range. Numerical optimization based on the loop-shaping method is used for the controller design. The designed controllers are implemented in the RETRAN model and tested in simulations achieving successful results.

  16. Current status of nuclear power generation in Japan and directions in water cooled reactor technology development

    International Nuclear Information System (INIS)

    Miwa, T.

    1991-01-01

    Electric power demand aspects and current status of nuclear power generation in Japan are outlined. Although the future plan for nuclear power generation has not been determined yet the Japanese nuclear research centers and institutes are investigating and developing some projects on the next generation of light water reactors and other types of reactors. The paper describes these main activities

  17. Reactor instrumentation and control in nuclear power plants in Germany

    International Nuclear Information System (INIS)

    Aleite, W.

    1993-01-01

    The pertinent legislation, guidelines and standards of importance for nuclear power plant construction as well as the relevant committees in Germany are covered. The impact of international developments on the German regulatory scene is mentioned. A series of 15 data sheets on reactor control, followed by 5 data sheets on instrumentation and control in nuclear power plants, which were drawn up for German plants, are compared and commented in some detail. Digitalization of instrumentation and control systems continues apace. To illustrate the results that can be achieved with a digitalized information system, a picture series that documents a plant test of behavior on simulated steam generator tube rupture is elaborately commented. An outlook on backfitting and upgrading applications concludes this paper. (orig.) [de

  18. Nuclear power reactors in the world. April 2000 ed.

    International Nuclear Information System (INIS)

    2000-01-01

    Nuclear data presented in this annual publication are based on actual statistical data collected by the IAEA Power Reactor Information System (PRIS). Energy and electricity data for 1999 are estimated since latest available information from the Statistical Office of the United Nations is for 1997. Population data originate from the 'World Population Prospects' (revision 1999) published by the Population Division of the UN Statistical office, and the 1999 values are estimates. The future growth of energy, electricity and nuclear power up to 2020 is presented as low and high estimates in order to encompass the uncertainties associated with the future. These estimates should be viewed as very general growth trends whose validity must constantly be subjected to critical review

  19. Nuclear power reactor analysis, methods, algorithms and computer programs

    International Nuclear Information System (INIS)

    Matausek, M.V

    1981-01-01

    Full text: For a developing country buying its first nuclear power plants from a foreign supplier, disregarding the type and scope of the contract, there is a certain number of activities which have to be performed by local stuff and domestic organizations. This particularly applies to the choice of the nuclear fuel cycle strategy and the choice of the type and size of the reactors, to bid parameters specification, bid evaluation and final safety analysis report evaluation, as well as to in-core fuel management activities. In the Nuclear Engineering Department of the Boris Kidric Institute of Nuclear Sciences (NET IBK) the continual work is going on, related to the following topics: cross section and resonance integral calculations, spectrum calculations, generation of group constants, lattice and cell problems, criticality and global power distribution search, fuel burnup analysis, in-core fuel management procedures, cost analysis and power plant economics, safety and accident analysis, shielding problems and environmental impact studies, etc. The present paper gives the details of the methods developed and the results achieved, with the particular emphasis on the NET IBK computer program package for the needs of planning, construction and operation of nuclear power plants. The main problems encountered so far were related to small working team, lack of large and powerful computers, absence of reliable basic nuclear data and shortage of experimental and empirical results for testing theoretical models. Some of these difficulties have been overcome thanks to bilateral and multilateral cooperation with developed countries, mostly through IAEA. It is the authors opinion, however, that mutual cooperation of developing countries, having similar problems and similar goals, could lead to significant results. Some activities of this kind are suggested and discussed. (author)

  20. Research works at the Physics Institute nuclear reactor for the nuclear power engineering

    International Nuclear Information System (INIS)

    Gavars, V.V.; Kalnin'sh, D.O.; Lapenas, A.A.; Tomsons, E.Ya.; Ulmanis, U.A.

    1985-01-01

    Methods for neutron spectra determination in the nuclear reactor core and vessel have been developed. On their base the neutron spectra at the Novo-Voronezh and kola NPPs have been measured. Such measurements are necessary for the determination of the nuclear fuel reprocessing coefficients, for the evaluation of the construction radiation-damage stability and the NPP economical efficiency on the whole. A new type of the reactor regulator - a liquid metal one - has been created. Such regulators are promising in respect of their use at the NPPs. The base for studying new radiation-damage-stable insulators has been founded. The materials obtained are now applied to designing the reactors of the second (fast) and the third (thermonuclear H) generations. There have developed and by a long-time exploitation checked a hot loop, used for materials irradiation. the nuclear reactor in Salaspils provides training of students being the new brain-power for the nuclear power engineering

  1. Nuclear power station with nuclear reactor accommodated largely secure against catastrophes

    International Nuclear Information System (INIS)

    Rosen, O.

    1987-01-01

    If the nuclear reactor is installed underground near the power station unit, then danger to the environment due to radiation contamination can be largely or nearly completely prevented by a covering of constant thickness or by a covering which can be installed by a catastrophic accident. The extinguishing of a burning reactor is also relatively simple for a reactor accommodated in a pit. The above-mentioned measures can be used individually or combined. (orig./HP) [de

  2. Severe accident considerations in Canadian nuclear power reactors

    International Nuclear Information System (INIS)

    Omar, A.M.; Measures, M.P.; Scott, C.K.; Lewis, M.J.

    1990-08-01

    This paper describes a current study on severe accidents being sponsored by the Atomic Energy Control Board (AECB) and provides background on other related Canadian work. Scoping calculations are performed in Phase I of the AECB study to establish the relative consequences of several permutations resulting from six postulated initiating events, nine containment states, and a selection of meteorological conditions and health effects mitigating criteria. In Phase II of the study, selected accidents sequences would be analyzed in detail using models suitable for the design features of the Canadian nuclear power reactors

  3. A new control strategy for nuclear power reactors

    International Nuclear Information System (INIS)

    Wakabayashi, H.; Sasaki, K.; Takegaki, M.

    1990-01-01

    A new automatic direct digital control strategy for nuclear power reactors is presented. It is based on a simple control logic of comparison between the available time (the time for the error signal to disappear) and the required time (the time for the time derivative to match that of the target trend). The method aims to control the system to an acceptable state within a minimum time under a number of restraints. The control capability of the method is shown for two typical transients. This method is generally applicable to process control in which time-optimal control based on the maximum principle is sought

  4. Micro-structured nuclear fuel and novel nuclear reactor concepts for advanced power production

    International Nuclear Information System (INIS)

    Popa-Simil, Liviu

    2008-01-01

    Many applications (e.g. terrestrial and space electric power production, naval, underwater and railroad propulsion and auxiliary power for isolated regions) require a compact-high-power electricity source. The development of such a reactor structure necessitates a deeper understanding of fission energy transport and materials behavior in radiation dominated structures. One solution to reduce the greenhouse-gas emissions and delay the catastrophic events' occurrences may be the development of massive nuclear power. The actual basic conceptions in nuclear reactors are at the base of the bottleneck in enhancements. The current nuclear reactors look like high security prisons applied to fission products. The micro-bead heterogeneous fuel mesh gives the fission products the possibility to acquire stable conditions outside the hot zones without spilling, in exchange for advantages - possibility of enhancing the nuclear technology for power production. There is a possibility to accommodate the materials and structures with the phenomenon of interest, the high temperature fission products free fuel with near perfect burning. This feature is important to the future of nuclear power development in order to avoid the nuclear fuel peak, and high price increase due to the immobilization of the fuel in the waste fuel nuclear reactor pools. (author)

  5. Closed Brayton cycle power conversion systems for nuclear reactors :

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-04-01

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

  6. Desalination of seawater with nuclear power reactors in cogeneration

    International Nuclear Information System (INIS)

    Flores E, R.M.

    2004-01-01

    The growing demand for energy and hydraulic resources for satisfy the domestic, industrial, agricultural activities, etc. has wakened up the interest to carry out concerning investigations to study the diverse technologies guided to increase the available hydraulic resources, as well as to the search of alternatives of electric power generation, economic and socially profitable. In this sense the possible use of the nuclear energy is examined in cogeneration to obtain electricity and drinkable water for desalination of seawater. The technologies are analysed involved in the nuclear cogeneration (desalination technology, nuclear and desalination-nuclear joining) available in the world. At the same time it is exemplified the coupling of a nuclear reactor and a process of hybrid desalination that today in day the adult offers and economic advantages. Finally, the nuclear desalination is presented as a technical and economically viable solution in regions where necessities of drinkable water are had for the urban, agricultural consumption and industrial in great scale and that for local situations it is possible to satisfy it desalinating seawater. (Author)

  7. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    International Nuclear Information System (INIS)

    Grossbeck, M. L.; Renier, J-P.A.; Bigelow, Tim

    2003-01-01

    Burnable poisons are used in nuclear reactors to produce a more level distribution of power in the reactor core and to reduce to necessity for a large control system. An ideal burnable poison would burn at the same rate as the fuel. In this study, separation of neutron-absorbing isotopes was investigated in order to eliminate isotopes that remain as absorbers at the end of fuel life, thus reducing useful fuel life. The isotopes Gd-157, Dy-164, and Er-167 were found to have desirable properties. These isotopes were separated from naturally occurring elements by means of plasma separation to evaluate feasibility and cost. It was found that pure Gd-157 could save approximately $6 million at the end of four years. However, the cost of separation, using the existing facility, made separation cost- ineffective. Using a magnet with three times the field strength is expected to reduce the cost by a factor of ten, making isotopically separated burnable poisons a favorable method of increasing fuel life in commercial reactors, in particular Generation-IV reactors. The project also investigated various burnable poison configurations, and studied incorporation of metallic burnable poisons into fuel cladding

  8. Core fusion accidents in nuclear power reactors. Knowledge review

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

  9. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

  10. 10 CFR 50.60 - Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for...

    Science.gov (United States)

    2010-01-01

    ... lightwater nuclear power reactors for normal operation. 50.60 Section 50.60 Energy NUCLEAR REGULATORY... lightwater nuclear power reactors for normal operation. (a) Except as provided in paragraph (b) of this section, all light-water nuclear power reactors, other than reactor facilities for which the...

  11. Robots in Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    Koizumi, Masumichi

    1984-01-01

    The Power Reactor and Nuclear Fuel Development Corp. has carried out the technical development concerning ATRs and FBRs, nuclear fuel cycle, the uranium enrichment by centrifugal separation, the reprocessing of spent fuel, and the treatment and disposal of wastes. For the purpose, the Corp. has operated diversified nuclear facilities, and for the operational management of these nuclear facilities, aiming at the reduction of radiation exposure of workers, the shortening of working time, or the rise of the capacity ratio of the facilities, the technical development related to robots has been advanced. Namely, the equipment for the remote maintenace and repair of facilities, the equipment for checkup and monitoring and the equipment for test and inspection are the main subjects of robot development. Hereafter, it is necessary to develop the equipment to which the function of high grade is given and to automate main processes and checkup and monitoring system as well as to improve the reliability and endurance of facilities. The development of the manipulator system for remote maintenance, the facility of handling high radioactive substances and a master-slave manipulator, a power manipulator and a remote transfer equipment, the development of a remote repair and checkup equipment in the reprocessing plant, a remote maintenance and checkup equipment for FBRs and a remote automatic inspection equipment for ATRs are reported. (Kako, I.)

  12. Use of Russian technology of ship reactors with lead-bismuth coolant in nuclear power

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Chitaykin, V.I.; Gromov, B.F.; Grigoryv, O.G.; Dedoul, A.V.; Toshinsky, G.I.; Dragunov, Yu.G.; Stepanov, V.S.

    2000-01-01

    The experience of using lead-bismuth coolant in Russian nuclear submarine reactors has been presented. The fundamental statements of the concept of using the reactors cooled by lead-bismuth alloy in nuclear power have been substantiated. The results of developments for using lead bismuth coolant in nuclear power have been presented. (author)

  13. Economic analysis of nuclear power reactor dissemination to less developed nations with implications for nuclear proliferation

    International Nuclear Information System (INIS)

    Gustavson, R.L.; Howard, J.S. II.

    1979-09-01

    We have applied an economic model to the transfer of nuclear-power reactors from industrialized nations to the less developed nations. The model includes demand and supply factors and predicts the success of US nonproliferation positions and policies. We conclude that economic forces dominate the transfer of power reactors to less developed nations. Our study shows that attempts to either restrict or promote the spread of nuclear-power technology by ignoring natural economic incentives would have only limited effect. If US policy is too restrictive, less developed nations will seek other suppliers and thereby lower US Influence substantially. Allowing less developed nations to develop nuclear-power technology as dictated by economic forces will result in a modest rate of transfer that should comply with nuclear-proliferation objectives

  14. Economic analysis of nuclear power reactor dissemination to less developed nations with implications for nuclear proliferation

    Energy Technology Data Exchange (ETDEWEB)

    Gustavson, R.L.; Howard, J.S. II

    1979-09-01

    An economic model is applied to the transfer of nuclear-power reactors from industrialized nations to the less developed nations. The model includes demand and supply factors and predicts the success of US nonproliferation positions and policies. It is concluded that economic forces dominate the transfer of power reactors to less developed nations. Our study shows that attempts to either restrict or promote the spread of nuclear-power technology by ignoring natural economic incentives would have only limited effect. If US policy is too restrictive, less developed nations will seek other suppliers and thereby lower US Influence substantially. Allowing less developed nations to develop nuclear-power technology as dictated by economic forces will result in a modest rate of transfer that should comply with nuclear-proliferation objectives.

  15. Studies on environment safety and application of advanced reactor for inland nuclear power plants

    International Nuclear Information System (INIS)

    Wei, L.; Jie, L.

    2014-01-01

    To study environment safety assessment of inland nuclear power plants (NPPs), the impact of environment safety under the normal operation was researched and the environment risk of serious accidents was analyzed. Moreover, the requirements and relevant provisions of site selection between international nuclear power plant and China's are comparatively studied. The conclusion was that the environment safety assessment of inland and coastal nuclear power plant have no essential difference; the advanced reactor can meet with high criteria of environment safety of inland nuclear power plants. In this way, China is safe and feasible to develop inland nuclear power plant. China's inland nuclear power plants will be as big market for advanced reactor. (author)

  16. Reactor control device for controlling load of nuclear power plant

    International Nuclear Information System (INIS)

    Hirota, Tadakuni; Yokoyama, Terukuni; Masuda, Jiro.

    1981-01-01

    Purpose: To improve the load follow-up capacity of a nuclear reactor by automatically controlling the width of the not-sensing band of a control rod inserting and removing discriminator circuit. Constitution: When load control operations such as automatic load control, automatic frequency control, governor free operation and so forth are conducted, the width of a not sensing band of a control rod inserting and removing discriminator circuit is ao automatically controlled that the not sensing band width may return to ordinary value in a normal operation by avoiding the fast repetition of inserting and removing control rods by increasing the width of the insensing band if the period of a control deviation signal produced due to the variation in the load is quickly repeated and varied in correspondence to the control deviation signal. That is, a circuit for varying the insensing band of the control circuit for driving a control mechanism is provided to reduce the amount of driving the control rods in a load control operation and to reduce the strain of the power distribution of the nuclear reactor, thereby improving the load control capacity. (Yoshihara, H.)

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  18. 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

  19. Materialistic Aspects of Raising Resource of Pressurized Water Reactors for Low-Power Nuclear Plants

    International Nuclear Information System (INIS)

    Parshin, A.M.; Muratov, O.E.

    2005-01-01

    The opportunity of using ships reactors for low-power nuclear plants is considered. Some aspects of working constructional materials on cases of water-water reactors of ships nuclear units are considered. Advantages of raising resource of ships reactors are shown

  20. Emergency medical assistance programs for nuclear power reactors

    International Nuclear Information System (INIS)

    Linnemann, R.E.; Mettler, F.A. Jr.

    1977-01-01

    This paper deals with a simple but practical medical support of geographically distributed nuclear reactors in isolated areas. A staff of experts at a centre devote their full attention to accident prevention and preparedness at reactor sites. They establish and maintain emergency medical programs at reactor sites and nearby support hospitals. The emphasis is on first aid and emergency treatment by medical attendants who are not and cannot be experts in radiation but do know how to treat patients. (author)

  1. Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Myers, Carl W [Los Alamos National Laboratory; Elkins, Ned Z [Los Alamos National Laboratory

    2008-01-01

    Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

  2. The working lifetime of nuclear power plants and new types of power reactors

    International Nuclear Information System (INIS)

    Bataille, Ch.; Birraux, C.

    2003-01-01

    The report on the working lifetime of nuclear power plants and new reactor types, by Mr Christian Bataille, deputy for the Nord, and Mr Claude Birraux, deputy for Haute-Savoie as well as President of the Office, supplements the studies carried out by the Parliamentary Office on the Safety of Nuclear Installations and Radioactive Wastes: it examines the remaining working life of the EDF nuclear power plants and the current status of projects that might, if circumstances were right, replace the reactors at present in service. The report investigates the different physical and other factors that influence the ageing of nuclear power plants and tackles the question of whether the design life of 40 years could be exceeded in practice. The whole issue of French nuclear power plant is put in perspective and compared with the situation of nuclear plants in Finland, Sweden, Germany and the United States, from the technical and regulatory standpoints. Believing that any attempt to optimise the working lifetime of the power plants currently in service must be accompanied by simultaneous moves aimed at their replacement, Messrs. Christian Bataille and Claude Birraux go on to review in detail the various light water reactor projects being proposed around the world for completion by 2015, as developments of existing models, in particular the EPR reactor of Framatome ANP, characterised by its competitiveness. They suggest that a first such reactor should be built as quickly as possible. Describing the other nuclear systems being investigated by research organisations not only in France but also in the United States and Sweden, Mrs. Christian Bataille and Claude Birraux review the objectives of these and the circumstances in which they might be developed, which would be unlikely to be before 2035 in view of the technological problems to be overcome and the industrial demonstration plants that would be needed

  3. Reactor/Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    Layton, J. P.

    1980-01-01

    Studies are currently underway to assess the technological feasibility of a nuclear-reactor-powered spacecraft propelled by electric thrusters. This vehicle would be capable of performing detailed exploration of the outer planets of the solar system during the remainder of this century. The purpose of this study was to provide comparative information on a closed cycle gas turbine power conversion system. The results have shown that the performance is very competitive and that a 400 kWe space power system is dimensionally compatible with a single Space Shuttle launch. Performance parameters of system mass and radiator area were determined for systems from 100 to 1000 kWe. A 400 kWe reference system received primary attention. The components of this system were defined and a conceptual layout was developed with encouraging results. The preliminary mass determination for the complete power system was very close to the desired goal of 20 kg/kWe. Use of more advanced technology (higher turbine inlet temperature) will substantially improve system performance characteristics.

  4. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  5. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  6. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  7. Job analysis of nuclear power reactor health physics technicians

    International Nuclear Information System (INIS)

    Davis, L.T.; Mazour, T.J.; Clark, P.V.; Todd, R.C.; Marotta, F.J.

    1984-06-01

    This report describes a project, an industry-wide Job Analysis of Nuclear Power Reactor Health Physics Technicians (HPTs), conducted by Brookhaven National Laboratory and Analysis and Technology, Inc. to provide the industry with job-performance data that can be used in systematically defining training programs in terms of required job functions responsibilities, and performance standards. The job-analysis methodology is consistent with that used by the Institute of Nuclear Power Operations (INPO) in similar industry-wide projects and includes administration of over 850 job task questionnaires to utility and contractor Health Physics Technicians throughout the country. Data collected includes task performance (difficulty, importance, and frequency) and industry-wide demographics (job levels, experience, education, and training). The results of this project discussed herein include model job descriptions for HPT positions, summaries of HPT experience, education, and training, industry-wide task listings with task-performance characteristics, and recommendations of selected tasks as a basis for HPT training development. Finally, potential future applications of the data base by utility and contractor organizations in training program development and evaluation and personnel qualifications are discussed

  8. Guidebook to nuclear reactors

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1976-05-01

    A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

  9. Nuclear power

    International Nuclear Information System (INIS)

    1987-01-01

    ''Nuclear Power'' describes how a reactor works and examines the different designs including Magnox, AGR, RBMK and PWR. It charts the growth of nuclear generation in the world and its contributions to world energy resources. (author)

  10. 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

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

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

  12. Estimates of power requirements for a Manned Mars Rover powered by a nuclear reactor

    Science.gov (United States)

    Morley, Nicholas J.; El-Genk, Mohamed S.; Cataldo, Robert; Bloomfield, Harvey

    1991-01-01

    This paper assesses the power requirement for a Manned Mars Rover vehicle. Auxiliary power needs are fulfilled using a hybrid solar photovoltaic/regenerative fuel cell system, while the primary power needs are meet using an SP-100 type reactor. The primary electric power needs, which include 30-kW(e) net user power, depend on the reactor thermal power and the efficiency of the power conversion system. Results show that an SP-100 type reactor coupled to a Free Piston Stirling Engine yields the lowest total vehicle mass and lowest specific mass for the power system. The second lowest mass was for a SP-100 reactor coupled to a Closed Brayton Cycle using He/Xe as the working fluid. The specific mass of the nuclear reactor power system, including a man-rated radiation shield, ranged from 150-kg/kW(e) to 190-kg/KW(e) and the total mass of the Rover vehicle varied depend upon the cruising speed.

  13. Power generation from nuclear reactors in aerospace applications

    Energy Technology Data Exchange (ETDEWEB)

    English, R.E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere. A program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  14. Power Generation from Nuclear Reactors in Aerospace Applications

    Science.gov (United States)

    English, Robert E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere; a program path is suggested to ease the conditions of first use of aircraft nuclear propulsion.

  15. Power generation from nuclear reactors in aerospace applications

    International Nuclear Information System (INIS)

    English, R.E.

    1982-01-01

    Power generation in nuclear powerplants in space is addressed. In particular, the states of technology of the principal competitive concepts for power generation are assessed. The possible impact of power conditioning on power generation is also discussed. For aircraft nuclear propulsion, the suitability of various technologies is cursorily assessed for flight in the Earth's atmosphere. A program path is suggested to ease the conditions of first use of aircraft nuclear propulsion

  16. On the domestically-made heavy forging for reactor pressure vessels of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Pan Xiren; Zhang Chen.

    1988-01-01

    The present situation of the foreign heavy forgings for nuclear reactor pressure vessels and the heavy forgings condition which is used for the Qinshan 300MWe nuclear power plant are described. Some opinions of domestic products is proposed

  17. Reactor control and protection of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhu Jinping; Sun Jiliang

    1996-01-01

    The control and protection simulation of Qinshan 300 MW Nuclear Power Unit, including the nuclear control, the pressurizer pressure control, the pressurizer level control, the rod control, the reactor shutdown protection and engineered safety feature etc are briefly introduced

  18. Nuclear reactor application for high temperature power industrial processes

    International Nuclear Information System (INIS)

    Dollezhal', N.A.; Zaicho, N.D.; Alexeev, A.M.; Baturov, B.B.; Karyakin, Yu.I.; Nazarov, E.K.; Ponomarev-Stepnoj, N.N.; Protzenko, A.M.; Chernyaev, V.A.

    1977-01-01

    This report gives the results of considerations on industrial heat and technology processes (in chemistry, steelmaking, etc.) from the point of view of possible ways, technical conditions and nuclear safety requirements for the use of high temperature reactors in these processes. Possible variants of energy-technological diagrams of nuclear-steelmaking, methane steam-reforming reaction and other processes, taking into account the specific character of nuclear fuel are also given. Technical possibilities and economic conditions of the usage of different types of high temperature reactors (gas cooled reactors and reactors which have other means of transport of nuclear heat) in heat processes are examined. The report has an analysis of the problem, that arises with the application of nuclear reactors in energy-technological plants and an evaluation of solutions of this problem. There is a reason to suppose that we will benefit from the use of high temperature reactors in comparison with the production based on high quality fossil fuel [ru

  19. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0010] Knowledge and Abilities Catalog for Nuclear Power... comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant... developed using this Catalog along with the Operator Licensing Examination Standards for Power Reactors...

  20. Confinement inertial fusion. Power reactors of nuclear fusion by lasers

    International Nuclear Information System (INIS)

    Velarde, G.; Ahnert, C.; Aragones, J.M.; Leira, G; Martinez-Val, J.M.

    1980-01-01

    The energy crisis and the need of the nuclear fusion energy are analized. The nuclear processes in the laser interation with the ablator material are studied, as well as the thermohydrodinamic processes in the implossion, and the neutronics of the fusion. The fusion reactor components are described and the economic and social impact of its introduction in the future energetic strategies.(author)

  1. Spatial effects on the fluctuations of a nuclear power reactor

    International Nuclear Information System (INIS)

    Salinas-Rodriguez, E.; Rodriguez, R.F.; Wio, H.S.

    1990-01-01

    The effects of spatial inhomogeneities in a nuclear system are studied by using the compounding moments method. In particular, the neutron density and temperature equilibrium correlation functions are explicitly calculated for a realistic linearized nuclear reactor model described in terms of a master equation. (author)

  2. 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

  3. 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

  4. Nuclear Reactors for Space Power, Understanding the Atom Series.

    Science.gov (United States)

    Corliss, William R.

    The historical development of rocketry and nuclear technology includes a specific description of Systems for Nuclear Auxiliary Power (SNAP) programs. Solar cells and fuel cells are considered as alternative power supplies for space use. Construction and operation of space power plants must include considerations of the transfer of heat energy to…

  5. Nuclear power plant monitoring method by neural network and its application to actual nuclear reactor

    International Nuclear Information System (INIS)

    Nabeshima, Kunihiko; Suzuki, Katsuo; Shinohara, Yoshikuni; Tuerkcan, E.

    1995-11-01

    In this paper, the anomaly detection method for nuclear power plant monitoring and its program are described by using a neural network approach, which is based on the deviation between measured signals and output signals of neural network model. The neural network used in this study has three layered auto-associative network with 12 input/output, and backpropagation algorithm is adopted for learning. Furthermore, to obtain better dynamical model of the reactor plant, a new learning technique was developed in which the learning process of the present neural network is divided into initial and adaptive learning modes. The test results at the actual nuclear reactor shows that the neural network plant monitoring system is successfull in detecting in real-time the symptom of small anomaly over a wide power range including reactor start-up, shut-down and stationary operation. (author)

  6. Nuclear-Mechanical Coupling: Small Amplitude Mechanical Vibrations and High Amplitude Power Oscillations in Nuclear Reactors

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2008-11-01

    The cores of nuclear reactors, including its structural parts and cooling fluids, are complex mechanical systems able to vibrate in a set of normal modes and frequencies, if suitable perturbed. The cyclic variations in the strain state of the core materials may produce changes in density. Changes in density modify the reactivity. Changes in reactivity modify thermal power. Modifications in thermal power produce variations in temperature fields. Variations in temperature produce variations in strain due to thermal-elastic effects. If the variation of the temperature field is fast enough and if the Doppler Effect and other stabilizing prompt effects in the fuel are weak enough, a fast oscillatory instability could be produced, coupled with mechanical vibrations of small amplitude. A recently constructed, simple mathematical model of nuclear reactor kinetics, that improves the one due to A.S. Thompson, is reviewed. It was constructed in order to study, in a first approximation, the stability of the reactor: a nonlinear nuclear-thermal oscillator (that corresponds to reactor point kinetics with thermal-elastic feedback and with frozen delayed neutron effects) is coupled nonlinearly with a linear mechanical-thermal oscillator (that corresponds to the first normal mode of mechanical vibrations excited by thermo-elastic effects). This mathematical model is studied here from the standpoint of mechanical vibrations. It is shown how, under certain conditions, a suitable mechanical perturbation could elicit fast and growing oscillatory instabilities in the reactor power. Applying the asymptotic method due to Krylov, Bogoliubov and Mitropolsky, analytical formulae that may be used in the calculation of the time varying amplitude and phase of the mechanical oscillations are given, as functions of the mechanical, thermal and nuclear parameters of the reactor. The consequences for the mechanical integrity of the reactor are assessed. Some conditions, mainly, but not exclusively

  7. Nuclear power reactor core melt accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Jacquemain, Didier; Cenerino, Gerard; Corenwinder, Francois; Raimond, Emmanuel IRSN; Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Couturier, Jean; Debaudringhien, Cecile; Duprat, Anna; Dupuy, Patricia; Evrard, Jean-Michel; Nicaise, Gregory; Berthoud, Georges; Studer, Etienne; Boulaud, Denis; Chaumont, Bernard; Clement, Bernard; Gonzalez, Richard; Queniart, Daniel; Peltier, Jean; Goue, Georges; Lefevre, Odile; Marano, Sandrine; Gobin, Jean-Dominique; Schwarz, Michel; Repussard, Jacques; Haste, Tim; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno; Durin, Michel; Andreo, Francois; Atkhen, Kresna; Daguse, Thierry; Dubreuil-Chambardel, Alain; Kappler, Francois; Labadie, Gerard; Schumm, Andreas; Gauntt, Randall O.; Birchley, Jonathan

    2015-11-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus-FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day. Following the introduction, which describes the structure of this book and highlights the objectives of R and D on core melt accidents, this book briefly presents the design and operating principles (Chapter 2) and safety principles (Chapter 3) of the reactors currently in operation in France, as well as the main accident scenarios envisaged and studied (Chapter 4). The objective of these chapters is not to provide exhaustive information on these subjects (the reader should refer to the general reference documents listed in the corresponding chapters), but instead to provide the information needed in order to understand, firstly, the general approach adopted in France for preventing and mitigating the consequences of core melt

  8. Development of Improved Burnable Poisons for Commercial Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Renier, J.A.

    2002-04-17

    Burnable poisons are used in all modern nuclear reactors to permit higher loading of fuel without the necessity of an overly large control rod system. This not only permits a longer core life but can also be used to level the power distribution. Commercial nuclear reactors commonly use B{sub 4}C in separate non-fueled rods and more recently, zirconium boride coatings on the fuel pellets or gadolinium oxide mixed with the fuel. Although the advantages are great, there are problems with using these materials. Boron, which is an effective neutron absorber, transmutes to lithium and helium upon absorption of a neutron. Helium is insoluble and is eventually released to the interior of the fuel rod, where it produces an internal pressure. When sufficiently high, this pressure stress could cause separation of the cladding from the fuel, causing overly high centerline temperatures. Gadolinium has several very strongly absorbing isotopes, but not all have large cross sections and result in residual burnable poison reactivity worth at the end of the fuel life. Even if the amount of this residual absorber is small and the penalty in operation small, the cost of this penalty, even if only several days, can be very high. The objective of this investigation was to study the performance of single isotopes in order to reduce the residual negative reactivity left over at the end of the fuel cycle. Since the behavior of burnable poisons can be strongly influenced by their configuration, four forms for the absorbers were studied: homogeneously mixed with the fuel, mixed with only the outer one-third of the fuel pellet, coated on the perimeter of the fuel pellets, and alloyed with the cladding. In addition, the numbers of fuel rods containing burnable poison were chosen as 8, 16, 64, and 104. Other configurations were chosen for a few special cases. An enrichment of 4.5 wt% {sup 235}U was chosen for most cases for study in order to achieve a 4-year fuel cycle. A standard pressurized

  9. Development of methods for monitoring and controlling power in nuclear reactors

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Santos, Andre Augusto Campagnole dos; Silva, Vitor Vasconcelos Araujo

    2012-01-01

    Redundancy and diversity are two important criteria for power measurement in nuclear reactors. Other criteria such as accuracy, reliability and response speed are also of major concern. Power monitoring of nuclear reactors is normally done by means of neutronic instruments, i.e. by the measurement of neutron flux. The greater the number of channels for power measuring the greater is the reliability and safety of reactor operations. The aim of this research is to develop new methodologies for on-line monitoring of nuclear reactor power using other reliable processes. One method uses the temperature difference between an instrumented fuel element and the pool water below the reactor core. Another method consists of the steady-state energy balance of the primary and secondary reactor cooling loops. A further method is the calorimetric procedure whereby a constant reactor power is monitored as a function of the temperature-rise rate and the system heat capacity. Another methodology, which does not employ thermal methods, is based on measurement of Cherenkov radiation produced within and around the core. The first three procedures, fuel temperature, energy balance and calorimetric, were implemented in the IPR-R1 TRIGA nuclear research reactor at Belo Horizonte (Brazil) and are the focus of the work described here. Knowledge of the reactor thermal power is very important for precise neutron flux and fuel element burnup calculations. The burnup is linearly dependent on the reactor thermal power and its accuracy is important in the determination of the mass of burned 235 U, fission products, fuel element activity, decay heat power generation and radiotoxicity. The thermal balance method developed in this project is now the standard methodology used for IPR-R1 TRIGA reactor power calibration and the fuel temperature measuring is the most reliable way of on-line monitoring of the reactor power. This research project primarily aims at increasing the reliability and safety of

  10. Nuclear reactor descriptions for space power systems analysis

    Science.gov (United States)

    Mccauley, E. W.; Brown, N. J.

    1972-01-01

    For the small, high performance reactors required for space electric applications, adequate neutronic analysis is of crucial importance, but in terms of computational time consumed, nuclear calculations probably yield the least amount of detail for mission analysis study. It has been found possible, after generation of only a few designs of a reactor family in elaborate thermomechanical and nuclear detail to use simple curve fitting techniques to assure desired neutronic performance while still performing the thermomechanical analysis in explicit detail. The resulting speed-up in computation time permits a broad detailed examination of constraints by the mission analyst.

  11. Nuclear power plants without the risk of a reactor disaster

    International Nuclear Information System (INIS)

    Merz, L.

    1980-01-01

    Nowadays everybody believes that all nuclear power plants are an inevitable risk for the life of the people in its environment; this is a prejudice. This article points out that it is possible to plan nuclear power plants in a way that no deaths in the population have to be feared, even in the heaviest possible disturbances. In order to realise such nuclear power plants, it is necessary either to improve the cooling of the fuel rods by passive additional safety measures of which one unlimitedly efficient thus excluding a nuclear melting accident or to improve the safety containers so that it remains sufficiently tight. (orig./HP) [de

  12. Progress in space nuclear reactor power systems technology development - The SP-100 program

    Science.gov (United States)

    Davis, H. S.

    1984-01-01

    Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

  13. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    Science.gov (United States)

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-10-09

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  14. 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.

  15. General view about reactor safety nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Gasparian, A.E.; Silva, D.E.; Salvatore, J.E.L.; Lima, J.M. de

    1991-01-01

    In this paper the authors describe the principles and goals that have guided, as well as the methods that have been used by the National Commission of Nuclear Energy (CNEN) to set forth measures aiming at providing safety to the Brazilian nuclear power plants. The status of the licensing process of these power plants is shown. The performance and the results obtained so far in relation to the nuclear safety are also described. (author)

  16. Identification and evaluation of accident sequences in nuclear power reactors

    International Nuclear Information System (INIS)

    Amendola, A.; Capobianchi, S.; Mancini, G.; Olivi, L.; Volta, G.; Reina, G.

    1981-01-01

    Probabilistic analysis techniques are being more and more used for the evaluation of accident progression in nuclear power plants, especially after the issue of the Reactor Safety Study (Report WASH-1400). This study and subsequent discussions have indicated the necessity of better investigating some major items, namely: adequate data base for the probabilistic evaluations; completeness of the analysis with respect both to accident initiation and behaviour; adequate treatment of uncertainties on the physical and operational parameters governing the accident behaviour. Furthermore, recent occurrences have stressed the importance of the operational aspects of reactor safety, such as plant-specific identification of possible occurrences, their prompt recognition, on-line prediction of subsequent developments and actions to be taken. The paper reviews the contributions in progress at JRC-Ispra to all these aspects, and specifically reports on the following: (1) The set-up of a European Reliability Data System for the acquisition and organisation of operational data of LWRs in the European Community. (2) The development of more complete and realistic models of systems. This work includes multistate static models of components and systems with a view to automatic fault-tree construction and dynamic models for accident sequence identification. The dynamic modelling approach ESCS (Event Sequence and Consequences Spectrum), shown in detail with an example, represents a step forward with respect to event-tree technique and opens new possibilities in dealing with human factors and on-line diagnosis problems. (3) The development of RSM (Response Surface Methodology) for the analysis of uncertainty propagations in consequence and in probability of accident chains. (author)

  17. Characterization of noise sources in nuclear power reactors

    International Nuclear Information System (INIS)

    Andhill, Gustav

    2004-03-01

    Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D

  18. Characterization of noise sources in nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Andhill, Gustav

    2004-03-01

    Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D.

  19. Maintenance management of nuclear power reactors at the stage of research and development

    International Nuclear Information System (INIS)

    Takaya, Shigeru; Chikazawa, Yoshitaka; Kubo, Shigenobu; Hayashida, Kiichi; Tagawa, Akihiro; Yamashita, Atsushi

    2016-07-01

    A maintenance management required to nuclear power reactors at the R and D stage was discussed in this report. It is the most important to ensure safety of nuclear power plants by taking account of characteristics of nuclear power reactors at the R and D stage. In addition, it is needed to establish a system of maintenance management technologies suitable for reactor types. In this report, objectives of maintenance management of nuclear power reactors at the R and D stage were clarified. Next, requirements and consideration for maintenance management of nuclear power reactors at the R and D stage were discussed according to the objectives. 'Code for Maintenance at Nuclear Power Plants' and 'Guide for Maintenance at Nuclear Power Plants' published by the Japan Electric Association were refereed in the discussion. Then, a draft of codes for maintenance management of nuclear power plants at the R and D stage was newly proposed. Finally, an example that the draft codes were applied to components containing sodium, typical components of sodium-cooled fast reactor, was presented. (author)

  20. 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.)

  1. A quality assurance program for nuclear power reactor materials tests at the Ford nuclear reactor

    International Nuclear Information System (INIS)

    Burn, R.R.

    1989-01-01

    The University of Michigan Nuclear Reactor Laboratory Quality Assurance Program has been established to assure that materials testing services provided to electric utilities produce accurate results in accordance with industry standards, sound engineering practice, and customer requirements. The program was prepared to comply with applicable requirements of 10CFR50, Appendix B, of the Code of Federal Regulations and a standard of the American National Standards Institute (ANSI), N45.2. The paper discusses the quality assurance program applicability, organization, qualification and training of personnel, material identification and control, examination and testing, measuring and test equipment, nonconforming test equipment, records, audits, and distribution

  2. Hydrogen in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    1992-01-01

    The Commission of the European Community (CEC) and the International Atomic Energy Agency (IAEA) decided in 1989 to update the state of the art concerning hydrogen in water cooled nuclear power reactors by commissioning a report which would review, all the available information to-date and make recommendations for the future. This joint report was prepared by committees formed by the IAEA and by the CEC. The aim of this report is to review the current understanding on the areas in which the research on hydrogen in LWR is conventionally presented, taking into account the results of the latest reported research developments. The main reactions through which hydrogen is produced are assessed together with their timings. An estimation of the amount of hydrogen produced by each reaction is given, in order to reckon their relative contribution to the hazard. An overview is then given of the state of knowledge of the most important phenomena taking place during its transport from the place of production and the phenomena which control the hydrogen combustion and the consequences of combustion under various conditions. Specific research work is recommended in each sector of the presented phenomena. The last topics reviewed in this report are the hydrogen detection and the prevent/mitigation of pressure and temperature loads on containment structures and structures and safety related equipment caused by hydrogen combustion

  3. Nuclear Power Reactor Core Melt Accidents. Current State of Knowledge

    International Nuclear Information System (INIS)

    Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

    2013-01-01

    For over thirty years, IPSN and subsequently IRSN has played a major international role in the field of nuclear power reactor core melt accidents through the undertaking of important experimental programmes (the most significant being the Phebus- FP programme), the development of validated simulation tools (the ASTEC code that is today the leading European tool for modelling severe accidents), and the coordination of the SARNET (Severe Accident Research Network) international network of excellence. These accidents are described as 'severe accidents' because they can lead to radioactive releases outside the plant concerned, with serious consequences for the general public and for the environment. This book compiles the sum of the knowledge acquired on this subject and summarises the lessons that have been learnt from severe accidents around the world for the prevention and reduction of the consequences of such accidents, without addressing those from the Fukushima accident, where knowledge of events is still evolving. The knowledge accumulated by the Institute on these subjects enabled it to play an active role in informing public authorities, the media and the public when this accident occurred, and continues to do so to this day

  4. Developing the European Center of Competence on VVER-Type Nuclear Power Reactors

    Science.gov (United States)

    Geraskin, Nikolay; Pironkov, Lyubomir; Kulikov, Evgeny; Glebov, Vasily

    2017-01-01

    This paper presents the results of the European educational projects CORONA and CORONA-II which are dedicated to preserving and further developing nuclear knowledge and competencies in the area of VVER-type nuclear power reactors technologies (Water-Water Energetic Reactor, WWER or VVER). The development of the European Center of Competence for…

  5. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

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

  6. Design of an optimal automatic regulator for regulating the power levels of a nuclear reactor in a wind power range

    International Nuclear Information System (INIS)

    Noori Khajavi, M.; Menhaj, M.B.; Ghofrani, M.B.

    2000-01-01

    Nuclear power reactors are, in nature nonlinear and time varying. These characteristics must be considered, if large power variations occur in their working regime. In this paper a robust optimal self-tuning regulator for regulating the power of a nuclear reactor has been designed and simulated. The proposed controller is capable of regulating power levels in a wide power range (10% to 100% power levels). The controller achieves a fast and good transient response. The simulation results show that the proposed controller outperforms the fixed optimal control recently cited in the literature for nuclear power plants

  7. Method and device for controlling nuclear reactor power

    International Nuclear Information System (INIS)

    Takigawa, Yukio; Ebata, Shigeo.

    1988-01-01

    Purpose: To detect and suppress the special power oscillations in the reactor core. Method: Four pairs of LPRM detectors, each pair comprising two detectors are disposed at an identical axial direction of the reactor core and situated at substantially insymmetrical positions at least in longitudinal, vertical and orthogonal directions with respect to the center of te reactor core and LPRM signals from them are inputted into a device for judging special power oscillations. In this case, a standardized mutual relation function is determined on every pair for the respective LPRM signals. Generation of special power oscillations in the reactor core is judged when it is detected that peaks appearing at least in one of the function forms for each pair are negative and have absolute values exceeding a predetermined value and that time of peak is within a predetermined time. The judged signal is inputted to a selected control rod insertion device. The selected control rod insertion device, upon preceiving the signal, inserts selected control rods into the reactor core to suppress the special power oscillations. Accordingly, it is possible to improve the fuel integrity. (Horiuchi, T.)

  8. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

  9. Nuclear power technology system with molten salt reactor for transuranium nuclides burning in closed fuel cycle

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Dudnikov, A.A.; Ignatiev, V.V.; Prusakov, V.N.; Ponomarev-Stepnoy, N.N.; Subbotin, S.A.

    2000-01-01

    A concept of nuclear power technology system with homogeneous molten salt reactors for burning and transmutation of long-lived radioactive toxic nuclides is considered in the paper. Disposition of such reactors in enterprises of fuel cycle allows to provide them with power and facilitate solution of problems with rad waste with minimal losses. (Authors)

  10. Nuclear power plant design characteristics. Structure of nuclear power plant design characteristics in the IAEA Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    2007-03-01

    One of the IAEA's priorities has been to maintain the Power Reactor Information System (PRIS) database as a viable and useful source of information on nuclear reactors worldwide. To satisfy the needs of PRIS users as much as possible, the PRIS database has included also a set of nuclear power plant (NPP) design characteristics. Accordingly, the PRIS Technical Meeting, organized in Vienna 4-7 October 2004, initiated a thorough revision of the design data area of the PRIS database to establish the actual status of the data and make improvements. The revision first concentrated on a detailed review of the design data completion and the composition of the design characteristics. Based on the results of the review, a modified set and structure of the unit design characteristics for the PRIS database has been developed. The main objective of the development has been to cover all significant plant systems adequately and provide an even more comprehensive overview of NPP unit designs stored in the PRIS database

  11. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive ''box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs

  12. Study on regimes of nuclear power plants with WWER-type reactors

    International Nuclear Information System (INIS)

    Akkerman, G.; Khampel', R.; Khentshel', G.; Kertsher, F.; Lyuttsov, K.

    1976-01-01

    The problems are considered of optimization of nuclear fuel loading, the peculiarities of the NPP operation at decreased power, and also the problem of stability operation of NPP with WWER type reactors taking into account specific features of these reactors (partial fuel overloads, change in reactor reactivity with power changes). The two particular interconnected problems discussed are: choice of such a sequence of partial rechargings which ensures the minimum cost of the electric power generated, and increasing the reactor operating time by reducing its power output. Besides the technical and economic estimates, much attention is given to analysing the stability of NPP operation

  13. Small nuclear power reactor emergency electric power supply system reliability comparative analysis

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

    This work presents an analysis of the reliability of the emergency power supply system, of a small size nuclear power reactor. Three different configurations are investigated and their reliability analyzed. The fault tree method is used as the main tool of analysis. The work includes a bibliographic review of emergency diesel generator reliability and a discussion of the design requirements applicable to emergency electrical systems. The influence of common cause failure influences is considered using the beta factor model. The operator action is considered using human failure probabilities. A parametric analysis shows the strong dependence between the reactor safety and the loss of offsite electric power supply. It is also shown that common cause failures can be a major contributor to the system reliability. (author)

  14. Global radioxenon emission inventory based on nuclear power reactor reports.

    Science.gov (United States)

    Kalinowski, Martin B; Tuma, Matthias P

    2009-01-01

    Atmospheric radioactivity is monitored for the verification of the Comprehensive Nuclear-Test-Ban Treaty, with xenon isotopes 131mXe, 133Xe, 133mXe and 135Xe serving as important indicators of nuclear explosions. The treaty-relevant interpretation of atmospheric concentrations of radioxenon is enhanced by quantifying radioxenon emissions released from civilian facilities. This paper presents the first global radioxenon emission inventory for nuclear power plants, based on North American and European emission reports for the years 1995-2005. Estimations were made for all power plant sites for which emission data were unavailable. According to this inventory, a total of 1.3PBq of radioxenon isotopes are released by nuclear power plants as continuous or pulsed emissions in a generic year.

  15. Questions to the reactors power upgrade of the Nuclear Power Plant of Laguna Verde

    International Nuclear Information System (INIS)

    Salas M, B.

    2014-08-01

    The two reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) were subjected to power upgrade labors with the purpose of achieving 20% upgrade on the original power; these labors concluded in August 24, 2010 for the Reactor 1 and in January 16, 2011 for the Reactor 2, however in January of 2014, the NNP-L V has not received by part of the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) the new Operation License to be able to work with the new power, because it does not fulfill all the necessary requirements of safety. In this work is presented and analyzed the information obtained in this respect, with data provided by the Instituto Federal de Acceso a la Informacion Publica y Proteccion de Datos (IFAI) and the Comision Federal de Electricidad (CFE) in Mexico, as well as the opinion of some workers of the NPP-L V. The Governing Board of the CFE announcement that will give special continuation to the behavior on the operation and reliability of the NPP-L V, because the frequency of not announced interruptions was increased 7 times more in the last three years. (Author)

  16. Acceptance criteria for the evaluation of nuclear power reactor security plans

    International Nuclear Information System (INIS)

    1982-08-01

    This guidance document contains acceptance criteria to be used in the NRC license review process. It contains specific criteria for use in evaluating the acceptability of nuclear power reactor security programs as detailed in security plans

  17. Hermetic cable penetrations for containments of nuclear power reactors meet high safety standards

    International Nuclear Information System (INIS)

    Kusserow, J.; Gurr, W.; Pflug, H.

    1985-05-01

    Different types of cable penetrations for containments of nuclear power reactors have been developed and fabricated in the GDR. The technical parameters achieved are in accordance with the radiation protection requirements

  18. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  19. Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

    International Nuclear Information System (INIS)

    1984-06-01

    The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

  20. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Executive summary: main report

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks

  1. Controlling the power output of a nuclear reactor with fuzzy logic

    NARCIS (Netherlands)

    Ruan, D.; Wal, A.J. van der

    1998-01-01

    The application of fuzzy logic control (FLC) in the domain of nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations

  2. Controlling the Power Output of a Nuclear Reactor with Fuzzy Logic

    NARCIS (Netherlands)

    Ruan, D.; Wal, A.J. van der

    1997-01-01

    The application of fuzzy logic control (FLC) in the domain of nuclear industry presents a tremendous challenge. The main reason for this is the public awareness of the risks of nuclear reactors and the very strict safety regulations in force for nuclear power plants. The very same regulations

  3. 10 CFR 50.72 - Immediate notification requirements for operating nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Immediate notification requirements for operating nuclear power reactors. 50.72 Section 50.72 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF... systems. (8) Emergency ac electrical power systems, including: Emergency diesel generators (EDGs...

  4. The present status of nuclear power and prospects for fast reactors - the IAEA outlook

    International Nuclear Information System (INIS)

    Dastidar, P.; Kupitz, J.; Arkhipov, V.

    1991-01-01

    Nuclear power continues to provide a significant amount of the world's electricity supply. Based on the experience gained from about 6000 reactor years of operation, improvements are continuing to be made in the design of nuclear power plants of all types including liquid metal cooled fast reactors. Five demonstration, prototypical or semi-commercial nuclear plants with liquid metal-cooled reactors (LMR) are in operation in the world. Although the commercial deployment of fast reactors has not been seen as urgent due to the availability of adequate low-cost uranium resources there is an awareness in many countries that breeder reactors will be needed in the early decades of the next century. Adequate energy supply for all countries of the world is vital. The exploitation of all non-polluting forms of energy, of which nuclear energy is the most abundant, must be planned now to meet the growing worldwide energy demand. (author)

  5. Inventory of nuclear power plants and research reactors temporary or definitively stopped in industrialized countries

    International Nuclear Information System (INIS)

    Clauzon, J.; Vaubert, B.

    1984-12-01

    This paper presents data and information on the end of the life of nuclear reactors. One deals more particularly with installations of industrialized countries. This report gives the motivations which have involved the definitive shut down of nuclear power plants and of research reactors in the concerned countries. A schedule of definitive reactor shutdowns is presented. Then, one deals with nuclear power plants of which the construction has been stopped. The reasons of these situations are also given. The temporary difficulties met during the construction or the starting of nuclear power plants these last years are mentioned. Most times, there are economical or political considerations, or safety reasons. Finally, the nuclear power plants stopped for more than two years are mentioned [fr

  6. Nuclear reactor capable of electric power generation during in-service inspection

    International Nuclear Information System (INIS)

    Nakamura, Shinsuke; Nogami, Hitoshi.

    1992-01-01

    The nuclear power plant according to the present invention can generate electric power even in a period when one of a pair of reactors is put to in-service inspection. That is, the nuclear power plant of the present invention comprises a system constitution of two nuclear reactors each of 50% thermal power and one turbine power generator of 100% electric power. Further, facilities of various systems relevant to the two reactors each of 50% thermal power, as a pair, are used in common as much as possible in order to reduce the cost for construction and maintenance/ inspection. Further, a reactor building and a turbine building disposed in adjacent with each for paired two reactors each of 50% thermal power are arranged vertically. This arrangement can facilitate the common use of the facilities for various systems and equipments to attain branching and joining of fluids in reactor feed water systems and main steam system pipelines easily with low pressure loss and low impact shocks. The facility utilization factor of such reactors is remarkably improved by doubling the period of continuous power generation. As a result, economic property is remarkably improved. (I.S.)

  7. Molten salt reactors and possible scenarios for future nuclear power deployment

    International Nuclear Information System (INIS)

    Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

    2004-01-01

    An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

  8. An underground nuclear power station using self-regulating heat-pipe controlled reactors

    Science.gov (United States)

    Hampel, V.E.

    1988-05-17

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast- acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor. 5 figs.

  9. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    Science.gov (United States)

    Hampel, Viktor E.

    1989-01-01

    A nuclear reactor for generating electricity is disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working flud in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor.

  10. Underground nuclear power station using self-regulating heat-pipe controlled reactors

    International Nuclear Information System (INIS)

    Hampel, V.E.

    1989-01-01

    The author presents a nuclear reactor for generating electricity disposed underground at the bottom of a vertical hole that can be drilled using conventional drilling technology. The primary coolant of the reactor core is the working fluid in a plurality of thermodynamically coupled heat pipes emplaced in the hole between the heat source at the bottom of the hole and heat exchange means near the surface of the earth. Additionally, the primary coolant (consisting of the working fluid in the heat pipes in the reactor core) moderates neutrons and regulates their reactivity, thus keeping the power of the reactor substantially constant. At the end of its useful life, the reactor core may be abandoned in place. Isolation from the atmosphere in case of accident or for abandonment is provided by the operation of explosive closures and mechanical valves emplaced along the hole. This invention combines technology developed and tested for small, highly efficient, space-based nuclear electric power plants with the technology of fast-acting closure mechanisms developed and used for underground testing of nuclear weapons. This invention provides a nuclear power installation which is safe from the worst conceivable reactor accident, namely, the explosion of a nuclear weapon near the ground surface of a nuclear power reactor

  11. Dynamic power behavior of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Moreira, F.J.

    1984-01-01

    A methodology for the power level evaluation (dynamic behavior) in a Pressurized Water Reactor, during a transient is developed, by solving the point kinetic equation related to the control rod insertion effects and fuel or moderator temperature 'feed-back'. A new version of the thermal-hydraulic code COBRA III P/MIT, is used. In this new version was included, as an option, the methodology developed. (E.G.) [pt

  12. Economic evaluation of nuclear reactor operation utilizing power effect

    International Nuclear Information System (INIS)

    Budinsky, M.; Mydliar, J.

    1988-01-01

    The operation of a reactor at the so-called power effect may substantially increase the burnup of fuel to be removed. The aim of the evaluation of such reactor operation is the optimal determination of the time over which the yield of the higher use of fuel exceeds economic losses resulting from the increased share of constant expenditure of the price of generated kWh of electric power which ensues from such operation. A mathematical model is presented for such evaluation of reactor operation with regard to benefits for the national economy which is the basis of the ESTER 2 computer program. The calculations show that the prices of generated and delivered kWh are minimally 2% less than the prices of generated power without the power effect use. The minimum ranges in the interval of 30 to 50 days. The dependence of the price of generated and delivered kWh from the point of view of the operator of the power plant as well as the component of fuel price of generated kWh will not reach the minimum even after 50 days of operation. From the operating and physical points of view the duration of power effect is not expected to exceed 20 to 30 days which means that from the point of view of the national economy the price of generated and delivered kWh will be 1.6 to 2% less and the fuel component of the price of the generated kWh will be 3 to 4.5% lower. (Z.M.). 5 figs., 3 refs

  13. Space nuclear-power reactor design based on combined neutronic and thermal-fluid analyses

    International Nuclear Information System (INIS)

    Koenig, D.R.; Gido, R.G.; Brandon, D.I.

    1985-01-01

    The design and performance analysis of a space nuclear-power system requires sophisticated analytical capabilities such as those developed during the nuclear rocket propulsion (Rover) program. In particular, optimizing the size of a space nuclear reactor for a given power level requires satisfying the conflicting requirements of nuclear criticality and heat removal. The optimization involves the determination of the coolant void (volume) fraction for which the reactor diameter is a minimum and temperature and structural limits are satisfied. A minimum exists because the critical diameter increases with increasing void fraction, whereas the reactor diameter needed to remove a specified power decreases with void fraction. The purpose of this presentation is to describe and demonstrate our analytical capability for the determination of minimum reactor size. The analysis is based on combining neutronic criticality calculations with OPTION-code thermal-fluid calculations

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  15. System of nuclear power reactor protection using dynamic logic

    International Nuclear Information System (INIS)

    Carvalho, P.V.R. de; Silva, L.C.R.P. da

    1990-01-01

    The aim of this work is the design of a Reactor Protection System (RPS) using dynamic logic as basic circuitry principle. This concept was developed to permit the electronic and eletromagnetic components employment in 'fail-safe' mode applied to automatic shutdown systems. 'Fail-safe' here means that a fail always yields a constant state that leads to a plant shutdown condition. So the normal condition of operation corresponds to an oscillating state response and the fail or abnormal condition to a static one. At present, almost all modern nuclear plant reactor protection systems use dynamic logic, just differing in the kind of technology employed in the construction of the system. In this work we define what technology best fits our necessities, setting out to design a RPS based on this philosophy. (author) [pt

  16. Nuclear power development on the basis of new concepts of nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    Adamov, E.O.; Orlov, V.V.

    2001-01-01

    Current state of nuclear power in the world has been considered and the reasons for its falling short of the great expectations relating to its vigorous development in the outgoing century are considered. Anticipated energy demand of the mankind in the next century is evaluated, suggesting that with exhausted resources of cheap fossil fuel and ecological restrictions it can be satisfied by means of a new nuclear technology meeting the requirements of large-scale power generation in terms of safety and economic indices, moreover, the technology can be elaborated in the context of achievements made in civil and military nuclear engineering. Since the developing countries are the most interested parties, it is just their initiative in the development of nuclear technology at the next stage that could provide an impetus for its actual advance. It is shown that large-scale development of nuclear power, being adequate to increase in energy demand, is possible even if solely large NPP equipped with breeders providing BR (1 are constructed). Requirements for the reactor and fuel cycle technologies are made, their major aspects being: efficient utilization of Pu accumulated and reduction of U specific consumption by at least an order of magnitude, natural inherent safety and deterministic elimination of accidents involving high radioactive releases, assurance of a balance between radiation hazard posed by radioactive wastes disposed and uranium extracted from the ground, nuclear weapons nonproliferation due to fuel reprocessing ruling out potentiality of Pu diversion, reduction of the new generation reactor costs below the costs of today's LWR. (authors)

  17. Power Nuclear Reactors: technology and innovation for development in future; Centrales Nucleares de Potencia: tecnologias actuales e innovaciones para el futuro

    Energy Technology Data Exchange (ETDEWEB)

    Suarez Antola, R [Universidad Catolica del Uruguay, Montevideo(Uruguay); Ministerio de Industria Energia y Minerria, Montevideo(Uruguay)

    2009-07-01

    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.

  18. Quality and Reliability Aspects in Nuclear Power Reactor Fuel Engineering

    International Nuclear Information System (INIS)

    2015-01-01

    In order to decrease costs and increase competitiveness, nuclear utilities use more challenging operational conditions, longer fuel cycles and higher burnups, which require modifications in fuel designs and materials. Different aspects of quality assurance and control, as well as analysis of fuel performance have been considered in a number of specialized publications. The present publication provides a concise but comprehensive overview of all interconnected quality and reliability issues in fuel fabrication, design and operation. It jointly tackles technical, safety and organizational aspects, and contains examples of state of the art developments and good practices of coordinated work of fuel designers, vendors and reactor operators

  19. The changing structure of the international commercial nuclear power reactor industry

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Hill, L.J.; Reich, W.J.; Rowan, W.J.

    1992-12-01

    The objective of this report is to provide an understanding of the international commercial nuclear power industry today and how the industry is evolving. This industry includes reactor vendors, product lines, and utility customers. The evolving structure of the international nuclear power reactor industry implies different organizations making decisions within the nuclear power industry, different outside constraints on those decisions, and different priorities than with the previous structure. At the same time, cultural factors, technical constraints, and historical business relationships allow for an understanding of the organization of the industry, what is likely, and what is unlikely. With such a frame of reference, current trends and future directions can be more readily understood

  20. Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

    International Nuclear Information System (INIS)

    Prelas, M.A.; Boody, F.P.; Zediker, M.S.

    1987-01-01

    An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

  1. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  3. Nuclear power

    International Nuclear Information System (INIS)

    Porter, Arthur.

    1980-01-01

    This chapter of the final report of the Royal Commission on Electric Power Planning in Ontario updates its interim report on nuclear power in Ontario (1978) in the light of the Three Mile Island accident and presents the commission's general conclusions and recommendations relating to nuclear power. The risks of nuclear power, reactor safety with special reference to Three Mile Island and incidents at the Bruce generating station, the environmental effects of uranium mining and milling, waste management, nuclear power economics, uranium supplies, socio-political issues, and the regulation of nuclear power are discussed. Specific recommendations are made concerning the organization and public control of Ontario Hydro, but the commission concluded that nuclear power is acceptable in Ontario as long as satisfactory progress is made in the disposal of uranium mill tailings and spent fuel wastes. (LL)

  4. Nuclear power

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The committee concludes that the nature of the proliferation problem is such that even stopping nuclear power completely could not stop proliferation completely. Countries can acquire nuclear weapons by means independent of commercial nuclear power. It is reasonable to suppose if a country is strongly motivated to acquire nuclear weapons, it will have them by 2010, or soon thereafter, no matter how nuclear power is managed in the meantime. Unilateral and international diplomatic measures to reduce the motivations that lead to proliferation should be high on the foreign policy agenda of the United States. A mimimum antiproliferation prescription for the management of nuclear power is to try to raise the political barriers against proliferation through misuse of nuclear power by strengthening the Non-Proliferation Treaty, and to seek to raise the technological barriers by placing fuel-cycle operations involving weapons-usable material under international control. Any such measures should be considered tactics to slow the spread of nuclear weapons and thus earn time for the exercise of statesmanship. The committee concludes the following about technical factors that should be considered in formulating nuclear policy: (1) rate of growth of electricity use is a primary factor; (2) growth of conventional nuclear power will be limited by producibility of domestic uranium sources; (3) greater contribution of nuclear power beyond 400 GWe past the year 2000 can only be supported by advanced reactor systems; and (4) several different breeder reactors could serve in principle as candidates for an indefinitely sustainable source of energy

  5. Nuclear Reactor Power Monitoring Using Silicon Carbide Semiconductor Radiation Detectors

    International Nuclear Information System (INIS)

    Thomas Blue; Don Miller

    2008-01-01

    To provide a perspective for our accomplishments, all of the sub-tasks in Task 1 (as they were identified in the proposal) are listed, and a brief description of the subtasks is given. Task 1--Define Generation IV Reactor Power Monitoring Requirements. Task 1.1--The power monitoring requirements for the IRIS and GT-MHR will be evaluated. Parameters considered will include maximum power level uncertainty, response time, etc. Task 1.2--The optimum locations for power monitors will be selected for both the IRIS and GT-MHR. Factors to be considered will include the power monitoring requirements defined in Task 1.1 as well as expected detector sensitivity and the presence of gamma ray background. Task 1.3--Other applications and opportunities offered by SiC power monitors will be evaluated. The prospects for on-line fault identification and diagnosis using pulse height and pulse shape analysis will be explored. The use of miniature SiC detectors to define axial, azimuthal, and radial flux profiles will be investigated

  6. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  7. Transferring knowledge and know-how from the nuclear power community to the research reactor community

    International Nuclear Information System (INIS)

    Wijtsma, F.J.

    2002-01-01

    Question What is the best way of transferring knowledge and know-how from the nuclear power community to the research reactor community, e.g. in the fields of quality assurance, safety culture, etc.? To answer the question on how to transfer knowledge and know-how from the nuclear power community to the research reactor community, one should first try to establish what are the differences and similarities between these types of nuclear facilities. Despite the big difference between the primary objectives of these two kinds of facilities, i.e. electricity production versus providing irradiation services, the underlying safety culture should be comparable. For historical reasons, nuclear power plant management took the lead in establishing fully accepted safety standards. However, research reactors can avail themselves of the wide body of nuclear safety experience accumulated at nuclear power plants. This should be applicable to all nuclear facilities. Nonetheless, in transferring their know-how, safety specialists should take into account the huge differences between critical assemblies, university reactors, small research reactors and multi-purpose high power research reactors. The goal to which a specific facility is dedicated bears heavily upon the outlook of its management Question: How can well run research reactors help problem research reactors? To answer this, a basic question should in turn be posed: Should one help a research reactor with operational difficulties? And, if so, to what extent? Who will benefit? Within the framework of this meeting, one should concentrate on nuclear safety, which is determined by: Safety culture (including quality assurance); The level of training of all staff; Ageing (installation, staff and documentation); The front/back end of the fuel cycle; A strong programme versus extended shutdown; Regulatory (nuclear regulatory) inspectorates; National (international) co-operation; The financial situation prevailing at the

  8. Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  9. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  10. Direct Estimation of Power Distribution in Reactors for Nuclear Thermal Space Propulsion

    Science.gov (United States)

    Aldemir, Tunc; Miller, Don W.; Burghelea, Andrei

    2004-02-01

    A recently proposed constant temperature power sensor (CTPS) has the capability to directly measure the local power deposition rate in nuclear reactor cores proposed for space thermal propulsion. Such a capability reduces the uncertainties in the estimated power peaking factors and hence increases the reliability of the nuclear engine. The CTPS operation is sensitive to the changes in the local thermal conditions. A procedure is described for the automatic on-line calibration of the sensor through estimation of changes in thermal .conditions.

  11. Fuel element for high-temperature nuclear power reactors

    International Nuclear Information System (INIS)

    Schloesser, J.

    1974-01-01

    The fuel element of the HTGR consists of a spherical graphite body with a spherical cavity. A deposit of fissile material, e.g. coated particles of uranium carbide, is fixed to the inner wall using binders. In addition to the fissile material, there are concentric deposits of fertile material, e.g. coated thorium carbide particles. The remaining cavity is filled with a graphite mass, preferably graphite powder, and the filling opening with a graphite stopper. At the beginning of the reactor operation, the fissile material layer provides the whole power. With progressing burn-up, the energy production is taken over by the fertile layer, which provides the heat production until the end of burn-up. Due to the relatively small temperature difference between the outer wall of the outer graphite body and the maximum fuel temperature, the power of the fuel element can be increased. (DG) [de

  12. Indian nuclear power programme with pressurised heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-09-01

    This compilation is a part of ongoing efforts by Nuclear Power Corporation (NPC) to enable persons: to visit the plants under construction and operation to see for themselves adoption of new and advanced techniques; to have contact with the realities of NPC`s facilities; to familiarize themselves with the regulatory aspects on radiological and environmental protection; and assess for themselves the extent of thrust and importance given to overall safety. figs., tabs.

  13. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing.

  14. The role of actinide burning and the Integral Fast Reactor in the future of nuclear power

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Lidsky, L.M.; Miller, M.M.

    1990-12-01

    A preliminary assessment is made of the potential role of actinide burning and the Integral Fast Reactor (IFR) in the future of nuclear power. The development of a usable actinide burning strategy could be an important factor in the acceptance and implementation of a next generation of nuclear power. First, the need for nuclear generating capacity is established through the analysis of energy and electricity demand forecasting models which cover the spectrum of bias from anti-nuclear to pro-nuclear. The analyses take into account the issues of global warming and the potential for technological advances in energy efficiency. We conclude, as do many others, that there will almost certainly be a need for substantial nuclear power capacity in the 2000--2030 time frame. We point out also that any reprocessing scheme will open up proliferation-related questions which can only be assessed in very specific contexts. The focus of this report is on the fuel cycle impacts of actinide burning. Scenarios are developed for the deployment of future nuclear generating capacity which exploit the advantages of actinide partitioning and actinide burning. Three alternative reactor designs are utilized in these future scenarios: The Light Water Reactor (LWR); the Modular Gas-Cooled Reactor (MGR); and the Integral Fast Reactor (FR). Each of these alternative reactor designs is described in some detail, with specific emphasis on their spent fuel streams and the back-end of the nuclear fuel cycle. Four separation and partitioning processes are utilized in building the future nuclear power scenarios: Thermal reactor spent fuel preprocessing to reduce the ceramic oxide spent fuel to metallic form, the conventional PUREX process, the TRUEX process, and pyrometallurgical reprocessing

  15. Alteration in reactor installation (addition of Unit 2) in Shimane Nuclear Power Station, Chugoku Electric Power Co., Inc. (inquiry)

    International Nuclear Information System (INIS)

    1983-01-01

    An inquiry was made by the Ministry of International Trade and Industry to Nuclear Safety Commission on the addition of Unit 2 in Shimane Nuclear Power Station of The Chugoku Electric Power Co., Inc., concerning the technical capability of Chugoku Electric Power Co., Inc., and the plant safety. The NSC requested the Committee on Examination of Reactor Safety to make a deliberation on this subject. Both the technical capability and the safety of Unit 1 were already confirmed by MITI. Unit 2 to be newly added in the Shimane Nuclear Power Station is a BWR power plant with electric output of 820 MW. The examination made by MITI is described: the technical capability of Chugoku Electric Power Co., Inc., the safety of Unit 2 about its siting, reactor proper, reactor cooling system, radioactive waste management, etc. (J.P.N.)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

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

  17. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  18. Nuclear reactor power as applied to a space-based radar mission

    Science.gov (United States)

    Jaffe, L.; Beatty, R.; Bhandari, P.; Chow, E.; Deininger, W.; Ewell, R.; Fujita, T.; Grossman, M.; Bloomfield, H.; Heller, J.

    1988-01-01

    A space-based radar mission and spacecraft are examined to determine system requirements for a 300 kWe space nuclear reactor power system. The spacecraft configuration and its orbit, launch vehicle, and propulsion are described. Mission profiles are addressed, and storage in assembly orbit is considered. Dynamics and attitude control and the problems of nuclear and thermal radiation are examined.

  19. The feature of high flux engineering test reactor and its role in nuclear power development

    International Nuclear Information System (INIS)

    Lu Guangquan

    1987-01-01

    The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

  20. Study of the Utilization BWR Type Nuclear Power Reactor for Desalination Process

    International Nuclear Information System (INIS)

    Itjeu Karliana; Sumijanto; Dhandhang Purwadi, M.

    2008-01-01

    The needs of fresh water increased by rapid population growth and industrials expansion, but these demands can not be prepared naturally. Following this case, seawater desalination becomes the primer option which can fulfill the need through the nuclear desalination technology. The coupled nuclear power reactor enables to supply thermal energy for auxiliary equipment and pumps operation. The utilization study of power reactor type BWR coupled with desalination process has been performed. The goal of study is to obtain characteristic data of desalted water specification which desalination system coupling with nuclear power plant produced energy for desalination process. The study is carried out by browsing data and information, and comprehensive review of thermal energy correlation between NPP with desalination process installation. According to reviewing are found that the thermal energy and electric power utilization from the nuclear power reactor are enable to remove the seawater to produce desalted water and also to operate auxiliary equipments. The assessment results is VK-300 reactor prototype, BWR type 250 MW(e) power are cogeneration unit can supplied hot steam temperature 285 °C to the extraction turbine to empower 150 MW electric power, and a part of hot steam 130 °C is use to operate desalination process and remind heat is distribute to the municipal and offices at that region. The coupled of VK-300 reactor power type BWR with desalination installation of MED type enable to produce desalted water with high quality distillate. Based on the economic calculation that the VK-300 reactor power of BWR type produced water distillate capacity is 300.000 m 3 /hour with cost US$ 0.58/m 3 . The coupling VK-300 reactor power type BWR with MED desalination plant is competitive economically. (author)

  1. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

  2. Study of reactor Brayton power systems for nuclear electric spacecraft

    Science.gov (United States)

    1979-01-01

    The feasibility of using Brayton power systems for nuclear electric spacecraft was investigated. The primary performance parameters of systems mass and radiator area were determined for systems from 100 to 1000 kW sub e. Mathematical models of all system components were used to determine masses and volumes. Two completely independent systems provide propulsion power so that no single-point failure can jeopardize a mission. The waste heat radiators utilize armored heat pipes to limit meteorite puncture. The armor thickness was statistically determined to achieve the required probability of survival. A 400 kW sub e reference system received primary attention as required by the contract. The components of this system were defined and a conceptual layout was developed with encouraging results. An arrangement with redundant Brayton power systems having a 1500 K (2240 F) turbine inlet temperature was shown to be compatible with the dimensions of the space shuttle orbiter payload bay.

  3. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    Fribourg, Ch.

    2001-01-01

    This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

  4. Development of Power Controller System based on Model Reference Adaptive Control for a Nuclear Reactor

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Ridzuan Abdul Mutalib

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP)-type TRIGA Mark II was installed in the year 1982. The Power Controller System (PCS) or Automated Power Controller System (APCS) is very important for reactor operation and safety reasons. It is a function of controlled reactivity and reactor power. The existing power controller system is under development and due to slow response, low accuracy and low stability on reactor power control affecting the reactor safety. The nuclear reactor is a nonlinear system in nature, and it is power increases continuously with time. The reactor parameters vary as a function of power, fuel burnup and control rod worth. The output power value given by the power control system is not exactly as real value of reactor power. Therefore, controller system design is very important, an adaptive controller seems to be inevitable. The method chooses is a linear controller by using feedback linearization, for example Model Reference Adaptive Control. The developed APCS for RTP will be design by using Model Reference Adaptive Control (MRAC). The structured of RTP model to produce the dynamic behaviour of RTP on entire operating power range from 0 to 1MWatt. The dynamic behavior of RTP model is produced by coupling of neutronic and thermal-hydraulics. It will be developed by using software MATLAB/Simulink and hardware module card to handle analog input signal. A new algorithm for APCS is developed to control the movement of control rods with uniformity and orderly for RTP. Before APCS test to real plant, simulation results shall be obtained from RTP model on reactor power, reactivity, period, control rod positions, fuel and coolant temperatures. Those data are comparable with the real data for validation. After completing the RTP model, APCS will be tested to real plant on power control system performance by using real signal from RTP including fail-safe operation, system reliable, fast response, stability and accuracy. The new algorithm shall be a satisfied

  5. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  6. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Science.gov (United States)

    2011-03-28

    ... design certification (DC) application for new nuclear power reactors under Title 10 of the Code of... NUCLEAR REGULATORY COMMISSION [NRC-2010-0033; DC/COL-ISG-021] Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby...

  7. 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.)

  8. Nuclear Reactor Physics

    Science.gov (United States)

    Stacey, Weston M.

    2001-02-01

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

  9. The simulation study on the Nuclear Heating Reactor's power auto-control system

    International Nuclear Information System (INIS)

    Yang Zhijun; Liu Longzhi; Hu Guifen

    2000-01-01

    The power automatic control system on nuclear heating reactor (NHR) is a multi-input and multi-output non-linear system. The power automatic control system on NHR is studied by modern control theory. Through the simulation experiments, it is clear that adopting μ outer-loop and LQR inner-loop feedback, the best control results are obtained

  10. Ensuring the operational safety of nuclear power plants with WWER reactors

    International Nuclear Information System (INIS)

    Shasharin, G.A.; Veretennikov, G.A.; Abagyan, A.A.; Lesnoj, S.A.

    1984-01-01

    At the start of 1983, 27 nuclear power producing units with reactor facilities of the WWER type were in operation in the Soviet Union and other countries. In 1982 the average load factor for nuclear power plants with WWER reactors was 73 per cent. There was not a single nuclear accident or even damage with any significant radiation consequences in the WWER reactors during the entire period of their operation. The most modern nuclear power plants with WWER-440 and WWER-1000 reactors meet all present-day international requirements. Safe operation of the plants is achieved by a variety of measures, the most important of which include: procedures for increasing the reliability of plant equipment and systems; ensuring exact compliance with plant operating instructions; ensuring reliable operation of plant safety systems; action directed towards maintaining the skills of plant personnel at a level adequate to ensure the taking of proper action during transient processes and accident situations. The paper discusses concrete steps for ensuring safe nuclear power plant operation along these lines. In particular, measures such as the following are described: the use of a system for collecting and processing information on equipment failures and defects; the development and introduction of methods of early defect diagnosis; the performance of complex testing of safety systems; the training of highly skilled personnel for nuclear power plants at educational combines and at teaching and training centres making use of simulators; arranging accident-prevention training and special instruction for personnel. (author)

  11. 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)

  12. 15 N utilization in nitride nuclear fuels for advanced nuclear power reactors and accelerator - driven systems

    International Nuclear Information System (INIS)

    Axente, D.

    2005-01-01

    15 N utilization for nitride nuclear fuels production for nuclear power reactors and accelerator - driven systems is presented. Nitride nuclear fuel is the obvious choice for advanced nuclear reactors and ADS because of its favorable properties: a high melting point, excellent thermal conductivity, high fissile density, lower fission gas release and good radiation tolerance. The application of nitride fuels in nuclear reactors and ADS requires use of 15 N enriched nitrogen to suppress 14 C production due to (n,p) reaction on 14 N. Accelerator - driven system is a recent development merging of accelerator and fission reactor technologies to generate electricity and transmute long - lived radioactive wastes as minor actinides: Np, Am, Cm. A high-energy proton beam hitting a heavy metal target produces neutrons by spallation. The neutrons cause fission in the fuel, but unlike in conventional reactors, the fuel is sub-critical and fission ceases when the accelerator is turned off. Nitride fuel is a promising candidate for transmutation in ADS of minor actinides, which are converted into nitrides with 15 N for that purpose. Tacking into account that the world wide market is about 20 to 40 Kg 15 N annually, the supply of that isotope for nitride fuel production for nuclear power reactors and ADS would therefore demand an increase in production capacity by a factor of 1000. For an industrial plant producing 100 t/y 15 N, using present technology of isotopic exchange in NITROX system, the first separation stage of the cascade would be fed with 10M HNO 3 solution of 600 mc/h flow - rate. If conversion of HNO 3 into NO, NO 2 , at the enriching end of the columns, would be done with gaseous SO 2 , for a production plant of 100 t/y 15 N a consumption of 4 million t SO 2 /y and a production of 70 % H 2 SO 4 waste solution of 4.5 million mc/y are estimated. The reconversion of H 2 SO 4 into SO 2 in order to recycle of SO 2 is a problem to be solved to compensate the cost of SO 2

  13. Multicriteria choice of a nuclear reactor and nuclear power project application in Tunisian

    International Nuclear Information System (INIS)

    Rihane, Mehdi

    2009-01-01

    qualitative choice of a nuclear reactor affects multiple domain, technological, industrial, economic and even political. The evaluation criteria for nuclear reactor are multiple, in this study we will cite the most important : technicoeconomique competitiveness, security at large, intrinsic and extrinsic impact on the environment, sustainable development, the impact on domestic industry and resistance nuclear proliferation. from each of these criteria, a classification has been proposed.

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

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

  15. Development of the fuel-cycle costs in nuclear power stations with light-water reactors

    International Nuclear Information System (INIS)

    Brosch, R.; Moraw, G.; Musil, G.; Schneeberger, M.

    1976-01-01

    The authors investigate the fuel-cycle costs in nuclear power stations with light-water reactors in the Federal Republic of Germany in the years 1966 to 1976. They determine the effect of the price development for the individual components of the nuclear fuel cycle on the fuel-cycle costs averaged over the whole power station life. Here account is taken also of inflation rates and the change in the DM/US $ parity. In addition they give the percentage apportionment of the fuel-cycle costs. The authors show that real fuel-cycle costs for nuclear power stations with light-water reactors in the Federal Republic of Germany have risen by 11% between 1966 and 1976. This contradicts the often repeated reproach that fuel costs in nuclear power stations are rising very steeply and are no longer competitive. (orig.) [de

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  17. Power probability density function control and performance assessment of a nuclear research reactor

    International Nuclear Information System (INIS)

    Abharian, Amir Esmaeili; Fadaei, Amir Hosein

    2014-01-01

    Highlights: • In this paper, the performance assessment of static PDF control system is discussed. • The reactor PDF model is set up based on the B-spline functions. • Acquaints of Nu, and Th-h. equations solve concurrently by reformed Hansen’s method. • A principle of performance assessment is put forward for the PDF of the NR control. - Abstract: One of the main issues in controlling a system is to keep track of the conditions of the system function. The performance condition of the system should be inspected continuously, to keep the system in reliable working condition. In this study, the nuclear reactor is considered as a complicated system and a principle of performance assessment is used for analyzing the performance of the power probability density function (PDF) of the nuclear research reactor control. First, the model of the power PDF is set up, then the controller is designed to make the power PDF for tracing the given shape, that make the reactor to be a closed-loop system. The operating data of the closed-loop reactor are used to assess the control performance with the performance assessment criteria. The modeling, controller design and the performance assessment of the power PDF are all applied to the control of Tehran Research Reactor (TRR) power in a nuclear process. In this paper, the performance assessment of the static PDF control system is discussed, the efficacy and efficiency of the proposed method are investigated, and finally its reliability is proven

  18. Small modulator reactors (SMRs) - the key to increased social acceptance of nuclear power?

    Energy Technology Data Exchange (ETDEWEB)

    Sam-Aggrey, H. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Small Modular Reactors (SMRs) have been touted as having the potential to spearhead a nuclear renaissance. Proponents of this view cite the relative advantages of SMRs over larger nuclear plants as reason for their optimism. Some of these merits include potential economies of mass production deriving from modular factory production, enhanced safety features, the ability to locate reactors in remote areas, and reduced requirements for technical workforces to install and maintain SMRs. Despite the optimism surrounding SMRs, nuclear power is still a contentious issue for many Canadians. The growth of nuclear power remains adversely affected by concerns over issues such as: the health and safety consequences of nuclear power generation, environmental impacts, proliferation of nuclear materials, terrorism, and radioactive waste disposal. Personal values, past nuclear accidents, and lack of trust in the industry are other factors influencing attitudes toward nuclear energy. Given that public support is pivotal to the growth of the nuclear power sector, the issues surrounding the social acceptance of SMRs should be given serious consideration. Drawing on previously published data, this paper analyses the purported qualities of SMRs with a view of determining whether these characteristics have the potential of changing public attitudes toward greater acceptance of nuclear power. (author)

  19. Small modulator reactors (SMRs) - the key to increased social acceptance of nuclear power?

    International Nuclear Information System (INIS)

    Sam-Aggrey, H.

    2014-01-01

    Small Modular Reactors (SMRs) have been touted as having the potential to spearhead a nuclear renaissance. Proponents of this view cite the relative advantages of SMRs over larger nuclear plants as reason for their optimism. Some of these merits include potential economies of mass production deriving from modular factory production, enhanced safety features, the ability to locate reactors in remote areas, and reduced requirements for technical workforces to install and maintain SMRs. Despite the optimism surrounding SMRs, nuclear power is still a contentious issue for many Canadians. The growth of nuclear power remains adversely affected by concerns over issues such as: the health and safety consequences of nuclear power generation, environmental impacts, proliferation of nuclear materials, terrorism, and radioactive waste disposal. Personal values, past nuclear accidents, and lack of trust in the industry are other factors influencing attitudes toward nuclear energy. Given that public support is pivotal to the growth of the nuclear power sector, the issues surrounding the social acceptance of SMRs should be given serious consideration. Drawing on previously published data, this paper analyses the purported qualities of SMRs with a view of determining whether these characteristics have the potential of changing public attitudes toward greater acceptance of nuclear power. (author)

  20. A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

    Energy Technology Data Exchange (ETDEWEB)

    Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

  1. A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

    Energy Technology Data Exchange (ETDEWEB)

    Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

  2. Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

    OpenAIRE

    Kaiser, Jan Christian

    2012-01-01

    Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES) level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI) 4; 62) severe accidents am...

  3. Technical limits on performance reserves and life expectancy in nuclear power stations with light water reactors

    International Nuclear Information System (INIS)

    Wanner, R.; Brosi, S.; Duijvestijn, G.

    1990-01-01

    The safety margin (i.e. the difference between the loads equipment can take and those actually imposed on components) in a reactor pressure vessel is a major factor in the life expectancy of a nuclear power station. This safety margin is reduced considerably by reductions in the toughness of equipment caused by neutron irradiation and growth of cracks. Once the minimum safety margin is infringed, the nuclear power station is at the end of its working life. 13 figs., 11 refs

  4. Legislative and regulatory aspects of nuclear power reactor licensing in the U.S.A

    International Nuclear Information System (INIS)

    Malsch, M.G.

    1976-01-01

    An explanation of the origins, statutory basis and development of the present regulatory system in the US. A description of the various actions which must be taken by a license applicant and by the USNRC before a nuclear power plant can be constructed and placed on-line. Account of the current regulatory practices followed by the USNRC in licensing nuclear power reactors. (orig./HP) [de

  5. Rock siting of nuclear power plants from a reactor safety standpoint. Status report October 1974

    International Nuclear Information System (INIS)

    1975-01-01

    The aim of this study is to clearify the advantages and disadvantages of an underground nuclear power plant from a reactor safety point of view, compared to a plant above ground. Principles for the technical design of a rock sited BWR nuclear power plant is presented. Also questions of sabotage and closing down the plant at the end of the operational period are treated. (K.K.)

  6. Determination of the lowest critical power levels of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Binh, Do Quang; Nghiem, Huynh Ton; Tuan, Nguyen Minh; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    This paper presents the experimental methods for determining critical states of the Dalat Nuclear Research Reactor containing an extraneous neutron source induced by gamma ray reactions on beryllium in the reactor. The lowest critical power levels are measured at various moments after the reactor is shut down following 100 hours of its continuous operation. Th power levels vary from (0.5-1.2) x 10{sup -4} of P{sub n}, i.e. (25-60)W to (1.1-1.6) x 10{sup -5} of P{sub n}, i.e. (5.5-8)W at corresponding times of 4 days to 13 days after the reactor is shut down. However the critical power must be chosen greater than 500 W to sustain the steady criticality of the reactor for a long time. (author). 3 refs. 4 figs. 1 tab.

  7. International nuclear reactor hazard study. Design and operational features, and hazards of commercial nuclear power reactors in the world. Vol. 1

    International Nuclear Information System (INIS)

    Anderson, R.; Benecke, J.; Donderer, R.; Etemad, S.; Hirsch, H.; Kjellstroem, B.; Large, J.; Scheer, J.; Schneider, M.; Schumacher, H.; Schumacher, O.; Scott, M.; Takagi, J.; Thompson, G.; Torrie, R.; Ziggel, H.

    1986-09-01

    Project management and scientific coordination was performed by Gruppe Oekologie Hannover. The experts from Canada, the Federal Republic of Germany, France, Great Britain, Sweden, USA, and Japan represented an annulated experience of work on nuclear safety. The panel members analyzed the reactor types of their respective countries. The draft reports on the individual reactor types were discussed. Also included are brief descriptions of the different reactor types with simple drawing, evolution of reactor types and a brief report on nuclear power plants in the world. (DG)

  8. Supply of appropriate nuclear technology for the developing world: small power reactors for electricity generation

    International Nuclear Information System (INIS)

    Heising-Goodman, C.D.

    1981-01-01

    This paper reviews the supply of small nuclear power plants (200 to 500 MWe electrical generating capacity) available on today's market, including the pre-fabricated designs of the United Kingdom's Rolls Royce Ltd and the French Alsthom-Atlantique Company. Also, the Russian VVER-440 conventionally built light-water reactor design is reviewed, including information on the Soviet Union's plans for expansion of its reactor-building capacity. A section of the paper also explores the characteristics of LDC electricity grids, reviewing methods available for incorporating larger plants into smaller grids as the Israelis are planning. Future trends in reactor supply and effects on proliferation rates are also discussed, reviewing the potential of the Indian 220 MWe pressurised heavy-water reactor, South Korean and Jananese potential for reactor exports in the Far East, and the Argentine-Brazilian nuclear programme in Latin America. This study suggests that small reactor designs for electrical power production and other applications, such as seawater desalination, can be made economical relative to diesel technology if traditional scaling laws can be altered by adopting and standardising a pre-fabricated nuclear power plant design. Also, economy can be gained if sufficient attention is concentrated on the design, construction and operating experience of suitably sized conventionally built reactor systems. (author)

  9. 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)

  10. The software safety analysis based on SFTA for reactor power regulating system in nuclear power plant

    International Nuclear Information System (INIS)

    Liu Zhaohui; Yang Xiaohua; Liao Longtao; Wu Zhiqiang

    2015-01-01

    The digitalized Instrumentation and Control (I and C) system of Nuclear power plants can provide many advantages. However, digital control systems induce new failure modes that differ from those of analog control systems. While the cost effectiveness and flexibility of software is widely recognized, it is very difficult to achieve and prove high levels of dependability and safety assurance for the functions performed by process control software, due to the very flexibility and potential complexity of the software itself. Software safety analysis (SSA) was one way to improve the software safety by identify the system hazards caused by software failure. This paper describes the application of a software fault tree analysis (SFTA) at the software design phase. At first, we evaluate all the software modules of the reactor power regulating system in nuclear power plant and identify various hazards. The SFTA was applied to some critical modules selected from the previous step. At last, we get some new hazards that had not been identified in the prior processes of the document evaluation which were helpful for our design. (author)

  11. Wishful thinking and real problems: Small modular reactors, planning constraints, and nuclear power in Jordan

    International Nuclear Information System (INIS)

    Ramana, M.V.; Ahmad, Ali

    2016-01-01

    Jordan plans to import two conventional gigawatt scale nuclear reactors from Russia that are expensive and too large for Jordan's current electricity grid. Jordan efforts to establish nuclear power might become easier in some ways if the country were to construct Small Modular Reactors, which might be better suited to Jordan's financial capabilities and its smaller electrical grid capacity. But, the SMR option raises new problems, including locating sites for multiple reactors, finding water to cool these reactors, and the higher cost of electricity generation. Jordan's decision has important implications for its energy planning as well as for the market for SMRs. - Highlights: •Jordan is planning to purchase two large reactors from Russia. •Large reactors would be inappropriate to Jordan's small electricity grid. •Small modular reactors would be more appropriate to Jordan's grid, but have problems. •The market for small modular reactors will be smaller than often projected. •Jordan should consider the financial impact of building a large nuclear reactor.

  12. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Huber, Horacio

    1989-01-01

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

  13. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

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

  16. NUCLEAR REACTOR

    Science.gov (United States)

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

    1958-01-21

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

  17. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  18. Computerization of the nuclear material accounting system for safeguards purposes at nuclear power plants with WWER-440 reactors

    International Nuclear Information System (INIS)

    Antonov, V.P.; Konnov, Yu.I.; Semenets, A.N.

    1983-01-01

    The paper sets forth the basic principles underlying nuclear material accounting at nuclear power plants with WWER-440 reactors. It briefly describes the general structure and individual units in a program for computerized accounting. The use of this program is illustrated by the actual accounting data from the fifth unit of the Novovoronezh nuclear power station. The NUMIS program seems to be of interest both for the purposes of IAEA safeguards and for nuclear power plant operators in countries where power plants with WWER-440 reactors subject to IAEA safeguards are either in operation or under construction. The research in question was conducted initially under an IAEA research contract; the system is now being developed further and tested under the IAEA-USSR technical and scientific co-operation programme on safeguards. (author)

  19. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

  20. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

    International Nuclear Information System (INIS)

    Pathania, R.S.; Tatone, O.S.

    1979-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

  1. Design considerations for micro nuclear reactors to supply power to off-grid mines

    Energy Technology Data Exchange (ETDEWEB)

    Gihm, B.; Cooper, G.; Morettin, D.; De Koning, P., E-mail: bgihm@hatch.ca [Hatch Ltd., Mississauga, Ontario (Canada); Carreau, M. [Hatch Ltd., Montreal, Quebec (Canada); Sarvinis, J. [Hatch Ltd., Mississauga, Ontario (Canada)

    2014-07-01

    Nuclear technology vendors have been proposing to develop small scale nuclear reactors to supply power and heat to remote industrial operations such as a mining site. Based on extensive experience in integrating different power generation technologies with captive mining power systems, Hatch examined the technical requirements of small scale nuclear reactor application in remote mine power generation. Mining power systems have unique characteristics and challenges that set them apart from utility grid connected power systems. Key examples of such unique characteristics are: A small number of large motor loads such as hoists, pumps, shovels, pumps and crushers represent a large fraction of the peak load. These equipment may cause significant load fluctuations and put the power systems under high stress; There is no organic demand growth (i.e., the load growth occurs as a step increase); and, The extreme environmental conditions and remoteness of the sites introduce a set of operational challenges and require specialized planning. This paper presents real remote mine operation data to demonstrate the load profile of remote mining sites. The operation characteristics and performance requirements of diesel reciprocating engines are discussed, which have to be matched or exceeded by a small scale nuclear power plant if it is to be a viable technical alternative to diesel power. The power quality control options from wind power integration in isolated grids are discussed as a parallel can be drawn between wind and nuclear power application in remote mine power systems. Finally the authors provided a list of technical constraints and design considerations for very small modular reactor development. (author)

  2. Design considerations for micro nuclear reactors to supply power to off-grid mines

    International Nuclear Information System (INIS)

    Gihm, B.; Cooper, G.; Morettin, D.; De Koning, P.; Carreau, M.; Sarvinis, J.

    2014-01-01

    Nuclear technology vendors have been proposing to develop small scale nuclear reactors to supply power and heat to remote industrial operations such as a mining site. Based on extensive experience in integrating different power generation technologies with captive mining power systems, Hatch examined the technical requirements of small scale nuclear reactor application in remote mine power generation. Mining power systems have unique characteristics and challenges that set them apart from utility grid connected power systems. Key examples of such unique characteristics are: A small number of large motor loads such as hoists, pumps, shovels, pumps and crushers represent a large fraction of the peak load. These equipment may cause significant load fluctuations and put the power systems under high stress; There is no organic demand growth (i.e., the load growth occurs as a step increase); and, The extreme environmental conditions and remoteness of the sites introduce a set of operational challenges and require specialized planning. This paper presents real remote mine operation data to demonstrate the load profile of remote mining sites. The operation characteristics and performance requirements of diesel reciprocating engines are discussed, which have to be matched or exceeded by a small scale nuclear power plant if it is to be a viable technical alternative to diesel power. The power quality control options from wind power integration in isolated grids are discussed as a parallel can be drawn between wind and nuclear power application in remote mine power systems. Finally the authors provided a list of technical constraints and design considerations for very small modular reactor development. (author)

  3. Application of powder metallurgy in production of nuclear fuels for research and power reactors

    International Nuclear Information System (INIS)

    Fukuda, Kosaku

    2000-01-01

    Powder metallurgy has been applied in many of the processes of nuclear fuel fabrication, which has contributed, to a great progress of the nuclear technology to date. Evolution of nuclear fuels still continues to meet various emerging demands in terms of enhanced safety, economical effectiveness, non-proliferation and environmental mitigation. This paper reviews recent progress of nuclear fuels of research and power reactors, in particular, focusing on the powder metallurgy application. First, the review is made on plate type fuels for research reactors, inter alia, silicide fuel which is prevailing worldwide from the viewpoint of non-proliferation. The relation between fabrication and irradiation behavior is also discussed. Next, oxide fuels including MOX are reviewed. Recent interests of UO 2 are directed toward large grain pellets and burnable absorber pellets, both of which arise from requirement of extended burnup. Finally, the MOX fuel for thermal reactors is reviewed. (author)

  4. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  6. Energy from nuclear reactors

    International Nuclear Information System (INIS)

    Hospe, J.

    1977-01-01

    This VDI-Nachrichten series has the target to provide a technical-objective basis for the discussion of the pros and cons of nuclear power. The first part deals with LWR-type reactors which so far have prevailed in nuclear power generation. (orig.) [de

  7. Thermal hydraulic aspects of uncertainty in power measurement of nuclear reactors

    International Nuclear Information System (INIS)

    Gupta, S.K.; Kumar, Rajesh; Gaikwad, A.J.; Majumdar, P.; Agrawal, R.A.

    2004-01-01

    Power measurement in Nuclear Reactors is carried out through in-core and ex-core neutron monitors which are continuously calibrated against thermal power. In Indian Pressurized Heavy Water Reactors (220 MWe) the temperature difference across steam generator hot and cold legs is taken to be a measure of thermal power as the flow through the primary heat transport system is assumed to be constant through out is operation. Gross flow is not measured directly. However, the flow depends on the characteristics of the primary heat transport pumps, which are centrifugal type and are affected by the grid frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable frequency. The paper quantifies the percentage increase in the reactor power for the sustained allowable high grid frequency. This uncertainty is in addition to instrument inaccuracy and should be accounted for in safety analysis. In some reactors thermal power is calculated from stem flow rate and pressure, here the location of steam flow measurement is important to avoid leakage related error in thermal power. Neutron absorption cross section in the power measurement instruments and the power production in the fuel varies with neutron energy levels, these aspects are also discussed in the paper. (author)

  8. Reactor units for power supply to the Russian Arctic regions: Priority assessment of nuclear energy sources

    Directory of Open Access Journals (Sweden)

    Mel'nikov N. N.

    2017-03-01

    Full Text Available Under conditions of competitiveness of small nuclear power plants (SNPP and feasibility of their use to supply power to remote and inaccessible regions the competition occurs between nuclear energy sources, which is caused by a wide range of proposals for solving the problem of power supply to different consumers in the decentralized area of the Russian Arctic power complex. The paper suggests a methodological approach for expert assessment of the priority of small power reactor units based on the application of the point system. The priority types of the reactor units have been determined based on evaluation of the unit's conformity to the following criteria: the level of referentiality and readiness degree of reactor units to implementation; duration of the fuel cycle, which largely determines an autonomy level of the nuclear energy source; the possibility of creating a modular block structure of SNPP; the maximum weight of a transported single equipment for the reactor unit; service life of the main equipment. Within the proposed methodological approach the authors have performed a preliminary ranking of the reactor units according to various criteria, which allows quantitatively determining relative difference and priority of the small nuclear power plants projects aimed at energy supply to the Russian Arctic. To assess the sensitivity of the ranking results to the parameters of the point system the authors have observed the five-point and ten-point scales under variations of importance (weights of different criteria. The paper presents the results of preliminary ranking, which have allowed distinguishing the following types of the reactor units in order of their priority: ABV-6E (ABV-6M, "Uniterm" and SVBR-10 in the energy range up to 20 MW; RITM-200 (RITM-200M, KLT-40S and SVBR-100 in the energy range above 20 MW.

  9. Preparation of processed nuclear data libraries for thermal, fast and fusion research and power reactor applications

    International Nuclear Information System (INIS)

    Ganesan, S.

    1994-03-01

    A Consultants Meeting on ''Preparation of Processed Nuclear Data Libraries for Thermal, Fast and Fusion Research and Power Reactor Applications'' was convened by the International Atomic Energy Agency and held during December 13-16, 1993 December 8-10, 1993 at the IAEA Headquarters, Vienna. The detailed agenda, the complete list of participants and the recommendations are presented in this report. (author)

  10. Temperature conditions of foundation plates under nuclear power plant reactor compartments

    International Nuclear Information System (INIS)

    Ehsaulov, S.L.

    1990-01-01

    Method for calculation of temperature conditions for foundation plates under reactor compartments located in the main building, used in construction of the second stage of the Kostroma nuclear power plant, is considered. The obtained calculation data can be used for determining the most suitable period of concrete placement, composition, initial temperature, manufacturing technology and ways of delivery of concrete mixture

  11. Application of associative emulator neural network for power control of nuclear reactor

    International Nuclear Information System (INIS)

    Datta, A.K.; Bandyopadhyay, Somnath

    1993-01-01

    This paper addresses the question of how to perform on-line training of emulator neural network for power control in a nuclear reactor. The computation and convergence problem can be reduced by judicious choice of bidirectional associative recall. (author). 10 refs., 2 figs

  12. Introduction to Nuclear Fusion Power and the Design of Fusion Reactors. An Issue-Oriented Module.

    Science.gov (United States)

    Fillo, J. A.

    This three-part module focuses on the principles of nuclear fusion and on the likely nature and components of a controlled-fusion power reactor. The physical conditions for a net energy release from fusion and two approaches (magnetic and inertial confinement) which are being developed to achieve this goal are described. Safety issues associated…

  13. Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-01-01

    Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

  14. The outlook for application of powerful nuclear thermionic reactor -powered space electric jet propulsion engines

    International Nuclear Information System (INIS)

    Semyonov, Y.P.; Bakanov, Y.A.; Synyavsky, V.V.; Yuditsky, V.D.

    1997-01-01

    This paper summarizes main study results for application of powerful space electric jet propulsion unit (EJPUs) which is powered by Nuclear Thermionic Power Unit (NTPU). They are combined in Nuclear Power/Propulsion Unit (NPPU) which serves as means of spacecraft equipment power supply and spacecraft movement. Problems the paper deals with are the following: information satellites delivery and their on-orbit power supply during 10-15 years, removal of especially hazardous nuclear wastes, mining of asteroid resources and others. Evaluations on power/time/mass relationship for this type of mission are given. EJPU parameters are compatible with Russian existent or being under development launch vehicle. (author)

  15. The reactor power control system based on digital control in nuclear power plant

    International Nuclear Information System (INIS)

    Liu Chong; Zhou Jianliang; Tan Ping

    2010-01-01

    The PLC (Programmable Logical Controller), digital communication and redundant techniques are applied in the rod control and position indication system(namely the reactor power control system) to perform the power control in the 300 MW reactor automatically and integrally in Qinshan Phase I project. This paper introduces the features, digital design methods of hardware of the instrumentation and control system (I and C) in the reactor power control. It is more convenient for the information exchange by human-machine interface (HMI), operation and maintenance, and the system reliability has been greatly improved after the project being reconstructed. (authors)

  16. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  18. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto, E-mail: alan@lmp.ufrj.br, E-mail: andressa@lmp.ufrj.br, E-mail: schirru@lmp.ufrj.br, E-mail: ffreire@eletronuclear.gov.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Rio de Janeiro, RJ (Brazil)

    2017-11-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10{sup 13} combinations and 10{sup 11} great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  19. Optimization of reload of nuclear power plants using ACO together with the GENES reactor physics code

    International Nuclear Information System (INIS)

    Lima, Alan M.M. de; Freire, Fernando S.; Nicolau, Andressa S.; Schirru, Roberto

    2017-01-01

    The Nuclear reload of a Pressurized Water Reactor (PWR) occurs whenever the burning of the fuel elements can no longer maintain the criticality of the reactor, that is, it cannot maintain the Nuclear power plant operates within its nominal power. Nuclear reactor reload optimization problem consists of finding a loading pattern of fuel assemblies in the reactor core in order to minimize the cost/benefit ratio, trying to obtain maximum power generation with a minimum of cost, since in all reloads an average of one third of the new fuel elements are purchased. This loading pattern must also satisfy constraints of symmetry and security. In practice, it consists of the placing 121 fuel elements in 121 core positions, in the case of the Angra 1 Brazilian Nuclear Power Plant (NPP), making this new arrangement provide the best cost/benefit ratio. It is an extremely complex problem, since it has around 1% of great places. A core of 121 fuel elements has approximately 10"1"3 combinations and 10"1"1 great locations. With this number of possible combinations it is impossible to test all, in order to choose the best. In this work a system called ACO-GENES is proposed in order to optimization the Nuclear Reactor Reload Problem. ACO is successfully used in combination problems, and it is expected that ACO-GENES will show a robust optimization system, since in addition to optimizing ACO, it allows important prior knowledge such as K infinite, burn, etc. After optimization by ACO-GENES, the best results will be validated by a licensed reactor physics code and will be compared with the actual results of the cycle. (author)

  20. Econometric modelling of certain nuclear power systems based on thermal and fast breeder reactors

    International Nuclear Information System (INIS)

    Pavelescu, M.; Pioaru, C.; Ursu, I.

    1988-01-01

    Certain known economic analysis models for a LMFBR fast breeder and CANDU thermal solitary reactors are presented, based on the concepts of discounting and levelization. These models are subsequently utilized as a basis for establishing an original model for the econometric analysis of certain thermal reactor systems or/and fast breeder reactors. Case studies are subsequently conducted with the systems: 1-CANDU, 2-LMFBR, 3-CANDU + LMFBR which enables us to draw certain interesting conclusions for a long range nuclear power policy. (author)

  1. Fast breeder reactors insertion in a D2O - natural U nuclear power plants park

    International Nuclear Information System (INIS)

    Gho, C.J.

    1985-01-01

    A model for the evolution of Argentine's installed nuclear power for the next 40 years is presented. The consequences of fast breeder reactors' introduction are studied in both autarchic Pu cycle and a limited reprocessing system. The passage of a reactor park like the national, of natural U - heavy water to one of fast breeder reactors, can only be obtained in a very long term due, fundamentally, to the need of Pu produced for those to feed the last ones. (M.E.L.) [es

  2. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    1997-01-01

    The purpose of the dissertation is to develop real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification plant transients (with and without scram). For this erps, probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents. The real - time information during transients and accidents can be obtained to assess the operator in his decision - making. Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. 5-15 figs., 42 refs

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  4. Nuclear power in Asia

    Energy Technology Data Exchange (ETDEWEB)

    Hagen, Ronald E.

    1998-08-01

    Contains Executive Summary and Chapters on: Nuclear Energy in the Asian context; Types of nuclear power reactors used in Asia; A survey of nuclear power by country; The economics of nuclear power; Fuels, fuel cycles and reprocessing; Environmental issues and waste disposal; The weapons issues and nuclear power; Conclusions. (Author)

  5. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  6. Advanced light water reactor program at ABB-Combustion Engineering Nuclear Power

    International Nuclear Information System (INIS)

    Cahn, H.

    1990-01-01

    To meet the needs of Electric Utilities ordering nuclear power plants in the 1990s, ABB-Combustion Engineering is developing two designs which will meet EPRI consensus requirements and new licensing issues. The System 80 Plus design is an evolutionary pressurized water reactor plant modelled after the successful System 80 design in operation in Palo Verde and under construction in Korea. System Plus is currently under review by the US Nuclear Regulatory Commission with final design approval expected in 1991 and design certification in 1992. The Safe Integral Reactor (SIR) plant is a smaller facility with passive safety features and modular construction intended for design certification in the late 1990s. (author)

  7. Considerations on safety against seismic excitations in the project of reactor auxiliary building and control building in nuclear power plants

    International Nuclear Information System (INIS)

    Santos, S.H.C.; Castro Monteiro, I. de

    1986-01-01

    The seismic requests to be considered in the project of main buildings of a nuclear power plant are discussed. The models for global seismic analysis of nuclear power plant structures, as well as models for global strength distribution are presented. The models for analysing reactor auxiliary building and control building, which together with the reactor building and turbine building form the main energy generation complex in a nuclear power plant, are described. (M.C.K.) [pt

  8. Catalogue and classification of technical safety standards, rules and regulations for nuclear power reactors and nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Fichtner, N.; Becker, K.; Bashir, M.

    1977-01-01

    The present report is an up-dated version of the report 'Catalogue and Classification of Technical Safety Rules for Light-water Reactors and Reprocessing Plants' edited under code No EUR 5362e, August 1975. Like the first version of the report, it constitutes a catalogue and classification of standards, rules and regulations on land-based nuclear power reactors and fuel cycle facilities. The reasons for the classification system used are given and discussed

  9. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    Baptista, Vinicius Damas

    1996-01-01

    The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

  10. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  11. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y H; Cha, K H; Lee, S H [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  12. Application of robust fuzzy control in power control of nuclear reactor

    International Nuclear Information System (INIS)

    Liu Lei; Luan Xiuchun; Jin Guangyuan; Yu Tao; Rao Su

    2013-01-01

    Robust-fuzzy controller based on T-S fuzzy model was designed for real-time controlling of nuclear reactor power and adapting to the load changing of power grid. Local controller was designed by means of state feedback technique, and the global controller was designed by parallel distributed compensation (PDC) method. The result of solving linear matrix inequalities (LMI) proves that this controller is stable. The simulation shows that the nuclear power can be well controlled in three typical conditions by this controller. (authors)

  13. An integrated nuclear reactor unit for a floating low capacity nuclear power plant designed for power supply in remote areas with difficult access

    International Nuclear Information System (INIS)

    Achkasov, A.N.; Grechko, G.I.; Gladkov, O.G.; Pavlov, V.L.; Pepa, V.N.; Shishkin, V.A.

    1997-01-01

    The paper describes the conceptual design of an integrated advanced safety nuclear reactor unit for a low capacity floating, NPP designed for power supply in areas which are remote with difficult access. The paper describes the major structural and lay-out components of the steam generator and reactor units with main technical characteristics. (author)

  14. The basic concepts of a fuel-power detector for nuclear power reactors

    International Nuclear Information System (INIS)

    Lynch, G.F.

    1979-01-01

    Fuel power is proposed as an alternative to neutron or gamma-ray flux for control and safety functions in CANDU power reactors. To satisfy in-core power monitoring requirements, a detector whose dynamic response corresponds to the heat production rate in the fuel is needed. This report explores the concept of tailoring the response characteristics of a mixed-response self-powered flux detector to match the requirements of an ideal fuel-power detector. (author)

  15. In-core fuel management programs for nuclear power reactors

    International Nuclear Information System (INIS)

    1984-10-01

    In response to the interest shown by Member States, the IAEA organized a co-ordinated research programme to develop and make available in the open domain a set of programs to perform in-core fuel management calculations. This report summarizes the work performed in the context of the CRP. As a result of this programme, complete in-core fuel management packages for three types of reactors, namely PWR's, BWR's and PHWR are now available from the NEA Data Bank. For some reactor types, these program packages are available with three levels of sophistication ranging from simple methods for educational purposes to more comprehensive methods that can be used for reactor design and operation. In addition some operating data have been compiled to allow code validation. (author)

  16. 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.)

  17. Nuclear reactor control method for maintaining an appreciably constant axial distribution of power with load variations

    International Nuclear Information System (INIS)

    Morita, Toshio.

    1975-01-01

    A nuclear reactor control method is described in which the power variations of the reactor are controlled partly by varying the concentration of the neutron absorbing element and partly by varying the positions of the control rods, in order to maintain the axial distribution of power appreciably symmetrical during the normal operation of the reactor. The control points are located in the upper and lower halves of the core. The controls are operated to maintain the output power difference between the upper and lower halves of the core, based on the total output power (axial deviation) significantly equal to a predetermined optimum figure during the entire running of the reactor, including when there are power variations. The optimum value is obtained by determining the axial deviation at full power with the xenon in balance and all the control rods withdrawn from the fuel area of the core. This optimum value is recalculated after a period appreciably equal to that of a month's operation at full power. This method applies in particular to PWR type reactors [fr

  18. Decommissioning and decontamination of licensed reactor facilities and demonstration nuclear power plants

    International Nuclear Information System (INIS)

    Lear, G.; Erickson, P.B.

    1975-01-01

    Decommissioning of licensed reactors and demonstration nuclear power plants has been accomplished by mothballing (protective storage), entombment, and dismantling or a combination of these three. The alternative selected by a licensee seems to be primarily based on cost. A licensee must, however, show that the decommissioning process provides adequate protection of the health and safety of the public and no adverse impact on the environment. To date the NRC has approved each of the alternatives in the decommissioning of different facilities. The decommissioning of small research reactors has been accomplished primarily by dismantling. Licensed nuclear power plants, however, have been decommissioned primarily by being placed in a mothballed state in which they continue to retain a reactor license and the associated licensee responsibilities

  19. Knowledges and abilities catalog for nuclear power plant operators: pressurized water reactors

    International Nuclear Information System (INIS)

    1985-07-01

    This document catalogs roughly 5300 knowledges and abilities of reactor operators and senior reactor operators. It results from a reanalysis of much larger job-task analysis data base compiled by the Institute of Nuclear Power Operations (INPO). Knowledges and abilities are cataloged for 45 major power plant systems and 38 emergency evolutions, grouped according to 11 fundamental safety functions (e.g., reactivity control and reactor coolant system inventory control). With appropriate sampling from this catalog, operator licensing examinations having content validity can be developed. A structured sampling procedure for this catalog is under development by the Nuclear Regulatory Commission (NRC) and will be published as a companion document, ''Examiners' Handbook for Developing Operator Licensing Examinations'' (NUREG-1121). The examinations developed by using the catalog and handbook will cover those topics listed under Title 10, Code of Federal Regulations, Part 55

  20. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  1. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  2. The contract of design of an atomic reactor system in Yonggwang - 5, 6 nuclear power plant

    International Nuclear Information System (INIS)

    1995-05-01

    This is a contract of design of an atomic reactor system in Yonggwang 5, 6 nuclear power plant. It has the general contract condition. In the appendix, it indicates the detail regulations between two parties which are the coverage of division on the responsibility, schedule of the delivery, standard of the technology, guarantee, drawing and paper support of the Korea Electric Power Corporation, support of technology drill, test, regulations of code and standard and list of items and prices.

  3. Fuel balance in nuclear power with fast reactors without a uranium blanket

    International Nuclear Information System (INIS)

    Naumov, V.V.; Orlov, V.V.; Smirnov, V.S.

    1994-01-01

    General aspects related to replacing the uranium blanket of a lead-cooled fast reactor burning uranium-plutonium nitride fuel with a more efficient lead reflector are briefly discussed in the article. A study is very briefly summarized, which showed that a breeding ratio of about 1 and electric power of about 300 MW were achievable. A nuclear fuel balance is performed to estimate the increased consumption of uranium to produce power and the gains achievable by eliminating the uranium blanket. Elimination of the uranium blanket has the advantages of simplifying and improving the fast reactor and eliminating the production of weapons quality plutonium. 3 figs

  4. Nuclear power - replacement of pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

  5. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendices VII, VIII, IX, and X. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy.

  6. Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendices VII, VIII, IX, and X

    International Nuclear Information System (INIS)

    1975-10-01

    Information is presented concerning the release of radioactivity in reactor accidents; physical processes in reactor meltdown accidents; safety design rationale for nuclear power plants; and design adequacy

  7. EPRI's nuclear power plant instrumentation and control program and its applicability to advanced reactors

    International Nuclear Information System (INIS)

    Naser, J.; Torok, R.; Wilkinson, D.

    1997-01-01

    I ampersand C systems in nuclear power plants need to be upgraded over the lifetime of the plant in a reliable and cost-effective manner to replace obsolete equipment, to reduce O ampersand M costs, to improve plant performance, and to maintain safety. This applies to operating plants now and will apply to advanced reactors in the future. The major drivers for the replacement of the safety, control, and information systems in nuclear power plants are the obsolescence of the existing hardware and the need for more cost-effective power production. Competition between power producers is dictating more cost-effective power production. The increasing O ampersand M costs to maintain systems experiencing obsolescence problems is counter to the needs for more cost-effective power production and improved competitiveness. This need for increased productivity applies to government facilities as well as commercial plants. Increasing competition will continue to be a major factor in the operation of both operating plants and advanced reactors. It will continue to dictate the need for improved productivity and cost-effectiveness. EPRI and its member nuclear utilities are working together on an industry wide I ampersand C Program to address I ampersand C issues and to develop cost-effective solutions. A majority of the I ampersand C products and demonstrations being developed under this program will benefit advanced reactors in both the design and operational phases of their life cycle as well as it will benefit existing plants. 20 refs

  8. Lightweight power bus for a baseload nuclear reactor in space

    International Nuclear Information System (INIS)

    Oberly, C.E.; Massie, L.D.; Hoffman, D.J.

    1989-01-01

    Space environmental interactions with the power distribution/power processing subsystem can become a serious problem for power systems rated at 10's to 100's of kilowatts. Utilization of ceramic superconductors at 1000 A/cm/sup 2/, which has already been demonstrated at 77 K in a conductor configuration may eliminate both bus mass and distribution voltage problems in a high power satellite. The analytical results presented here demonstrate that a superconducting coaxial power transmission bus offers significant benefits in reduced distribution voltage and mass

  9. Nuclear power plant control room task analysis. Pilot study for pressurized water reactors

    International Nuclear Information System (INIS)

    Barks, D.B.; Kozinsky, E.J.; Eckel, S.

    1982-05-01

    The purposes of this nuclear plant task analysis pilot study: to demonstrate the use of task analysis techniques on selected abnormal or emergency operation events in a nuclear power plant; to evaluate the use of simulator data obtained from an automated Performance Measurement System to supplement and validate data obtained by traditional task analysis methods; and to demonstrate sample applications of task analysis data to address questions pertinent to nuclear power plant operational safety: control room layout, staffing and training requirements, operating procedures, interpersonal communications, and job performance aids. Five data sources were investigated to provide information for a task analysis. These sources were (1) written operating procedures (event-based); (2) interviews with subject matter experts (the control room operators); (3) videotapes of the control room operators (senior reactor operators and reactor operators) while responding to each event in a simulator; (4) walk-/talk-throughs conducted by control room operators for each event; and (5) simulator data from the PMS

  10. Brazilian nuclear power plants decommissioning plan for a multiple reactor site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R., E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas. Programa de Pos-Graduacao em Energia e Engenharia da Energia

    2015-07-01

    Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

  11. Brazilian nuclear power plants decommissioning plan for a multiple reactor site

    International Nuclear Information System (INIS)

    Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R.

    2015-01-01

    Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

  12. Utilization of nuclear energy for generating electric power in the FRG, with special regard to LWR-type reactors

    International Nuclear Information System (INIS)

    Vollradt, J.

    1977-01-01

    Comments on interdependencies in energy industry and energy generation as seen by energy supply utilities, stating that the generation of electric power in Germany can only be based on coal and nuclear energy in the long run, are followed by the most important, fundamental, nuclear-physical, technological and in part political interdependencies prevailing in the starting situation of 1955/58 when the construction of nuclear power plant reactors began. Then the development ranging to the 28000 MW nuclear power output to be expected in 1985 is outlined, totalling in 115000 MW electric power in the FRG. Finally, using the respectively latest order, the technical set up of each of the reactor types with 1300 MWe unit power offered by German manufacturers are described: BBC/BBR PWR-type reactor Neupotz, KWU-PWR-type reactor Hamm and KWU PWR-type reactor double unit B+C Gundremmingen. (orig.) [de

  13. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  14. Problems and legislative remedies of the parallel law systems in Japan for nuclear power reactors

    International Nuclear Information System (INIS)

    Irie, Kazutomo

    2011-01-01

    There are two established laws governing nuclear power reactors in Japan. One is the Electricity Utilities Industry Law, which regulates the nuclear power reactors, and the other is the so-called 'Reactor Regulation Law', which dually regulates the reactors in some phases. When a graded approach on the regulation of nuclear reactors was adopted, it extended over these two laws and was legislatively imperfect. Such imperfection created problems from the beginning. Also, the original regulatory structures presented by these laws had become obscure during the operation process of the graded regulation. The situation becomes further complicated by the revision of these laws in recent years. It appears that the trait of the regulatory procedural structure of the Electricity Utilities Industry Law has been weakened. As there is a pressing need to review the entire regulatory structure and to propose a unified regulatory system by combining these laws, this paper examines the merits and demerits of combining these laws under a unified regulation. (author)

  15. Multi-reactor power system configurations for multimegawatt nuclear electric propulsion

    Science.gov (United States)

    George, Jeffrey A.

    1991-01-01

    A modular, multi-reactor power system and vehicle configuration for piloted nuclear electric propulsion (NEP) missions to Mars is presented. Such a design could provide enhanced system and mission reliability, allowing a comfortable safety margin for early manned flights, and would allow a range of piloted and cargo missions to be performed with a single power system design. Early use of common power modules for cargo missions would also provide progressive flight experience and validation of standardized systems for use in later piloted applications. System and mission analysis are presented to compare single and multi-reactor configurations for piloted Mars missions. A conceptual design for the Hydra modular multi-reactor NEP vehicle is presented.

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  18. On the selfacting safe limitation of fission power and fuel temperature in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Scherer, W.; Brockmann, H.; Drecker, S.; Gerwin, H.; Haas, K.A.; Kugeler, K.; Ohlig, U.; Ruetten, H.J.; Teuchert, E.; Werner, H.; Wolf, L.

    1994-08-01

    Nuclear energy probably will not contribute significantly to the future worldwide energy supply until it can be made catastrophe-free. Therefore it has to be shown, that the consequences of even largest accidents will have no major impact to the environment of a power plant. In this paper one of the basic conditions for such a nuclear technology is discussed. Using mainly the modular pebble-bed high-temperature reactor as an example, the design principles, analytical methods and the level of knowledge as given today in controlling reactivity accidents by inherent safety features of innovative nuclear reactors are described. Complementary possibilities are shown to reach this goal with systems of different types of construction. Questions open today and resulting requirements for future activities are discussed. Today's knowledge credibly supports the possibility of a catastrophe-free nuclear technology with respect to reactivity events. (orig.)

  19. Main research results in the field of nuclear power engineering of the Nuclear Reactors and Thermal Physics Institute in 2014

    International Nuclear Information System (INIS)

    Trufanov, A.A.; Orlov, Yu.I.; Sorokin, A.P.; Chernonog, V.L.

    2015-01-01

    The main results of scientific and technological activities for last years of the Nuclear Reactors and Thermal Physics Institute FSUE SSC RF - IPPE in solving problems of nuclear power engineering are presented. The work have been carried out on the following problems: justification of research and development solutions and safety of NPPs with fast reactors of new generation with sodium (BN-1200, MBIR) and lead (BREST-OD-300) coolants, justification of safety of operating and advanced NPPs with WWER reactor facilities (WWER-1000, AEhS 2006, WWER-TOI), development and benchmarking of computational codes, research and development support of Beloyarsk-3 (BN-600) and Bilibino (BN-800) NPPs operation, decommissioning of AM and BR-10 research reactors, pilot scientific studies (WWER-SKD, ITER), international scientific and technical cooperation. Problems for further investigations are charted [ru

  20. Life time of nuclear power plants and new types of reactors

    International Nuclear Information System (INIS)

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  1. Small and medium size nuclear power reactors for desalination

    International Nuclear Information System (INIS)

    Raisic, N.; Goodman, E.I.

    1976-01-01

    Taking the water needs, e.g. of some of the world's major towns, as a basis, it is investigated whether nuclear energy can be utilized economically for desalination. When a certain distance for the transport of water from other regions is exceeded it is quite possible that nuclear desalination becomes economical. Taking the example of Honkong, it is shown that this method can find application for other reasons, too, e.g. if the need exceeds the possibilities there are of meeting this need from natural sources. (UA) [de

  2. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    Science.gov (United States)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  3. 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)

  4. HTGR nuclear power plants: features of the VGR-50 high temperature reactor

    International Nuclear Information System (INIS)

    Glebov, V.P.; Bogoyavlenskii, R.G.; Glushkov, E.S.; Grebennik, V.N.; Ponomarev-Stepnoi, N.N.; Vinogradov, V.P.

    1983-01-01

    Current developmental trends in the power industry are guided to an appreciable extent by the increasing shortages of fossil fuels (coal, petroleum, natural gas) and by ecological problems. Assuming a continuing trend in worldwide consumption of energy resources, we see the electric power industry using up 20%, the other 80% (petroleum, coal, natural gas) going into generating industrial process heat and space heat, transportation, the chemical processing industry, the metallurgical industry, and other branches of industry. In the future, nuclear power will have the job of not only meeting the needs of the electric power industry, but also generating process heat. The most promising type of nuclear power plant available for solving complex problems in generation of electric power and heat for technological processes in the metallurgical processing industry and chemical processing industry is the one based around high-temperature reactors

  5. Shield materials recommended for space power nuclear reactors

    Science.gov (United States)

    Kaszubinski, L. J.

    1973-01-01

    Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  7. 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

  8. 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.

  9. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs.

  10. Development of public acceptance and market success with very small nuclear power reactors (VSR)

    International Nuclear Information System (INIS)

    Lang, P

    2008-01-01

    This paper presents a case for public acceptance of very small nuclear reactors (VSR) as a socially welcomed, locally produced source of energy. It is argued that the subconscious associations and emotional responses that the public makes with size and appearance will often outweigh technical or economic advantages as the determining factor in their ultimate acceptance or rejection of a new power plant. The unique technical features of VSR effectively refute the three most commonly cited reasons for opposing nuclear power: cost, safety, and waste management. VSR also address the demands of environmentalists for small-scale, clean distributed energy production. (author)

  11. Draft report on compilation of generic safety issues for light water reactor nuclear power plants

    International Nuclear Information System (INIS)

    1997-07-01

    A generally accepted approach to characterizing the safety concerns in nuclear power plants is to express them as safety issues which need to be resolved. When such safety issues are applicable to a generation of plants of a particular design or to a family of plants of similar design, they are termed generic safety issues. Examples of generic safety issues are those related to reactor vessel embrittlement, control rod insertion reliability or strainer clogging. The safety issues compiled in this document are based on broad international experience. This compilation is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. Refs

  12. The Perspective of Small and Medium Size Nuclear Power Reactors in the Brazilian Isolated Electricity Grids

    International Nuclear Information System (INIS)

    Moutinho dos Santos, E.

    1998-01-01

    As a consequence of the international crisis that has shaken the nuclear industry almost all over the world during the late 1980s, new academic contributions, which could settle a new basis for the nuclear energy in Brazil, have been scarce. This paper aims to partially fulfill this gap. We discuss some perspectives for the nuclear option to regain some prominence in the Brazilian energy matrix. Recent developments in the nuclear industry, including advances in plant design, have been opening interesting markets for small and medium nuclear power reactors (SMNPRs). We access the suitability and feasibility of such new technology in the Brazilian isolated electricity grids. We conclude by saying that, although the difficulties are still huge, SMNPRs may be a good strategy for Brazil to revitalize its nuclear policy. (author)

  13. Nuclear power development

    International Nuclear Information System (INIS)

    Povolny, M.

    1980-01-01

    The development and uses of nuclear power in Czechoslovakia and other countries are briefly outlined. In the first stage, the Czechoslovak nuclear programme was oriented to the WWER 440 type reactor while the second stage of the nuclear power plant construction is oriented to the WWER 10O0 type reactor. It is envisaged that 12 WWER 440 type reactors and four to five WWER 1000 type reactors will be commissioned till 1990. (J.P.)

  14. Thorium fuel for light water reactors - reducing proliferation potential of nuclear power fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Galperin, A; Radkowski, A [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

    1996-12-01

    The proliferation potential of the light water reactor fuel cycle may be significantly reduced by utilization of thorium as a fertile component of the nuclear fuel. The main challenge of Th utilization is to design a core and a fuel cycle, which would be proliferation-resistant and economically feasible. This challenge is met by the Radkowsky Thorium Reactor (RTR) concept. So far the concept has been applied to a Russian design of a 1,000 MWe pressurized water reactor, known as a WWER-1000, and designated as VVERT. The following are the main results of the preliminary reference design: * The amount of Pu contained in the RTR spent fuel stockpile is reduced by 80% in comparison with a VVER of a current design. * The isotopic composition of the RTR-Pu greatly increases the probability of pre-initiation and yield degradation of a nuclear explosion. An extremely large Pu-238 content causes correspondingly large heat emission, which would complicate the design of an explosive device based on RTR-Pu. The economic incentive to reprocess and reuse the fissile component of the RTR spent fuel is decreased. The once-through cycle is economically optimal for the RTR core and cycle. To summarize all the items above: the replacement of a standard (U-based) fuel for nuclear reactors of current generation by the RTR fuel will provide an inherent barrier for nuclear weapon proliferation. This inherent barrier, in combination with existing safeguard measures and procedures is adequate to unambiguously disassociate civilian nuclear power from military nuclear power. * The RTR concept is applied to existing power plants to assure its economic feasibility. Reductions in waste disposal requirements, as well as in natural U and fabrication expenses, as compared to a standard WWER fuel, provide approximately 20% reduction in fuel cycle (authors).

  15. Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

    Science.gov (United States)

    LeVert, Francis E.; Cox, Samson A.

    1978-01-01

    A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor.

  16. Threshold self-powered gamma detector for use as a monitor of power in a nuclear reactor

    International Nuclear Information System (INIS)

    LeVert, F.E.; Cox, S.A.

    1978-01-01

    A self-powered gamma monitor for placement near the core of a nuclear reactor comprises a lead prism surrounded by a coaxial thin nickel sheet, the combination forming a collector. A coaxial polyethylene electron barrier encloses the collector and is separated from the nickel sheet by a vacuum region. The electron barrier is enclosed by a coaxial stainless steel emitter which, in turn, is enclosed within a lead casing. When the detector is placed in a flux of gamma rays, a measure of the current flow in an external circuit between emitter and collector provides a measure of the power level of the reactor

  17. State-space model predictive control method for core power control in pressurized water reactor nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Guo Xu; Wu, Jie; Zeng, Bifan; Wu, Wangqiang; Ma, Xiao Qian [School of Electric Power, South China University of Technology, Guangzhou (China); Xu, Zhibin [Electric Power Research Institute of Guangdong Power Grid Corporation, Guangzhou (China)

    2017-02-15

    A well-performed core power control to track load changes is crucial in pressurized water reactor (PWR) nuclear power stations. It is challenging to keep the core power stable at the desired value within acceptable error bands for the safety demands of the PWR due to the sensitivity of nuclear reactors. In this paper, a state-space model predictive control (MPC) method was applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, the MPC model, and quadratic programming (QP). The mathematical models of the reactor core were based on neutron dynamic models, thermal hydraulic models, and reactivity models. The MPC model was presented in state-space model form, and QP was introduced for optimization solution under system constraints. Simulations of the proposed state-space MPC control system in PWR were designed for control performance analysis, and the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  18. Human Factors Engineering Review Model for advanced nuclear power reactors

    International Nuclear Information System (INIS)

    O'Hara, J.; Higgins, J.; Goodman, C.; Galletti, G.: Eckenrode, R.

    1993-01-01

    One of the major issues to emerge from the initial design reviews under the certification process was that detailed human-systems interface (HSI) design information was not available for staff review. To address the lack of design detail issue. The Nuclear Regulatory Commission (NRC) is performing the design certification reviews based on a design process plan which describes the human factors engineering (HFE) program elements that are necessary and sufficient to develop an acceptable detailed design specification. Since the review of a design process is unprecedented in the nuclear industry. The criteria for review are not addressed by current regulations or guidance documents and. therefore, had to be developed. Thus, an HFE Program Review Model was developed. This paper will describe the model's rationale, scope, objectives, development, general characteristics. and application

  19. Removal of toxic uranium from synthetic nuclear power reactor effluents using uranyl ion imprinted polymer particles.

    Science.gov (United States)

    Preetha, Chandrika Ravindran; Gladis, Joseph Mary; Rao, Talasila Prasada; Venkateswaran, Gopala

    2006-05-01

    Major quantities of uranium find use as nuclear fuel in nuclear power reactors. In view of the extreme toxicity of uranium and consequent stringent limits fixed by WHO and various national governments, it is essential to remove uranium from nuclear power reactor effluents before discharge into environment. Ion imprinted polymer (IIP) materials have traditionally been used for the recovery of uranium from dilute aqueous solutions prior to detection or from seawater. We now describe the use of IIP materials for selective removal of uranium from a typical synthetic nuclear power reactor effluent. The IIP materials were prepared for uranyl ion (imprint ion) by forming binary salicylaldoxime (SALO) or 4-vinylpyridine (VP) or ternary SALO-VP complexes in 2-methoxyethanol (porogen) and copolymerizing in the presence of styrene (monomer), divinylbenzene (cross-linking monomer), and 2,2'-azobisisobutyronitrile (initiator). The resulting materials were then ground and sieved to obtain unleached polymer particles. Leached IIP particles were obtained by leaching the imprint ions with 6.0 M HCl. Control polymer particles were also prepared analogously without the imprint ion. The IIP particles obtained with ternary complex alone gave quantitative removal of uranyl ion in the pH range 3.5-5.0 with as low as 0.08 g. The retention capacity of uranyl IIP particles was found to be 98.50 mg/g of polymer. The present study successfully demonstrates the feasibility of removing uranyl ions selectively in the range 5 microg - 300 mg present in 500 mL of synthetic nuclear power reactor effluent containing a host of other inorganic species.

  20. Gaseous-fuel nuclear reactor research for multimegawatt power in space

    Science.gov (United States)

    Thom, K.; Schneider, R. T.; Helmick, H. H.

    1977-01-01

    In the gaseous-fuel reactor concept, the fissile material is contained in a moderator-reflector cavity and exists in the form of a flowing gas or plasma separated from the cavity walls by means of fluid mechanical forces. Temperatures in excess of structural limitations are possible for low-specific-mass power and high-specific-impulse propulsion in space. Experiments have been conducted with a canister filled with enriched UF6 inserted into a beryllium-reflected cavity. A theoretically predicted critical mass of 6 kg was measured. The UF6 was also circulated through this cavity, demonstrating stable reactor operation with the fuel in motion. Because the flowing gaseous fuel can be continuously processed, the radioactive waste in this type of reactor can be kept small. Another potential of fissioning gases is the possibility of converting the kinetic energy of fission fragments directly into coherent electromagnetic radiation, the nuclear pumping of lasers. Numerous nuclear laser experiments indicate the possibility of transmitting power in space directly from fission energy. The estimated specific mass of a multimegawatt gaseous-fuel reactor power system is from 1 to 5 kg/kW while the companion laser-power receiver station would be much lower in specific mass.

  1. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  2. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    EL-Kafas, A.E.A.E.

    1996-01-01

    the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

  3. Safety problems of nuclear power plants with channel-type graphite boiling water reactors

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Vasilevskij, V.P.; Volkov, V.P.; Gavrilov, P.A.; Kramerov, A.Ya.; Kuznetsov, S.P.; Kunegin, E.P.; Rybakov, N.Z.

    1977-01-01

    Construction of nuclear power plants in a highly populated region near large industrial centres necessitates to pay a special attention to their nuclear and radiation safety. Safety problems of nuclear reactor operation are discussed, in particular, they are: reliable stoppage of fission chain reaction at any emergency cases; reliable core cooling with failure of various equipment; emergency core cooling with breached pipes of a circulating circuit; and prevention of radioactive coolant release outside the nuclear power plant in amount exceeding the values adopted. Channel-type water boiling reactors incorporate specific features requiring a new approach to safety operation of a reactor and a nuclear power plant. These include primarily a rather large steam volume in the coolant circuit, large amount of accumulated heat, void reactivity coefficient. Channel-type reactors characterized by fair neutron balance and flexible fuel cycle, have a series of advantages alleviating the problem of ensuring their safety. The possibility of reliable control over the state of each channel allows to replace failed fuel elements by the new ones, when operating on-load, to increase the number of circulating loops and reduce the diameter of main pipelines, simplifies significantly the problem of channel emergency cooling and localization of a radioactive coolant release from a breached circuit. The concept of channel-type reactors is based on the solution of three main problems. First, plant safety should be assured in emergency switch off of separate units and, if possible, energy conditions should be maintained, this is of particular importance considering the increase in unit power. Second, the system of safety and emergency cooling should eliminate a great many failures of fuel elements in case of potential breaches of any tube in the circulating circuit. Finally, rugged boxes and localizing devices should be provided to exclude damage of structural elements of the nuclear power

  4. Use mobile pumps and liquid chilling water units to provide chilled water for nuclear reactor during nuclear power plant accident

    International Nuclear Information System (INIS)

    Zhang Guobin; Feng Jiaxuan

    2012-01-01

    From the nuclear accident in Japan Fuksuhima in March this year, despite a shut down of the reactor, the residue heat inside the reactor was not able to remove due to the failure of the cooling system and it finally caused the catastrophe. It was observed that when the failure of the cooling system after an earthquake of magnitude 9 and a tsunami of 28 meters height, the containment vessel for the reactor core was still able to maintain its integrity in the first 24 hours before the first explosion was happened. A backup emergency heat removal system for nuclear power plants using mo- bile pumps and liquid chilling units has been proposed 20 years ago by Cheung [Ref. 1]. Due to the fact that there are more than 400 nuclear power plants around the world and 10% of them are located in earthquake active zone, together with the aging of some of the power plants which were built more than 30 years ago, the risk of another nuclear accident becomes high. An emergency safety measure has to be designed in order to deal with the unforeseen scenario. This re- port re-visits the proposal again; to re-design to the suit the need and to integrate with the current situation of the nuclear industry. (authors)

  5. Estimates of the financial consequences of nuclear-power-reactor accidents

    International Nuclear Information System (INIS)

    Strip, D.R.

    1982-09-01

    This report develops preliminary techniques for estimating the financial consequences of potential nuclear power reactor accidents. Offsite cost estimates are based on CRAC2 calculations. Costs are assigned to health effects as well as property damage. Onsite costs are estimated for worker health effects, replacement power, and cleanup costs. Several classes of costs are not included, such as indirect costs, socio-economic costs, and health care costs. Present value discounting is explained and then used to calculate the life cycle cost of the risks of potential reactor accidents. Results of the financial consequence estimates for 156 reactor-site combinations are summarized, and detailed estimates are provided in an appendix. The results indicate that, in general, onsite costs dominate the consequences of potential accidents

  6. Outline of construction planning on No. 2 Reactor of the Shika Nuclear Power Plant

    International Nuclear Information System (INIS)

    Nakagawa, Tetsuro; Kadoki, Shuichi; Kubo, Tetsuji

    1999-01-01

    The Hokuriku Electric Co., Ltd. carries out the expansion of the Shika Nuclear Power Plant No.2 (ABWR) to start its in March 2006. It is situated in north neighboring side of No. 1 reactor under operation at present, and its main buildings are planned to position a reactor building at mountain side and a turbine building at sea side as well as those in the No. 1 reactor. And, cooling water for steam condenser was taken in from an intake opening built at north side of the lifting space situated at the front of the power plant, and discharged into seawater from a flashing opening positioned about 600 m offing. Here were described on outline of main civil engineering such as base excavation engineering, concrete caisson production, oceanic establishment engineering, and facility for steam condenser, and characteristics of the engineering. (G.K.)

  7. Occupational radiation exposure at commercial nuclear power reactors, 1981. Annual report

    International Nuclear Information System (INIS)

    Brooks, B.G.

    1982-11-01

    This report presents an updated compilation of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1981. This year's report contains data received from the 70 light water cooled reactors (LWRs) and one high temperature gas cooled reactor that had been declared to be in commercial operation for at least one full year as of December 31, 1981. This represents an increase of two reactors over the number contained in last year's report. The total number of personnel monitored at LWRs in 1981 was 124,504, a slight decrease from that found in 1980. The number of workers that received measurable doses during 1981 was 82,183 which is about 2000 more than that found in 1980. The total collective dose at LWRs for 1981 is estimated to be 54,142 man-rems, which is only about 350 man-rems more than that reported in 1980. The report also presents a summary and some analyses of the exposure data contained in the termination reports that have been submitted by nuclear power licensees to the Commission pursuant to 10 CFR Section 20.408. As of December 31, 1981, personal identification and exposure information had been collected and computerized for some 210,000 of these terminating reactor personnel

  8. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  9. Source driven breeding fission power reactors and the nuclear energy strategy

    International Nuclear Information System (INIS)

    Greenspan, E.

    The nuclear energy economy is facing severe difficulties associated with low utilization of uranium resources, safety, non-proliferation and environmental issues. Energy policy makers face the dilemma: commercialize LMFBRs immediately with the risk of negative economical, proliferation or other consequences, or continue with R and D programs that will provide the information needed for sounder decisions, but now taking the risk of running out of economically exploitable uranium ore resources. The development of hybrid reactors can provide an assurance against the latter risk and offers many interesting new options for the nuclear energy strategy. Being based on the technology of LWRs and HWRs, Light Water Hybrid Reactors (LWHR) provide a most natural link between the fission reactor technology of the present and the fusion power technology of the future. The investment in their development in excess of that required for the development of fusion power reactors is expected to be relatively small, thus making the development of LWHRs potentially a high benefit-to-cost ratio program. It is recommended that the fission and fusion communities will cooperate in hybrids R and D programs aimed at assessing the technological and economical viability of hybrid reactors as reliably and soon as possible. (author)

  10. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1980-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

  11. Expert systems for analysis and management assistance of nuclear reactor electrical power supplies

    International Nuclear Information System (INIS)

    Evrard, J.M.; Souchet, Y.

    1988-01-01

    Electrical power supplies of nuclear plants are very complex systems. They are studied in many ways: failure consequences, probabilistic risk assessement, failure diagnosis, corrective actions in case of incident ... Knowledge base technology (expert systems) is very suited to solve these problems. A common structural representation can generate specific functional representations; thus we get coherency and easy evolution. This paper shows a two facet methodology: the plant description (both static and dynamic) and reasoning about it. Current applications are developed for pressurized water reactors and gas cooled reactors, using the SPIRAL expert system shell buil at CEA [fr

  12. Application in nuclear engineering: methodology of innovative nuclear reactors: approaches to the safety of future nuclear power plants

    International Nuclear Information System (INIS)

    Alramady, A.M.K

    2008-01-01

    This thesis describes RELAP5 and MATLAB/SIMULINK computer codes for thermal hydraulic analysis of a typical pressurized water reactor (PWR). The two codes are used to calculate the thermal-hydraulic characteristics of the reactor core and the primary loop under steady-state and hypothetical accidents conditions.New designs of nuclear power plants are directed to increase safety by many methods like reducing the dependence on active parts (such as safety pumps, fans, and diesel generators ) and replacing them with passive features such as gravity draining of cooling water from tanks, and natural circulation of water and air. In this work, high and medium pressure injection pumps are replaced by passive injection components. Different break sizes in cold leg pipe are simulated to analyze to what degree the plant is safe (without any operator action) by using only these passive components. The passive design means operators would not need to take immediate action after an accident, with the reactor ,instead, safely shutting down on its own. Different accident scenarios were simulated in this thesis as loss of coolant accidents and station blackout accidents, and complete passive safety systems used to mitigate theses accidents.

  13. Facing the facts: the suspension of a design requirement applicable to nuclear power reactor effluents

    International Nuclear Information System (INIS)

    Amado, Valeria; Biaggio, Alfredo; Canoba, Analia; Curti, Adriana

    2008-01-01

    A design requirement aimed at limiting the discharge of globally dispersed long-lived radionuclides released by nuclear power reactors was in force in Argentine since 1979 till recently. The practical results of such regulatory measure was the need to retain C-14 in the PHWR under construction, as well as in furniture heavy water reactors to be built in the country. This paper explains the basis of such requirement, which was formulated as a collective dose constraint per unit of energy generated, and the main assumptions that triggered it. The differences between the assumptions made at the time and the reality of nuclear power generation at the beginning of the 21 st century, as well as their implications in relation to the requirement are described, including the Suess effect and its impact in the total dose due to C-14. Finally, the facts that made no longer reasonable to keep in force the above mentioned requirement are presented. (author)

  14. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    Science.gov (United States)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  16. Applications of nuclear reactor power systems to electric propulsion missions.

    Science.gov (United States)

    Schaupp, R. W.; Sawyer, C. D.

    1971-01-01

    The performance of nuclear electric propulsion systems (NEP) has been evaluated for a wide variety of missions in an attempt to establish the commonality of NEP system requirements. Emphasis was given to those requirements and system characteristics that serve as guidelines for current technology development programs. Various interactions and tradeoffs between NEP system and mission parameters are described. The results show that the most significant factors in selecting NEP system size are launch mode (direct or spiral escape) and, to a weaker extent, launch vehicle capability. Other factors such as mission, payload, and thrust time constraints, have little influence, thus allowing one NEP system to be used for many missions. The results indicated that a 100 kWe NEP would be suitable for most direct escape missions and a 250 kWe NEP system would be suitable for more demanding missions that use the spiral escape mode.

  17. Implementation of nuclear power plant simulation in start-up commissioning of reactor control system

    International Nuclear Information System (INIS)

    Yang Zongwei; Huang Tieming; Feng Guangyu; Luan Zhenhua; Lin Meng; Zhu Lizhi

    2009-01-01

    Based on the nuclear power thermal-hydraulic model, Labview graphical programming language and virtual instrument data acquisition technology, this paper describes a dedicate test platform to solve the problem that the reactor control system (RRC) can not be evaluated and analyzed far before the actual startup of the unit. By connecting the test platform to the nuclear Digital Control System (DCS), the step-by-step closed-looped test and global function test of RRC system were performed, the dynamic validation and logical function demonstration for RRC were realized, and a lot of configuration mistakes of RRC and nonconformity were solved. The test for unit 3 of Ling'ao phase II has proved that the implementation of nuclear power plant simulation in the start-up commissioning of RRC can greatly reduce the risk of normal power operation and great transient tests, with which the term of startup for overall unit test can be greatly shortened. (authors)

  18. Background information document to support NESHAPS rulemaking on nuclear power reactors. Draft report

    International Nuclear Information System (INIS)

    Colli, A.; Conklin, C.; Hoffmeyer, D.

    1991-08-01

    The purpose of this Background Information Document (BID) is to present information relevant to the Administrator of the Environmental Protection Agency's (EPA) reconsideration of the need for a NESHAP to control radionuclides emitted to the air from commercial nuclear power reactors. The BID presents information on the relevant portions of the regulatory framework that NRC has implemented for nuclear power plant licensees, under the authority of the Atomic Energy Act, as amended, to protect the public's health and safety. To provide context, it summarizes the rulemaking history for Subpart I. It then describes NRC's regulatory program for routine atmospheric emissions of radionuclides and evaluates the doses caused by actual airborne emissions from nuclear power plants, including releases resulting from anticipated operational occurrences

  19. Power thresholds for fast oscillatory instabilities in nuclear reactors: a simple mathematical model

    International Nuclear Information System (INIS)

    Suarez-Antola, Roberto; Uruguay)

    2007-01-01

    The cores of nuclear reactors, including its structural parts and cooling fluids, are complex mechanical systems able to vibrate in a set of normal modes and frequencies, if suitable perturbed. The cyclic variations in the strain state of the core materials may modify the reactivity, and thus thermal power, producing variations in strain due to thermal-elastic effects. If the variation of the temperature field is fast enough and if the Doppler Effect and other stabilizing prompt effects in the fuel are weak enough, a fast oscillatory instability could be produced, coupled with mechanical vibrations of small enough amplitude that they will not be excluded by the procedures of conventional mechanical design. After a careful discussion of the time scales of neutron kinetics, thermal-elastic and vibration phenomena, a simple lumped parameter mathematical model is constructed in order to study, in a first approximation, the stability of the reactor. An integro-differential equation for power kinetics is derived. Under certain conditions, fast oscillatory instabilities are found when power is greater than a threshold value, and the delay in the global power feedback loop is big enough. Approximate analytical formulae are given for the power threshold, critical delay and power oscillation frequency. It is shown that if prompt stabilizing fuel effects are strong enough, dangerous fast power oscillations due to mechanical thermal-nuclear coupling phenomena can not appear at any power level. (author)

  20. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

  1. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  2. Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

    International Nuclear Information System (INIS)

    Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

    1979-01-01

    The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

  3. Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

    International Nuclear Information System (INIS)

    Devgun, Jas S.

    2013-01-01

    Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

  4. Nuclear power status 1999

    International Nuclear Information System (INIS)

    2000-01-01

    The document gives statistical information on nuclear power plants status in the world in 1999, including the number of reactors in operation or under construction, the electricity supplied by nuclear power reactors and the respective percentage of electricity produced by nuclear energy in 1999, and the total operating experience to 31 December 1999, by country

  5. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  6. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  7. Development of a tool for comparing different nuclear power reactor technologies: the Mexican choice

    International Nuclear Information System (INIS)

    Martin-del-Campo, C.; Francois, J.L.; Reyes, R.

    2007-01-01

    This paper describes a methodology which has allowed us to make a comparative assessment of nuclear power reactor options. The methodology was divided in 3 steps. The first step consists in searching of common indicators to be compared. A total of twenty indicators were considered and grouped in 3 main criteria. The second step is to obtain the values of all the indicators for each of the reactor technologies being compared. The third step is to utilize an aggregation method to integrate all the indicators in an overall qualification. Fuzzy Logic was selected as multi criteria aggregation method because it copes with imprecisely defined data; it can model non-linear functions of arbitrary complexity; and it is able to build on top of the experience of experts. The Fuzzy Logic inference system was built using the MATLAB toolbox; 3 fuzzy sets were described for each entry variable (Indicator) and 5 fuzzy sets for the output variable (Qualification). Both, the set of membership function and the set of rules were defined in combination. The methodology is simple but at the same time is powerful; it allows the use of all the indicators with their own magnitudes and units. Five reactors were compared: the Advanced Boiling Water Reactor (ABWR), the Economic Simplified Boiling Water Reactor (ESBWR), the Evolutionary Pressurized water Reactor (EPR), the Advanced Pressurized water reactor 1000 (AP1000) and the Pebbled Bed Modular Reactor (PBMR). Preliminary results were obtained using non official data obtained from public information. The qualifications of the reactors appear to be quite near. This work should be improved by taking into account which indicator is important and grading the indicators according to the situation in Mexico. (authors)

  8. Solution of the Lambda modes problem of a nuclear power reactor using an h–p finite element method

    International Nuclear Information System (INIS)

    Vidal-Ferrandiz, A.; Fayez, R.; Ginestar, D.; Verdú, G.

    2014-01-01

    Highlights: • An hp finite element method is proposed for the Lambda modes problem of a nuclear reactor. • Different strategies can be implemented for increasing the accuracy of the solutions. • 2D and 3D benchmarks have been studied obtaining accurate results. - Abstract: Lambda modes of a nuclear power reactor have interest in reactor physics since they have been used to develop modal methods and to study BWR reactor instabilities. An h–p-Adaptation finite element method has been implemented to compute the dominant modes the fundamental mode and the next subcritical modes of a nuclear reactor. The performance of this method has been studied in three benchmark problems, a homogeneous 2D reactor, the 2D BIBLIS reactor and the 3D IAEA reactor

  9. Pressurized water reactor nuclear power plant. Environmental characterization information report

    International Nuclear Information System (INIS)

    1981-01-01

    The typical plant chosen for characterization is a 10000-MWe nameplate rating with wet-natural-draft cooling towers and modern radwaste control and processing equipment. The process, plant operating parameters, resources needed, and the environmental residuals and products associated with the power plant are presented. Annual resource usage and pollutant discharges are shown in English and metric units, assuming an annual plant capacity factor of 70%. In addition to annual quantities, the summary table gives quantities in terms of 10 12 Btu (about 293 million kWh) of electrical energy produced for comparison among energy processes. Supporting information and calculation procedures for the data are given. Thirteen environmental points of interest are discussed individually. Cost information, typical radioactive releases, and use of cooling ponds as an alternative cooling method are discussed in appendixes. A glossary and list of acronyms and abbreviations are provided

  10. Methods and means of the radioisotope flaw detection of the nuclear power reactors components

    International Nuclear Information System (INIS)

    Dekopov, A.S.; Majorov, A.N.; Firsov, V.G.

    1979-01-01

    Methods and means are considered for the radioisotopic flaw detection of the nuclear reactors pressure vessels and structural components of the reactor circuit. Methods of control are described as in the technological process of fabrication of the power reactors assemblies as during the systematic-preventive repair of the nuclear power station equipment during exploitation. Methodological base is given of the technology of radiation control of welded joints of the pressure vessel branch piper of the WWER-440 and WWER-1000 reactors in the process of assembling and exploitation and joining pipes with the pipe-plate of the steamgenerator in the process of fabrication. Methods of the radioisotope flaw detection in the process of exploitation take into consideration the influence of the radioisotope background, and ensure obtaining of the demanded by the rules of control, sensitivity. Methods of control of welded joints of the steamgenerator of nuclear power plants are based on the simultaneous examination of all joints with application of the shaped radiographic plate-holders. Special gamma-flaw-detection equipment is developed for control of the welded joints of the main branch-pipes. Design peculiarities are given of the installation for flaw detection. These installations are equipped with the system for emergency return of the radiation source into the storage position from the position for exposure. They have automatic exposure-meters for determination of the exposure time. Successfull exploitation of such installations in the Finland during assembling equipment for the nuclear reactor of the nuclear power plant ''Loviisa-1'' and in the USSR on the Novovoronezh nuclear power plant has shown possibility for detection of flaws having dimensions about 1% of the equipment used. For control of welded joints of pipes with pipe-plates at the steam generators, portable flaw-detectors are used. Sensitivity of these flaw-detectors towards detection of the wire standards has

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    Schweiger, F.; Glahe, E.

    1976-01-01

    In a nuclear reactor of the kind which is charged with spherical reaction elements and in which control rods are arranged to be thrust directly into the charge, each control rod has at least one screw thread on its external surface so that as the rod is thrust into the charge it is caused to rotate and thus make penetration easier. The length of each control rod may have two distinct portions, a latter portion which carries a screw thread and a lead-in portion which is shorter than the latter portion and which may carry a thread of greater pitch than that on the latter portion or may have a number of axially extending ribs instead of a thread

  12. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  13. 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.)

  14. Design and optimization of fuzzy-PID controller for the nuclear reactor power control

    International Nuclear Information System (INIS)

    Liu Cheng; Peng Jinfeng; Zhao Fuyu; Li Chong

    2009-01-01

    This paper introduces a fuzzy proportional-integral-derivative (fuzzy-PID) control strategy, and applies it to the nuclear reactor power control system. At the fuzzy-PID control strategy, the fuzzy logic controller (FLC) is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region and the genetic algorithm to improve the 'extending' precision through quadratic optimization for the membership function (MF) of the FLC. Thus the FLC tunes the gains of PID controller to adapt the model changing with the power. The fuzzy-PID has been designed and simulated to control the reactor power. The simulation results show the favorable performance of the fuzzy-PID controller.

  15. Transfer parameters of fission and activation products present in effluents of nuclear power reactors

    International Nuclear Information System (INIS)

    Cancio, D.; Menossi, C.A.; Ciallella, N.R.

    1978-01-01

    The paper presents results of research carried out in Argentina on transfer parameters of fission and activation products which may be present in the effluents of nuclear power reactors. For some nuclides, as Sr-90, Co-137 and I-131, the parameters were obtained by studies of the fallout, from measurements of integrated levels in the environment and in the food chains. Other values are concentration factors derived from laboratory and field experiments. They refer to fish, molluscs, crustaces and fresh water plants, for several fission and activation nuclides. Transfer parameters obtained have been of significant importance for environmental assessments, relating to nuclear installations in Argentina. (author)

  16. Electric failure on the reactor n.3 of the nuclear power plant of Dampierre

    International Nuclear Information System (INIS)

    2007-05-01

    This note of information resumes the progress of the electric failure on the reactor n.3 of the nuclear power plant of Dampierre, the organization during the incident, it establishes then a comparison with the incident arisen to Forsmark in 2006 and reminds that it lead in an inspection on behalf of the Asn which noticed that all the procedures had been respected by the operators and did not noticed any abnormality in the maintenance. This event was classified at the level 1 of the international nuclear event scale (INES). (N.C.)

  17. Installation modification of the reactor No.2 of Ikata nuclear power plant of Shikoku Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1980-01-01

    The application was made on August 25, 1979, from the president of the Shikoku Electric Power Company, Inc., to the Minister of International Trade and Industry, relating to the installation modification of the reactor No. 2 in the Ikata nuclear power plant. The inquiry was submitted on September 28, 1979, from the Minister of International Trade and Industry to the Nuclear Safety Commission, after the safety evaluation in the Ministry of International Trade and Industry, and the investigation and deliberation were started on October 1, 1979, in the Nuclear Safety Commission. The content of the modification is to add the circuit actuated by the abnormal low pressure signal of the reactor to the actuating circuit of the emergency core cooling system (ECCS) and to increase the new fuel storage capacity from about 1/3 core to about 2/3 core. The additional signal circuit is composed of the logic circuit of ''2 out of 4'' and is multichannel design. The circuit is independent from the reactor control system and the conventional signal circuit of the concurrence of low pressure in the reactor and low level in the pressurizer. With the addition of the circuit of abnormal low pressure signal of the reactor, the countermeasures for preventing ECCS start by mistake are also added. These modifications give no influence to the functions of the reactor control system and reactor protection system. The function and the performance of ECCS were analyzed and evaluated accompanying these modifications assuming the loss of coolant accident. Concerning the new fuel storage capacity, the type of racks is modified from angle type to can type, and the subcriticality is kept even at the time of water flood. (Nakai, Y.)

  18. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  19. Nuclear research reactors in Brazil

    International Nuclear Information System (INIS)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias

    2011-01-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  20. Installation and commissioning of operation nuclear power plant reactor protection system modernization project

    International Nuclear Information System (INIS)

    Lu Weiwei

    2010-01-01

    Qinshan Nuclear Power Plant is the first nuclear power plant in mainland China; it is also the first one which realizes the modernization of analog technology based Reactor Protection System in the operation nuclear power plant of China. The implementation schedule is the shortest one which use same digital technology platform (TELEPERM XS of AREVA NP) to modifying the safety class I and C system in the world, the whole project spent 28 months from equipment contract signed to putting system into operation. It open up a era for operation nuclear power plant using mature digital technology to make safety class I and C system modernization in China. The important practical significance of this successful project is very obvious. This article focus on two important project stage--equipment installation and system commissioning, it is based on a large number of engineering implementation fact, it covers the problems and solutions happened during the installation and commission. The purpose of the article is to share the experience and lessons of safety I and C system modernization for other operation nuclear power plant. (authors)

  1. Radiation protection for repairs of reactor's internals at the 2nd Unit of the Nuclear Power Plant Temelin

    International Nuclear Information System (INIS)

    Zapletal, P.; Konop, R.; Koc, J.; Kvasnicka, O.; Hort, M.

    2011-01-01

    This presentation describes the process and extent of repairs of the 2 nd unit of the Nuclear power plant Temelin during the shutdown of the reactor. All works were optimized in terms of radiation protection of workers.

  2. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Pearson, K.G.

    1977-01-01

    Reference is made to auxiliary means of cooling the nuclear fuel clusters used in light or heavy water cooled nuclear reactors. One method is to provide one or more spray cooling tubes. From holes in the side walls of those tubes coolant water may be sprayed laterally into the cluster against the rods. The flow of main coolant may thus be supplemented or even replaced by the auxiliary coolant. A difficulty, however, is that only those fuel rods close to a spray cooling tube can readily be reached by the auxiliary coolant. In the arrangement described, where the fuel rods are spaced apart by transverse grids, at least one of the interspaces between the grids is provided with an axially extending auxiliary coolant conduit having lateral holes through which an auxiliary coolant is sprayed into the cluster. A deflector is provided that extends from a transverse grid into a position in front of the holes and deflects auxiliary coolant on to parts of the fuel rods otherwise inaccessible to the auxiliary coolant. The construction of the deflector is described. (U.K.)

  4. Rock siting of nuclear power plants from a reactor safety standpoint

    International Nuclear Information System (INIS)

    1975-11-01

    The study has aimed at surveying the advantages and disadvantages of a rock sited nuclear power plant from a reactor safety standpoint. The studies performed are almost entirely concentrated on the BWR alternative. The design of a nuclear power plant in rock judged most appropriate has been studied in greater detail, and a relatively extensive safety analysis has been made. It is found that the presented technical design of the rock sited alternative is sufficiently advanced to form a basis for further projecting treatment. The chosen technical design of the reactor plant demands a cavern with a 45-50 metre span. Caverns without strengthening efforts with such spans are used in mines, but have no previously been used for industrial plants. Studies of the stability of such caverns show that a safety level is attainable corresponding to the safety required for the other parts of the nuclear power plant. The conditions are that the rock is of high quality, that necessary strengthening measures are taken and that careful studies of the rock are made before and during the blasting, and also during operation of the plant. When locating a rock sited nuclear power plant, the same criteria must be considered as for an above ground plant, with additional stronger demands for rock quality. The presented rock sited nuclear power plant has been assessed to cost 20 % more in total construction costs than a corresponding above ground plant. The motivations for rock siting also depend on whether a condensing plant for only electricity production, or a plant for combined power production and district heating, is considered. The latter would under certain circumstances make rock siting look more attractive. (author)

  5. Influence of regulatory requirements for nuclear power plants on the backfitting of Austrian research reactors

    International Nuclear Information System (INIS)

    Boeck, H.; Hammer, J.

    1985-01-01

    In general the licensing and backfitting activities have once more demonstrated the fact that safety assessment of a research reactor is by no means just a scaled-down version of a nuclear power plant licensing procedure. Naturally the risk potential is much lower, however, the very nature of research calls for much more flexibility in operation, for temporary installations and for experimental methods which cannot be covered by detailed regulations in advance. Therefore the application of nuclear power reactor criteria to such facilities has to be considered with extreme caution. If NPP standards are applicable at all, they have to be carefully interpreted in each individual case. It is interesting to compare the original reactor safety reports with their modern versions: emphasis has shifted from reactivity accident calculations to thermal-hydraulic considerations, to better instrumentation (both in quality and quantity) and to more effort in reducing, measuring and documenting all radioactive effluents. This tendency is also reflected in most of the backfitting requirements. In summary, the result of the lengthy licensing and backfitting process is certainly a considerable improvement in performance and safety of the Austrian research reactors

  6. Nuclear reactor

    International Nuclear Information System (INIS)

    Schabert, H.P.; Weber, R.; Bauer, A.

    1975-01-01

    The refuelling of a PWR power reactor of about 1,200 MWe is performed by a transport pipe in the containment leading from an external to an internal fuel pit. A wagon to transport the fuel elements can go from a vertical loading position to an also vertical deloading position in the inner fuel pit via guide rollers. The necessary horizontal movement is effected by means of a cable line through the transport pipe which is inclined at least 10 0 . Gravity thus helps in the movement to the deloading position. The cable line with winch is fastened outside the containment. Swivelling devices tip the wagon from the horizontal to the vertical position or vice versa. Loading and deloading are done laterally. (TK/LH) [de

  7. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Executive summary: main report. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks.

  8. Nuclear reactor power for a space-based radar. SP-100 project

    Science.gov (United States)

    Bloomfield, Harvey; Heller, Jack; Jaffe, Leonard; Beatty, Richard; Bhandari, Pradeep; Chow, Edwin; Deininger, William; Ewell, Richard; Fujita, Toshio; Grossman, Merlin

    1986-01-01

    A space-based radar mission and spacecraft, using a 300 kWe nuclear reactor power system, has been examined, with emphasis on aspects affecting the power system. The radar antenna is a horizontal planar array, 32 X 64 m. The orbit is at 61 deg, 1088 km. The mass of the antenna with support structure is 42,000 kg; of the nuclear reactor power system, 8,300 kg; of the whole spacecraft about 51,000 kg, necessitating multiple launches and orbital assembly. The assembly orbit is at 57 deg, 400 km, high enough to provide the orbital lifetime needed for orbital assembly. The selected scenario uses six Shuttle launches to bring the spacecraft and a Centaur G upper-stage vehicle to assembly orbit. After assembly, the Centaur places the spacecraft in operational orbit, where it is deployed on radio command, the power system started, and the spacecraft becomes operational. Electric propulsion is an alternative and allows deployment in assembly orbit, but introduces a question of nuclear safety.

  9. Nuclear power desalinating complex with IRIS reactor plant and Russian distillation desalinating unit

    International Nuclear Information System (INIS)

    Kostin, V. I.; Panov, Yu.K.; Polunichev, V. I.; Fateev, S. A.; Gureeva, L. V.

    2004-01-01

    This paper has been prepared as a result of Russian activities on the development of nuclear power desalinating complex (NPDC) with the IRIS reactor plant (RP). The purpose of the activities was to develop the conceptual design of power desalinating complex (PDC) and to evaluate technical and economical indices, commercial attractiveness and economical efficiency of PDC based on an IRIS RP with distillation desalinating plants. The paper presents the main results of studies as applied to dual-purpose PDC based on IRIS RP with different types of desalinating plants, namely: characteristics of nuclear power desalinating complex based on IRIS reactor plant using Russian distillation desalinating technologies; prospective options of interface circuits of the IRIS RP with desalinating plants; evaluations of NPDC with IRIS RP output based on selected desalinating technologies for water and electric power supplied to the grid; cost of water generated by NPDC for selected interface circuits made by the IAEA DEEP code as well as by the Russian TEO-INVEST code; cost evaluation results for desalinated water of PDC operating on fossil fuel and conditions for competitiveness of the nuclear PDC based on IRIS RP compared with analog desalinating complexes operating on fossil fuel.(author)

  10. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

    International Nuclear Information System (INIS)

    Tatone, O.S.; Meindl, P.; Taylor, G.F.

    1986-06-01

    A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

  11. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    Energy Technology Data Exchange (ETDEWEB)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens, E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas

    2015-07-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  12. Decommissioning strategy and schedule for a multiple reactor nuclear power plant site

    International Nuclear Information System (INIS)

    Monteiro, Deiglys Borges; Moreira, Joao M.L.; Maiorino, Jose Rubens

    2015-01-01

    The decommissioning is an important part of every Nuclear Power Plant life cycle gaining importance when there are more than one plant at the same site due to interactions that can arise from the operational ones and a decommissioning plant. In order to prevent undesirable problems, a suitable strategy and a very rigorous schedule should implemented and carried. In this way, decommissioning tasks such as fully decontamination and dismantling of activated and contaminated systems, rooms and structures could be delayed, posing as an interesting option to multiple reactor sites. The present work aims to purpose a strategy and a schedule for the decommissioning of a multiple reactor site highlighting the benefits of delay operational tasks and constructs some auxiliary services in the site during the stand by period of the shutdown plants. As a case study, will be presented a three-reactor site which the decommissioning process actually is in planning stage and that should start in the next decade. (author)

  13. The inherently-safe power reactor DYONISOS (Dynamic Nuclear Inherently-Safe Reactor Operating with Spheres)

    International Nuclear Information System (INIS)

    Taube, M.; Lanfranchi, M.; Weissenfluh, Th. von; Ligou, J.; Yadigaroglu, G.; Taube, P.

    1986-01-01

    A philosophy of inherent safety is formulated and an inherently-safe thermal power reactor is presented. Solid fuel in the form of spheres a few centimetres in diameter is suspended under the hydro-dynamic pressure of molten lead coolant in vertical channels within the graphite moderator. Loss of main pump pressure, or a loss-of-coolant accident (LOCA), results in immediate removal of the fuel to rigid sieves below the core, with consequent subcriticality. Residual and decay heat are carried away by thermal conduction through the coolant or, in the case of a LOCA, by a combination of radiation and natural convection of cover gas or incoming air from the fuel to the reactor vessel and convection of air between the vessel and steel containment wall. All decay heat removal systems are passive, though actively initiated external spray cooling of the containment can be used to reduce wall temperature. This, however, is only necessary in the case of a LOCA and after a period of 24 h. (author)

  14. Cutting Technology for Decommissioning of the Reactor Pressure Vessels in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jeong, Kwan Seong; Kim, Geun Ho; Moon, Jei Kwon; Choi, Byung Seon

    2012-01-01

    Lots of nuclear power plants have been decommissioned during the last 2 decades. An essential part of this work is the dismantling of the Reactor Pressure Vessel and its Internals. For this purpose a wide variety of different cutting technologies have been developed, adapted and applied. A detailed introduction to Plasma Arc cutting, Contact Arc Metal cutting and Abrasive Water Suspension Jet cutting is given, as it turned out that these cutting technologies are particularly suitable for these type of segmentation work. A comparison of these technologies including gaseous emissions, cutting power, manipulator requirements as well as selected design approaches are given. Process limits as well as actual limits of application are presented

  15. Device for measuring the local power in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Thor, E.R.; Korpaas, Harald; Leyse, R.H.; Smith, R.D.

    1977-01-01

    The purpose of the invention is to make possible the fabrication of a gamma ray thermometer, capable of being accurately calibrated, before being mounted in the core of the reactor, by means of a system simulating the power released by the fuel. As already indicated, the principle used for performing the measurement with a gamma ray thermometer consists, schematically, in establishing a constant and given conductance path in the mass of an absorbent body and, after calibration, deducting from the temperature difference recorded at the ends of this path the value of the calorific power absorbed by the body and therefore generated by the surrounding nuclear fuel [fr

  16. System for determining the local power in the fuel assembly of a nuclear reactor

    International Nuclear Information System (INIS)

    Rolstad, Erik; Korpas, T.-H.; Leyse, R.H.; Smith, R.D.

    1979-01-01

    System for determining the local power in the fuel assembly of a nuclear reactor which includes a rod conducting the heat and electricity, along which axial areas act as a gamma radiation thermometer. Each area includes a thermal bridge, a cold source and a pair of junctions acting as thermocouples so placed that they measure the temperature difference between the thermal bridge and the cold source. The power created by the fuel assembly near each area acting as gamma thermometer is found from this difference in temperature [fr

  17. Biological hazards of radioactivity and the biological consequences of radionuclide emissions from routine operation of nuclear power reactors

    International Nuclear Information System (INIS)

    Stendig-Lindberg, G.

    1978-01-01

    The biological hazards of radioactivity and the biological consequences of radionuclide emissions from the routine operation of nuclear power reactors are reviewed. ICRP and Scandinavian recommendations for the limitation of annual radiation doses are presented. The contribution of environmental conditions to radiation hazard is also discussed. It is concluded that a review of the justification of nuclear power is urgently needed. (H.K.)

  18. Status and trends in nuclear technology for power plants with WWER-1000 reactors. Review

    Energy Technology Data Exchange (ETDEWEB)

    Zorev, N N

    1977-04-01

    The problems of improving quality of nuclear equipment for WWER-1000 power plants and associated nuclear technology automation are surveyed. Examples of technological innovations are presented which significantly reduce labour intensity, time consumption and increase quality standards of the products. Some new automated equipments for materials welding, working, machining and quality control are described. The discussion is centering around heavy-section steel technologies. Some mechanical properties of new-developed nuclear grade steels designed for producing reactor vessels and steamgenerators, volume compensators and pipes, as well as steam separators and steamsuperheaters are also presented. Their properties (impact strength and radiation resistance) are pointed out to be superior to that of steels used abroad. The basic trend in nuclear structural material developments is towards integrated optimization of strength, performance and workability.

  19. Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

    International Nuclear Information System (INIS)

    Brooks, B.G.

    1985-03-01

    This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

  20. The safety of Ontario's nuclear power reactor. A scientific and technical review. Report to the Minister

    International Nuclear Information System (INIS)

    Hare, F.K.

    1988-01-01

    In December 1986 a study of the safety of the design, operating procedures and emergency plans associated with Ontario Hydro's nuclear generating plants was commissioned by the government of the province of Ontario. After receiving briefs from many interested groups and individuals, visiting the power plants, and consulting with nuclear industry and regulatory representatives in Canada and other countries, the commissioner presented this report to the Minister of Energy for Ontario. His major conclusion is that Ontario Hydro reactors are being operated safely and at high standards of technical performance. No significant adverse impact has been detected in either the work force or the public. The risk of accidents serious enough to affect the public adversely can never be zero, but is very remote. Major recommendations are that: Ontario Hydro re-examine its operational organization closely and commission a study of factors affecting human performance; and, that priority be given to finding a solution to pressure tube performance problems and to improving in-reactor monitoring. Sixteen other recommendations are presented relating to research and development, information exchange with other organizations, reactor performance, training, severe accident analysis, the provincial nuclear emergency plan, epidemiological studies, the Atomic Energy Control Board, public hearings, and women in the nuclear industry

  1. Nuclear reactor container

    International Nuclear Information System (INIS)

    Hosaka, Seiichi.

    1988-01-01

    Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

  2. A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

    International Nuclear Information System (INIS)

    Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

    2017-01-01

    This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

  3. Technology of nuclear reactors

    International Nuclear Information System (INIS)

    Ravelet, F.

    2016-01-01

    This academic report for graduation in engineering first presents operation principles of a nuclear reactor core. It presents core components, atomic nuclei, the notions of transmutation and radioactivity, quantities used to characterize ionizing radiations, the nuclear fission, statistical aspects of fission and differences between fast and slow neutrons, a comparison between various heat transfer fluids, the uranium enrichment process, and different types of reactor (boiling water, natural uranium and heavy water, pressurized water, and fourth generation). Then, after having recalled the French installed power, the author proposes an analysis of a typical 900 MWe nuclear power plant: primary circuit, reactor, fuel, spent fuel, pressurizer and primary pump, secondary circuit, aspects related to control-command, regulation, safety and exploitation. The last part proposes a modelling of the thermodynamic cycle of a pressurized water plant by using an equivalent Carnot cycle, a Rankine cycle, and a two-phase expansion cycle with drying-overheating

  4. Xenon spatial oscillation in nuclear power reactors:an analytical approach through non linear modal analysis

    International Nuclear Information System (INIS)

    Suarez Antola, R.

    2005-01-01

    It was proponed recently to apply an extension of Lyapunov's first method to the non-linear regime, known as non-linear modal analysis (NMA), to the study of space-time problems in nuclear reactor kinetics, nuclear power plant dynamics and nuclear power plant instrumentation and control(1). The present communication shows how to apply NMA to the study of Xenon spatial oscillations in large nuclear reactors. The set of non-linear modal equations derived by J. Lewins(2) for neutron flux, Xenon concentration and Iodine concentration are discussed, and a modified version of these equations is taken as a starting point. Using the methods of singular perturbation theory a slow manifold is constructed in the space of mode amplitudes. This allows the reduction of the original high dimensional dynamics to a low dimensional one. It is shown how the amplitudes of the first mode for neutron flux field, temperature field and concentrations of Xenon and Iodine fields can have a stable steady state value while the corresponding amplitudes of the second mode oscillates in a stable limit cycle. The extrapolated dimensions of the reactor's core are used as bifurcation parameters. Approximate analytical formulae are obtained for the critical values of this parameters( below which the onset of oscillations is produced), for the period and for the amplitudes of the above mentioned oscillations. These results are applied to the discussion of neutron flux and temperature excursions in critical locations of the reactor's core. The results of NMA can be validated from the results obtained applying suitable computer codes, using homogenization theory(3) to link the complex heterogeneous model of the codes with the simplified mathematical model used for NMA

  5. Nuclear engineering laboratory self regulated power oscillation experiments at the Health Physics Research Reactor

    International Nuclear Information System (INIS)

    Miller, L.F.; Mihalczo, J.T.; Bailiff, E.G.; Woody, N.D.; Gardner, G.D.

    1983-01-01

    Self regulated power oscillation experiments with a variety of initial conditions have been performed with the ORNL Health Physics Research Reactor (HPRR) by undergraduate nuclear engineering students from The University of Tennessee for several years. These experiments demonstrate the coupling between reactor kinetics and heat transfer and show how the temperature coefficient of reactivity affects reactor behavior. A model that consists of several coupled first order nonlinear differential equations is used to calculate the temperature of the core center and surface and power as a function of time which are compared with the experimental data; also, the model is also used to study the effects of various model parameters and initial conditions on the amplitude, frequency and damping of the power and temperature oscillations. A previous paper presented some limited experimental results and demonstrated the correspondence between a simple point model and the experimental data. This paper presents the results of experiments for: (1) the initial power fixed at 9 kW with central core temperatures of 300 0 F and 500 0 F, annd (2) the initial central core temperature fixed at 500 0 F with initial powers of 6 and 8 kW

  6. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Werner, James Elmer [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; McKellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Kennedy, John Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Biersdorf, John Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division

    2017-04-01

    The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

  7. 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

  8. Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

    Science.gov (United States)

    Tsibulskiy, V. F.; Andrianova, E. A.; Davidenko, V. D.; Rodionova, E. V.; Tsibulskiy, S. V.

    2017-12-01

    A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium-plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

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

  10. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su'ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  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. Impact of nuclear power and public acceptance. Licensing of nuclear reactors and public acceptance in Austria

    International Nuclear Information System (INIS)

    Plochl, Gerhardt.

    1977-01-01

    This paper first reviews the main stages in the development of nuclear activities in Austria and recalls the principal legal and regulatory texts adopted in this field. In this context, the author describes the circumstances which, following announcement of the project for a second nuclear power plant in 1974, gave rise to hostile reactions and led to the organisation by the federal authorities of a wide public information campaign in 1976 and 1977. This resulted in increasing the number of opponents to nuclear energy and raised the question of the long term storage of radioactive waste, for which at present Austrian regulations provide no satisfactory solution. (NEA) [fr

  13. Energy Research Advisory Board, Civilian Nuclear Power Panel: Subpanel 1 report, Light water reactor utilization and improvement: Volume 2

    International Nuclear Information System (INIS)

    1986-10-01

    The Secretary of Energy requested that the Office of Nuclear Energy prepare a strategic national plan that outlines the Department's role in the future development of civilian nuclear power and that the Energy Research Advisory Board establish an ad hoc panel to review and comment on this plan. The Energy Research Advisory Board formed a panel for this review and three subpanels were formed. One subpanel was formed to address the institutional issues surrounding nuclear power, one on research and development for advanced nuclear power plants and a third subpanel on light water reactor utilization and improvement. The subpanel on light water reactors held two meetings at which representatives of the DOE, the NRC, EPRI, industry and academic groups made presentations. This is the report of the subpanel on light water reactor utilization and improvement. This report presents the subpanel's assessment of initiatives which the Department of Energy should undertake in the national interest, to develop and support light water reactor technologies

  14. Concept of a nuclear powered submersible research vessel and a compact reactor

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa; Nishimura, Hajime; Tokunaga, Sango

    2001-07-01

    A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

  15. Economic analysis of multiple-module high temperature gas-cooled reactor (MHTR) nuclear power plants

    International Nuclear Information System (INIS)

    Liu Yu; Dong Yujie

    2011-01-01

    In recent years, as the increasing demand of energy all over the world, and the pressure on greenhouse emissions, there's a new opportunity for the development of nuclear energy. Modular High Temperature Gas-cooled Reactor (MHTR) received recognition for its inherent safety feature and high outlet temperature. Whether the Modular High Temperature Gas-cooled Reactor would be accepted extensively, its economy is a key point. In this paper, the methods of qualitative analysis and the method of quantitative analysis, the economic models designed by Economic Modeling Working Group (EMWG) of the Generation IV International Forum (GIF), as well as the HTR-PM's main technical features, are used to analyze the economy of the MHTR. A prediction is made on the basis of summarizing High Temperature Gas-cooled Reactor module characteristics, construction cost, total capital cost, fuel cost and operation and maintenance (O and M) cost and so on. In the following part, comparative analysis is taken measures to the economy and cost ratio of different designs, to explore the impacts of modularization and standardization on the construction of multiple-module reactor nuclear power plant. Meanwhile, the analysis is also adopted in the research of key factors such as the learning effect and yield to find out their impacts on the large scale development of MHTR. Furthermore, some reference would be provided to its wide application based on these analysis. (author)

  16. Investigation for calculation methods used in analyzing the physics characteristics of nuclear power reactor

    International Nuclear Information System (INIS)

    Nguyen Tuan Khai; Hoang Van Khanh; Phan Quoc Vuong; Tran Viet Phu; Tran Vinh Thanh; Nguyen Thi Mai Huong; Nguyen Thi Dung; Le Tran Chung; Nguyen Minh Tuan; Tran Quoc Duong

    2014-01-01

    The project aims at nuclear human resource development and enhancement in research capability in reactor physics and kinetics at Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat). The main research items of the project can be summarized as follows: i) Considering possibility on using modern calculation techniques and methods in investigating neutronic characteristics and neutronics-thermal hydraulics coupling. This item is proposed to carry out based on international collaboration with Prof. Le Trong Thuy, San Jose University, US; ii) Carrying out the collaborative activities in research and training between Nuclear Energy Center (Institute for Nuclear Science and Technology) and Nuclear Reactor Center (Nuclear Research Institute, Dalat); iii) Opening two-week training course on nuclear reactor engineering (25 Nov - 12 Dec 2013) in collaboration with Japan Atomic Energy Agency (JAEA). (author)

  17. Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

    International Nuclear Information System (INIS)

    Foerster, S.; Hewing, G.

    1977-01-01

    On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

  18. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    International Nuclear Information System (INIS)

    1986-09-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

  19. Present state of inspection robot technology in nuclear power facilities. Case of fast breeder reactors

    International Nuclear Information System (INIS)

    Ara, Kuniaki

    1995-01-01

    In the maintenance works in nuclear power facilities such as checkup, inspection and repair, for the main purpose of radiation protection, remote operation technology was introduced since relatively early stage, and at present, the robots that carry out the inspection works for confirming the soundness of main equipment have been developed and put to practical use. At the time of introducing these technologies, in addition to the research and development of robots proper, the coordination with the design of plant machinery and equipment facilities as the premise of introducing robots is an important requirement. In this report, the present state of the development of remote inspection technology for fast breeder reactors is introduced, and the matters to which attention is paid in the plant design for introducing robots are explained. First, fast breeder reactors are described. The needs of robotizing and adopting remote operation in nuclear power facilities are explained, using the examples of the inspection system for a reactor vessel and the inspection system for steam generator heat transfer tubes. (K.I.)

  20. 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

  1. Development of a power-period calculation unit for nuclear reactor Control

    International Nuclear Information System (INIS)

    Martin, J.

    1966-10-01

    The apparatus studied is a digital calculating assembly which makes it possible to prepare and to present numerically the period and power of a nuclear reactor during operation, from start-up to nominal power. The pulses from a fission chamber are analyzed continuously, using real time. A small number of elements is required because of the systematic use of a calculation technique comprising the determination of a base 2 logarithm by a linear approximation. The accuracy obtained for the period is of the order of 14%; the response time of the order of the calculated period value. An approximate value of the power (30%) is given at each calculation cycle together with the power thresholds required for the control. (author) [fr

  2. Medical surveillance of nuclear power plant workers during reactor shutdown using whole-body counting and excretion analysis

    International Nuclear Information System (INIS)

    Le Roux-Desmis, C.

    1987-01-01

    After a review of radioactivity basis and radiation protection principles, the various aspects of medical surveillance of nuclear power plant workers during reactor shutdown, are presented. Internal contamination incidents that happened during 1986-1987 shutdown of Paluel reactor are exposed. Internal contamination levels are evaluated using whole-body counting and radionuclide determination in feces and urine and compared with dose limits [fr

  3. Indian experience with radionuclide transport, deposition and decontamination in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Narasimhan, S.V.; Das, P.C.; Lawrence, D.A.; Mathur, P.K.; Venkateswarlu, K.S.

    1983-01-01

    The present generation of water-cooled nuclear reactors uses construction materials chosen with utmost care so that minimum corrosion occurs during the life of the reactor. As interaction between the primary coolant and the construction materials is unavoidable, the coolant is chemically treated to achieve maximum compatibility. First measurements of the chemical and radiochemical composition of the crud present on the in-core and out-of-core primary heat transport system surfaces of a pressurized heavy-water-moderated and cooled reactor (PHWR) are given; then experience in India in the development of a low temperature, one-stage decontaminating formulation for chemical decontamination of the radioactive deposits formed on stainless steel surfaces under BWR conditions is discussed. The effect of the magnitude of the transients in parameters such as reactor power, system temperature, dissolved oxygen content in the coolant, etc. on the nature and migration behaviour of primary heat transport system crud in a PHWR is described. Contributions to radioactive sources and insoluble crud from different primary heat transport system materials are identified and correlated with reactor operations in a PHWR. Man-rem problems faced by nuclear reactors, especially during off-line maintenance, stress the need for reducing the deposited radioactive sources from system surfaces which would otherwise be accessible. Laboratory and on-site experimentation was carried out to effect chemical decontamination on the radioactive deposits formed on the stainless steel surfaces under BWR conditions. Both the reducing and oxidizing formulations were subsequently used in a small-scale, in-plant trial in the clean-up system of a BWR. More than 85% of the deposited 60 Co activity was found to have been removed by the oxidizing formulation. Efforts to develop a decontaminating mixture containing a reducing agent with the help of a circulating loop are in progress in the laboratory. (author)

  4. The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Maoyao; Peng, Minjun; Cheng Shouyu [Harbin Engineering University, Harbin (China)

    2014-08-15

    In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions.

  5. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  6. The Research on Operation Strategy of Nuclear Power Plant with Multi-reactors

    International Nuclear Information System (INIS)

    Fang, Maoyao; Peng, Minjun; Cheng Shouyu

    2014-01-01

    In this paper, the operation characteristics and control strategy of nuclear power plant (NPP) with multi-modular pressurized water reactors (PWR) were researched through simulation. The main objective of this research was to ensure the coordinated operation and satisfy the convenience of turbine-generator and reactor's load adjustment in NPP with multi-reactors (MR). According to the operation characteristics of MR-NPP, the operation and control strategy was proposed, which was 'he average allocation of load for each reactor and maintaining average temperature of coolant at a constant? The control system was designed based the operation and control strategy. In order to research the operation characteristics and control strategy of MR-NPP, the paper established the transient analysis model which included the reactors and thermal hydraulic models, turbine model, could simulate and analyze on different operating conditions such as load reducing, load rising. Based on the proposed operation and control strategy and simulation models, the paper verified and validated the operation strategy and control system through load reducing, load rising. The results of research simulation showed that the operation strategy was feasible and can make the MR-NPP running safely as well as steadily on different operating conditions

  7. Training courses for the staff of the nuclear power station KRSKO conducted at the TRIGA reactor center in Ljubljana

    International Nuclear Information System (INIS)

    Pregl, G.; Najzer, M.

    1976-01-01

    The training program for the Nuclear Power Station Krsko was divided into two modules: fundamentals of nuclear engineering and specialized training according to duties that candidates are supposed to take at the power station. Basic training was organized at the TRIGA Reactor Center in Ljubljana in two different versions. The first version intended for plant operators and all engineers lasted for six months and included about 500 hours of classroom lessons and seminars and 31 laboratory experiments. The educational program was conventional. The following topics were covered: nuclear and atomic physics, reactor theory, reactor dynamics, reactor instrumentation and control, heat transfer in nuclear power plants, nuclear power plant systems, reactor materials, reactor safety, and radiation protection. Until now, two groups, consisting of 37 candidates altogether, have attended this basic course. Plans have been made to conduct two additional courses of about 20 students each for technicians other than operators. The program of this second version will be reduced, with the emphasis on reactor core physics and radiation protection. Classroom lessons will be strongly supported by laboratory experiments. (author)

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

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

  9. Views on quality assurance at Finnish and Swedish nuclear power plants and at Halden Reactor

    International Nuclear Information System (INIS)

    Hammar, L.; Lidh, B.; Wahlstroem, B.; Reiman, T.

    2001-06-01

    The paper reports on a study within the Nordic Nuclear Safety Research, NKS on quality systems at nuclear installations in Finland, Norway and Sweden. In the study a total of 74 people at the NPPs in Barsebaeck, Forsmark, Loviisa, Olkiluoto, Oskarshamn and Ringhals, and at the research reactor in Halden were interviewed in the period 30 August to 13 December 2000 concerning their views in regard of quality and quality systems. The study was concluded with a seminar held in the Ringhals nuclear power plant in Januar 2001. The study covered a number of aspects in regard of quality management, including the quality concept, quality systems, topical quality issues and approaches, rules and procedures, competency and training, the process approach to quality management, the promotion of quality consciousness and future prospects. The study reflects the significant progress made in the management of quality in nuclear power in the Nordic countries since the early phase in the seventies. The most distinctive characteristic of today's approach to quality is seen in that responsibility for the quality is assumed directly in conjunction with the working processes. It could be noted that the work patterns at the nuclear installations have been largely modified during the recent years as a result of persistent endeavours to continuously improve the quality of operation. Challenges were seen in currently reduced revenues due to descending electricity prices and the likely prospect of further increased regulatory safety requirements. The report is aimed for those working with quality issues at the nuclear power plants as well as for those interested in quality management in general or in the safety aspects of nuclear power in particular. (au)

  10. Floating nuclear heat. And power station 'Pevec' with KLT-40S type reactor plant for remote regions of Russia

    International Nuclear Information System (INIS)

    Veshnyakov, K.B.; Kiryushin, A.I.; Panov, Yu.K.; Polunichev, V.I.

    2000-01-01

    Floating small nuclear power plants power for local energy systems of littoral regions of Russia, located far from central energy system, open a new line in nuclear power development. Designing a floating power unit of a lead nuclear heat and power generating station for port Pevec at the Chuckchee national district is currently nearing completion. Most labor-intensive components are being manufactured. The co-generation NPP Pevec is to be created on the basis of a floating power unit with KLT-40S type reactor plant. KLT-40S reactor plant is based on similar propulsion plants, verified at operation of Russia's nuclear-powered civil ships, evolutionary improved by elimination of 'weak points' revealed during its prototypes operation or on the basis of safety analysis. KLT-40S reactor plant uses the most wide-spread and developed in the world practice PWR-type reactor. KLT-40S meets contemporary national and international requirements imposed to future reactor plants. The NHPS description, its main technical-economic data, environmental safety indices, basic characteristics of KLT-40S reactor plant are presented. Prospects of small NPPs utilization outside Russia, particularly as an energy source for sea water desalination, are considered. (author)

  11. Reactor pressure vessel life cycle management at the Calvert Cliffs Nuclear Power Plant

    International Nuclear Information System (INIS)

    Doroshuk, B.W.; Bowman, M.E.; Henry, S.A.; Pavinich, W.A.; Lapides, M.E.

    1993-01-01

    Life Cycle Management (LCM) seeks to manage the aging process of important systems, structures, and components during licensed operation. The goal of Baltimore Gas and Electric Company's (BG and E) Life Cycle Management Program is to assure attainment of 40 years of operation and to preserve the option of an additional 20 years of operation for the Calvert Cliffs Nuclear Power Plant (CCNPP). Since the reactor pressure vessel (RPV) has been identified as one of the most critical components with regard to long-term operation of a nuclear power plant, BG and E initiated actions to manage life limiting or aging issues for the CCNPP RPVs. To achieve long-term operation, technical RPV issues must be effectively managed. This paper describes methods BG and E uses for managing RPV age-related degradation. (author)

  12. Transportable nuclear power plant T3C-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Ogloblin, B.G.; Gromov, B.F.

    1993-01-01

    Development and cultivating of districts in Siberia, North, Far East, Kamchatka and other remote or almost inaccessible district of the country depends to a large degree on their providing with power. The specific character of these districts imposes in turn a wide variety of special requirements upon the power sources. In particular, it is essential to provide the following; maximum manufacture availability of the whole equipment at the minimum volume of construction and installation work on operation site, high safety, longterm service life, ecologically, minimum scope of work on equipment in-service maintenance and inspection, etc. Taking into account the well-known difficulties connected with the delivery of conventional energy carriers to the above-mentioned districts and the situation with the alternative power sources, the application of the low-power nuclear plants (NPP) for these purposes looks definitely promising. Among the probable trends in creating the NPPs of this type as very promising is considered the possibility to apply the two-circuit reactor plant of the vessel type with the liquid lead as a primary coolant and free air as a secondary coolant and working medium in the open gas-turbine cycle. The nuclear plant T3C-M of improved safety with two of this type reactor plants with total electric power of 8 MW is developed by CDB of Machine Building with participation of several enterprises of St. Petersburg under the scientific leadership and is intended for generation of electric power and up to 4 Gcal/h of heat for populated areas and installations placed at long distance from the main electric power supply sources where it is difficult or non-efficient economically to deliver the conventional kinds of fuel. The main principles being laid as a basis when developing the proposed NPP will allow one to create mobile power sources which possess a high degree of safety and inherent self-protection

  13. Modular nuclear reactor for a land-based power plant and method for the fabrication installation and operation thereof

    International Nuclear Information System (INIS)

    Craig, E. R.; Blumberg, B. Jr.

    1985-01-01

    A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors

  14. Nuclear reactors to come

    International Nuclear Information System (INIS)

    Lung, M.

    2002-01-01

    The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

  15. Modification of reactor installation in the Genkai nuclear power plant No. 1 of Kyushu Electric Power Company, Inc

    International Nuclear Information System (INIS)

    1979-01-01

    The Nuclear Safety Commission recognized the adequacy concerning the inquiry which was offered from the Minister of International Trade and Industry on July 25, 1979, following the safety evaluation in the Ministry of International Trade and Industry, and decided to submit a report to the Minister of International Trade and Industry on July 26, 1979, about the modification of reactor installation in the Genkai nuclear power plant No. 1 of the Kyushu Electric Power Company, Inc. This is concerned to the application which was made from the president of the Kyushu Electric Power Company, Inc., to the Minister of International Trade and Industry on July 24, 1979. The content of the modification is to add a control circuit which is actuated by the signal of abnormal low pressure in a reactor to the circuit of actuating the emergency core cooling system of the plant. The influences on the safety protection system by the addition of the circuit transmitting safety injection signal and by the additions of an interlock circuit preventing the misoperation of pressurizer spray and of a block circuit of safety injection signal in case of the abnormal low pressure in a reactor were evaluated. The effects on the function and characteristics of the emergency core cooling system due to the addition of the control circuit were investigated, and it was recognized by the analysis that there is no effect in the pipe ruptures of both small and large scales. (Nakai, Y.)

  16. Nuclear reactor design

    CERN Document Server

    2014-01-01

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

  17. Radioactive waste management in nuclear power plants with WWER-type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dlouhy, Z; Napravnik, J; Safar, O

    1975-05-01

    The possibilities of radioactive waste solidification in nuclear power plants with LWR reactors (of the WWER type) and the problems of their safe storage in Czechoslovakia are discussed. The most suitable method for the treatment of emitted sorbents and concentrates seems to be their incorporation in bitumen or concrete. In the disposal of solidified blocks all requirements should be met including the selection of suitable sites and of convenient methods of transportation. A preliminary economic estimate shows that the storage of bitumen-incorporated wastes in trenches seems to be less expensive from the point of view of exploitation of the storage facility as well as from the point of view of investment.

  18. Design of power auto-regulating system's high reliability controller for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    An Zhencai; Liu Longzhi; Chen Yuan

    1996-01-01

    The paper mainly introduces power auto-regulating system's high reliability controller for 200 MW Nuclear Heating Reactor. The controller is implemented with excellent performance 16 bit single chip microcomputer 8097. Master controller and 10 digit samplers are blocked. Each and every block's hardware is identical. These blocks communicate each other through 8 bit BUS and operate synchronously by united clock and reset signal and are designed with three redundancies. The identity comparison principle through two-out-of three is also introduced. The test proves that designing scheme is feasible

  19. International conference on management of spent fuel from nuclear power reactors. Book of extended synopses

    International Nuclear Information System (INIS)

    2006-01-01

    This document contains 48 extended synopses of the International Conference on Management of Spent Fuel from Nuclear Power Reactors. The major topics covered related to national programmes in spent fuel management as well as regional trends, technology and safety/security aspects of wet and dry storage, licensing and regulation, quality assurance, design control, operating experience, R and D, and special aspects of spent fuel storage including in-service inspection, robotics, heat removal, and other engineering considerations. Each of the extended synopses was indexed separately

  20. Data list of nuclear power plants of pressurized-water reactor type in Japan

    International Nuclear Information System (INIS)

    Izumi, Fumio; Harayama, Yasuo

    1981-08-01

    This report has collected and compiled the data concerning performances, equipments and installations for nuclear power plants of the pressurized-water reactor type in Japan. The data used in the report are based on informations that were collected before December in 1980. The report is edited by modifing changes of the data appeared after publication of 1979 edition (JAERI-M 8947), and extending the data-package to cover new plants proposed thereafter. All data have been processed and tabulated with a computer program FREP, which has been developed as an exclusive use of data processing. (author)

  1. Federal supervisory authorities' power to issue directives in nuclear licensing procedures (Kalkar reactor)

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The Federal Constitutional Court (BVerfG) is concerned in its judgement with the legal instrument of Federal supervisory authorities' power to issue directives in nuclear licensing procedures (Kalkar reactor). Dealt with are questions concerning material and excercising competences, the Federal Government's right to issue directives and possible violation of Laender rights, as well as the legal position between Land and Federal Government, the necessity to issue clear directives, questions concerning the competence to issue directives and the Federal Government's duty to consider the overall interests (Federal Government and Laender interests), questions concerning constitutional state principles and the limits in the relations between Federal Government - Laender concerning legal competence. (RST) [de

  2. On the reliability of steam generator performance at nuclear power plants with WWER type reactors

    International Nuclear Information System (INIS)

    Styrikovich, M.A.; Margulova, T.Kh.

    1974-01-01

    The problem of ensuring reliable operation of steam generators in a nuclear power plant with a water-cooled, water-moderated reactor (WWER) was studied. At a nuclear power plant with a vertical steam generator (specifically, a Westinghouse product) the steam generator tubes were found to have been penetrated. Shutdown was due to corrosion disintegration of the austenitic stainless steel, type 18/8, used as pipe material for the heater surface. The corrosion was the result of the action of chlorine ions concentrated in the moisture contained in the iron oxide films deposited in low parts of the tube bundle, directly at the tube plate. Blowing through did not ensure complete removal of the film, and in some cases the construction features of the steam generator made removal of the film practically impossible. Replacement of type 18/8 stainless steel by other construction material, e.g., Inconel, did not give good results. To ensure reliable operation of vertical steam generators in domestic practice, the generators are designed without a low tube plate (a variant diagram of the vertical steam generator of such construction for the water-cooled, water-moderated reactor 1000 is presented). When low tube plates are used the film deposition is intolerable. For organization of a non-film regime a complex treatment of the feed water is used, in which the amount of complexion is calculated from the stoichmetric ratios with the composition of the feed water. It is noted that, if 100% condensate purification is used with complexon processing of the feed water to the generator, we can calculate the surface of the steam-generator heater without considering the outer placement on the tubes. In this the cost of the steam generator and all the nuclear power plants with WWER type reactors is decreased even with installation of a 100% condensate purification. It is concluded that only simultaneous solution of construction and water-regime problems will ensure relaible operation of

  3. CPU time reduction strategies for the Lambda modes calculation of a nuclear power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vidal, V.; Garayoa, J.; Hernandez, V. [Universidad Politecnica de Valencia (Spain). Dept. de Sistemas Informaticos y Computacion; Navarro, J.; Verdu, G.; Munoz-Cobo, J.L. [Universidad Politecnica de Valencia (Spain). Dept. de Ingenieria Quimica y Nuclear; Ginestar, D. [Universidad Politecnica de Valencia (Spain). Dept. de Matematica Aplicada

    1997-12-01

    In this paper, we present two strategies to reduce the CPU time spent in the lambda modes calculation for a realistic nuclear power reactor.The discretization of the multigroup neutron diffusion equation has been made using a nodal collocation method, solving the associated eigenvalue problem with two different techniques: the Subspace Iteration Method and Arnoldi`s Method. CPU time reduction is based on a coarse grain parallelization approach together with a multistep algorithm to initialize adequately the solution. (author). 9 refs., 6 tabs.

  4. Nuclear power flies high

    International Nuclear Information System (INIS)

    Friedman, S.T.

    1983-01-01

    Nuclear power in aircraft, rockets and satellites is discussed. No nuclear-powered rockets or aircraft have ever flown, but ground tests were successful. Nuclear reactors are used in the Soviet Cosmos serles of satellites, but only one American satellite, the SNAP-10A, contained a reactor. Radioisotope thermoelectric generators, many of which use plutonium 238, have powered more than 20 satellites launched into deep space by the U.S.A

  5. Researching and improving the reliability of reactor protection system of Qinshan nuclear power plant

    International Nuclear Information System (INIS)

    Jiang Zuyue; Sheng Jiannan

    1997-01-01

    Due to the original design defects of the Reactor Protection System (RPS) of Qinshan Nuclear Power Plant, this system has brought about a number of reactor shutdown accidents and Engineered Safety Features (ESF) mis-activation events which have seriously endangered safe and steady operation of the nuclear power plant. So over three years have been spent on research on the reform of the original design on the premise that the general wiring of the system should remain the same and that the system size should also remain small to be contained in the original cabinets. The following improvements were made: (1) Increase the system's anti-disturbance capability. The system's zero volt bus floating designs were modified to surmount the disturbance resulting from the bad isolation performance of impedance-isolated amplifier; Double grounds have been added to logical modules to surmount the disturbance resulting from zero volt floating bus during the replacement of single module with two connectors; The opto-coupling circuit in its oscillation input stage of Engineered Safety Features have been improved to increase its reliability. (2)Modify to output activation part of the system. The new type of output relays were selected and the relay activation circuits were redesigned in which switcher activation mode is used instead of amplifier activation mode so as to increase the reliability of relay operation and reduce the power consumption; CMOS buffer gates in the input and output stage of the circuit were used to match TTL circuits to CMOS circuits of the system

  6. Technical and economic proposal for the extension of the Laguna Verde Nuclear Power plant with an additional nuclear reactor

    International Nuclear Information System (INIS)

    Leal C, C.D.; Francois L, J.L.

    2006-01-01

    The increment of the human activities in the industrial environments and of generation of electric power, through it burns it of fossil fuels, has brought as consequence an increase in the atmospheric concentrations of the calls greenhouse effect gases and, these in turn, serious repercussions about the environment and the quality of the alive beings life. The recent concern for the environment has provoked that industrialized countries and not industrialized carry out international agreements to mitigate the emission from these gases to the atmosphere. Our country, like part of the international community, not is exempt of this problem for what is necessary that programs begin guided toward the preservation of the environment. As for the electric power generation, it is indispensable to diversify the sources of primary energy; first, to knock down the dependence of the hydrocarbons and, second, to reduce the emission of polluting gases to the atmosphere. In this item, the nucleo electric energy not only has proven to be safe and competitive technical and economically, able to generate big quantities of electric power with a high plant factor and a considerable cost, but rather also, it is one of the energy sources that less pollutants it emits to the atmosphere. The main object of this work is to carry out a technical and economic proposal of the extension of the Laguna Verde Nuclear power plant (CNLV) with a new nuclear reactor of type A BWR (Advanced Boiling Water Reactor), evolutionary design of the BWR technology to which belong the two reactors installed at the moment in the plant, with the purpose of increasing the installed capacity of generation of the CNLV and of the Federal Commission of Electricity (CFE) with foundation in the sustainable development and guaranteeing the protection of the environment by means of the exploitation of a clean and sure technology that counts at the moment with around 12,000 year-reactor of operational experience in more of

  7. Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

    International Nuclear Information System (INIS)

    Boldyrev, V.M.; Mikhaj, V.I.

    1985-01-01

    Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation

  8. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

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

  9. Low-temperature thermionics in space nuclear power systems with the safe-type fast reactor

    International Nuclear Information System (INIS)

    Zrodnikov, A.V.; Yarygin, V.I.; Lazarenko, G.E.; Zabudko, A.N.; Ovcharenko, M.K.; Pyshko, A.P.; Mironov, V.S.; Kuznetsov, R.V.

    2007-01-01

    The potentialities of the use of the low-temperature thermionic converters (TIC) with the emitter temperature ≤ 1500 K in the space nuclear power system (SNPS) with the SAFE-type (Safe Affordable Fission Engine) fast reactor proposed and developed by common efforts of American experts have been considered. The main directions of the 'SAFE-300-TEG' SNPS (300 kW(thermal)) design update by replacing the thermoelectric converters with the low-temperature high-performance thermionic converters (with the barrier index V B ≤ 1.9 eV and efficiency ≥ 10%) meant for a long-term operation (5 years at least) as the components of the SAFE-300-TIC SNPS for a Lunar base have been discussed. The concept of the SNPS with the SAFE-type fast reactor and low-temperature TICs with specific electric power of about 1.45 W/cm 2 as the components of the SAFE-300-TIC system meeting the Nasa's initial requirements to a Lunar base with the electric power demand of about 30 kW(electrical) for robotic mission has been considered. The results, involving optimization and mass-and-size estimation, show that the SAFE-300-TIC system meets the initial requirements by Nasa to the lunar base power supply. The main directions of the system update aimed at the output electric power increase up to 100 kW(electrical) have also been presented. (authors)

  10. HYPER (hybrid power extraction reactor): a system for clean nuclear energy

    International Nuclear Information System (INIS)

    Park, W.S.; Shin, U.; Han, S.-J.; Song, T.Y.; Choi, B.H.; Park, C.K.

    2000-01-01

    The Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development (RID) called HYPER (hybrid power extraction reactor) for the transmutation of nuclear waste and energy production through the transmutation process. HYPER program is within the frame work of the national mid and long-term nuclear research plan. KAERI is aiming to develop the elemental technologies for the subcritical transmutation system by the year of 2001 and build a small bench scale test facility (∝5 MW) by the year of 2006. Some major features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead-bismuth (Pb-Bi) is adopted as a coolant and spallation target material. 1 GeV 16 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MW of power. The support ratio of HYPER for LWR units producing the same power is believed to be 5∝6. (orig.)

  11. Transference of know-how for the fabrication of heavy components for nuclear power reactors

    International Nuclear Information System (INIS)

    Meier, F.

    1977-01-01

    1) Heavy components for nuclear power reactors. Reactor pressure vessels with total weight of 540 tons; steam generators: heat exchangers with U-type tube bundles, total weight 420 tons. 2) Choice of know-how recipient. Technical criteria, i.e. manufacturing facilities, existing quality assurance system, location of the workshops, possibilities for training, infrastructures. 3. Measures for transferring know-how to a newly established company. Planning and erection of the factory: organisational set up of the company; personnel selection and training; transfer of documentation; transfer of know-how that cannot be transferred in a written form. 4) Contracts for assuring the transfer of know-how. Stipulation of mutual rights and obligations of the know-how owner and receiver in individual contracts: engineering services contract, technical information contract, personnel training contract, license contract. (orig.) [de

  12. Aging management strategy for reactor internals of Korean nuclear power plants

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

    2010-01-01

    This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

  13. Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

    2011-01-01

    The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

  14. Training simulator for nuclear power plant reactor control model and method

    International Nuclear Information System (INIS)

    Czerbuejewski, F.R.

    1975-01-01

    A description is given of a method and system for the real-time dynamic simulation of a nuclear power plant for training purposes, wherein a control console has a plurality of manual and automatic remote control devices for operating simulated control rods and has indicating devices for monitoring the physical operation of a simulated reactor. Digital computer means are connected to the control console to calculate data values for operating the monitoring devices in accordance with the control devices. The simulation of the reactor control rod mechanism is disclosed whereby the digital computer means operates the rod position monitoring devices in a real-time that is a fraction of the computer time steps and simulates the quick response of a control rod remote control lever together with the delayed response upon a change of direction

  15. Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1981-01-01

    The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

  16. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1991

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1993-07-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1991. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1991 annual reports submitted by about 436 licensees indicated that approximately 206,732 individuals were monitored, 182,334 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.15 rem (cSv) and an average measurable dose of about 0.31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 96,231 individuals completed their employment with one or more of the 436 covered licensees during 1991. Some 68,115 of these individuals terminated from power reactor facilities, and about 7,763 of them were considered to be transient workers who received an average dose of 0.52 rem (cSv)

  17. Characteristics of Soil Structure Interaction for Reactor Building of Kashiwazaki-Kariwa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gil, Moon Joo; Jung, Rae Young; Hyun, Chang Hun; Kim, Moon Soo; Lim, Nam Hyoung

    2010-01-01

    On 16 July 2007, the Nigataken-chuetsu-oki earthquake registering a moment magnitude of 6.8 occurred at a depth of about 15 km. As a result of this earthquake, noticeable shaking exceeding the design ground motion was measured at the Tokyo Electric Power Company (TEPCO) Kashiwazaki-Kariwa Nuclear Power Station (KKN), the biggest nuclear power plant in the world, located at about 16 km away from the epicenter. This earthquake triggered a fire at an electrical transformer and insignificant damage on some parts of facilities. This event gave an impulse to study on the damage and safety margin of nuclear power plant due to the strong earthquake exceeding design basis. As a part of those efforts, KARISMA (KAshiwazaki-Kariwa Research Initiative for Seismic Margin Assessment) benchmark study was launched by the IAEA in terms of an international collaborative research. The main objectives of this research are to estimate the structural behavior and to evaluate the seismic margin of reactor building considering the effects of Soil-Structure Interaction (SSI). This paper presents verification of structural model developed here and validation of soil foundation characteristics through soil-column analysis. It has also been demonstrated that the spring constants and damping coefficient obtained from impedance analysis represent well the soil foundation characteristics

  18. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1989

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1992-04-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC 1 licensees during the years 1969 through 1989. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC 1 licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1989 annual reports submitted by about 448 licensees indicated that approximately 216,294 individuals were monitored 111,000 of whom were monitored by nuclear power facilities. They incurred an average individual does of 0.18 rem (cSv) and an average measurable dose of 0.36 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,535 individuals completed their employment with one or more of the 448 covered licensees during 1989. Some 76,561 of these individuals terminated from power reactor facilities, and about 10, 344 of them were considered to be transient workers who received an average dose of 0.64 rem (cSv)

  19. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1988

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1991-07-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1988. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1988 annual reports submitted by about 429 licensees indicated that approximately 220,048 individuals were monitored, 113,00 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.20 rem (cSv) and an average measurable dose of 0.41 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,072 individuals completed their employment with one or more of the 429 covered licensees during 1988. Some 80,211 of these individuals terminated from power reactor facilities, and about 8,760 of them were considered to be transient workers who received an average dose of 0.27 rem (cSv). 17 refs., 11 figs., 29 tabs

  20. Power generation by nuclear power plants

    International Nuclear Information System (INIS)

    Bacher, P.

    2004-01-01

    Nuclear power plays an important role in the world, European (33%) and French (75%) power generation. This article aims at presenting in a synthetic way the main reactor types with their respective advantages with respect to the objectives foreseen (power generation, resources valorization, waste management). It makes a fast review of 50 years of nuclear development, thanks to which the nuclear industry has become one of the safest and less environmentally harmful industry which allows to produce low cost electricity: 1 - simplified description of a nuclear power generation plant: nuclear reactor, heat transfer system, power generation system, interface with the power distribution grid; 2 - first historical developments of nuclear power; 3 - industrial development and experience feedback (1965-1995): water reactors (PWR, BWR, Candu), RBMK, fast neutron reactors, high temperature demonstration reactors, costs of industrial reactors; 4 - service life of nuclear power plants and replacement: technical, regulatory and economical lifetime, problems linked with the replacement; 5 - conclusion. (J.S.)