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Sample records for reactor protections formulaire

  1. Handbook for the calculation of reactor protections; Formulaire sur le calcul de la protection des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-07-01

    This note constitutes the first edition of a Handbook for the calculation of reactor protections. This handbook makes it possible to calculate simply the different neutron and gamma fluxes and consequently, to fix the minimum quantities of materials necessary under general safety conditions both for the personnel and for the installations. It contains a certain amount of nuclear data, calculation methods, and constants corresponding to the present state of our knowledge. (authors) [French] Cette note constitue la premiere edition du 'Formulaire sur le calcul de la protection des reacteurs'. Ce formulaire permet de calculer de facon simple les difterents flux de neutrons et de gamma et, par suite, de fixer les quantites minima de materiaux a utiliser pour que les conditions generales de securite soient respectees, tant pour le personnel que pour les installations. Il contient un certain nombre de donnees nucleaires, de methodes de calcul et de constantes correspondant a l'etat actuel de nos connaissances. (auteurs)

  2. Basic nuclear data and reactor shielding design formulaire PROPANE Do

    International Nuclear Information System (INIS)

    Estiot, J.C.; Salvatores, M.; Trapp, J.P.

    1979-01-01

    This paper presents a calculational scheme - formulaire PROPANE - to calculate the deep neutron penetration in the fast reactor shield. The emphasis is put on the multigroup data and method approximations. The performances of this formulaire are presented

  3. European ERANOS formulaire for fast reactor core analysis

    International Nuclear Information System (INIS)

    Rimpault, Gerald

    2003-01-01

    ERANOS code scheme was developed within the European collaboration on fast reactors. It contains all the functions required to calculate a complete set of core, shielding and fuel cycle parameters for LMFR cores. Nuclear data are taken from recent evaluations (JEF2.2) and adjusted on integral experiments (ERALIB1). Calculational scheme uses the ECCO cell code to generate cross section data. Whole core calculations are carried out using the spatial modules BISTRO (Sn) and TGVNARIANT (nodal method). Validation is based on integral and power reactor experiments. Integral experiments are also used for adjustment of nuclear data

  4. Tables of formulae for calculating the mechanics of stacks in gas-graphite reactors; Formulaire pour le calcul de la mecanique des empilements des reacteurs graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-07-01

    This collection of formulae only gives, for nuclear graphite stacks. The mechanical effects due to the strains, thermal or not, of steel structures supporting or surrounding graphite blocks. Equations have been established by mean of experiments made at Chinon with large pile models. Thus, it is possible to calculate displacement, strain and stress in the EDF type stacks of horizontal triangular block lattice. (authors) [French] Le domaine de ce formulaire est strictement limite aux effets mecaniques, pour les empilements, des deformations, thermiques ou autres, des structures metalliques de soutien (aire - support et corset). On propose un ensemble de relations qui ont ete etablies a la suite des essais de CHINON sur des maquettes de grande taille. Ces relations permettent le calcul des mouvements, des deformations et des contraintes dans les empilements du type EDF, a reseau horizontal triangulaire regulier. (auteurs)

  5. Joint European contribution to phase 5 of the BN600 hybrid reactor benchmark core analysis (European ERANOS formulaire for fast reactor core analysis)

    International Nuclear Information System (INIS)

    Rimpault, G.

    2004-01-01

    Hybrid UOX/MOX fueled core of the BN-600 reactor was endorsed as an international benchmark. BFS-2 critical facility was designed for full size simulation of core and shielding of large fast reactors (up tp 3000 MWe). Wide experimental programme including measurements of criticality, fission rates, rod worths, and SVRE was established. Four BFS-62 critical assemblies have been designed to study changes in BN-600 reactor physics-when moving to a hybrid MOX core. BFS-62-3A assembly is a full scale model of the BN-600 reactor hybrid core. it consists of three regions of UO 2 fuel, axial and radial fertile blankets, MOX fuel added in a ring between MC and OC zones, 120 deg sector of stainless steel reflector included within radial blanket. Joint European contribution to the Phase 5 benchmark analysis was performed by Serco Assurance Winfrith (UK) and CEA Cadarache (France). Analysis was carried out using Version 1.2 of the ERANOS code; and data system for advanced and fast reactor core applications. Nuclear data is based on the JEF2.2 nuclear data evaluation (including sodium). Results for Phase 5 of the BN-600 benchmark have been determined for criticality and SVRE in both diffusion and transport theory. Full details of the results are presented in a paper posted on the IAEA Business Collaborator website nad a brief summary is provided in this paper

  6. The functionalities of the Darwin radioactivity calculation form and the radiation protection studies; Les fonctionnalites du formulaire de calcul de la radioactivite Darwin et les etudes de radioprotection

    Energy Technology Data Exchange (ETDEWEB)

    Tsilanizara, A; Huynh, T D; Luneville, L; Diop, C M; Eid, M [CEA Saclay, Dir. de l' Energie Nucleaire (DEN/DM2S/SERMA), Service d' Etudes des reacteurs et de Modelisation Avancee, 91 - Gif-sur-Yvette (France)

    2003-07-01

    The characterisation of the radioactive sources relative to the evolution of nuclear fuels or to the activation under particles flux (generally neutrons) of structures of a nuclear equipment or a simple isotope decay is a step in the radiation protection studies. This characterisation needs to know a fundamental knowledge: the radionuclides concentration. This one changes with time, and follows the coupled differential equations of first order in time, the generalised Bateman equations. The objective of this paper is to present the functionalities of the Darwin form, developed by the Cea and dedicated to the study of radioactivity. (N.C.)

  7. Reactor protection system

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Lesniak, L.M.; Orgera, E.G.

    1977-10-01

    The report describes the reactor protection system (RPS-II) designed for use on Babcock and Wilcox 145-, later 177-, and 205-fuel assembly pressurized water reactors. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low-pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, a description of the software programmed in the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W

  8. Reactor safety protection system

    International Nuclear Information System (INIS)

    Nishi, Hiroshi; Yokoyama, Tsuguo.

    1989-01-01

    A plurality of neutron detectors are disposed around a reactor core and detection signals from optional two neutron detectors are inputted into a ratio calculation device. If the ratio between both of the neutron flux level signals exceeds a predetermined value, a reactor trip signal is generated from an alarm setting device. Further, detection signals from all of the neutron detection devices are inputted into an average calculation device and the reactor trip signal is generated also in a case where the average value exceeds a predetermined set value. That is, when the reactor core power is increased locally, the detection signal from the neutron detector nearer to the point of power increase is greater than the increase rate for the entire reactor core power, while the detection signal from the neutron detector remote from the point of power increase is smaller. Thus, the local power increase ratio in the FBR reactor core can be detected efficiently by calculating the ratio for the neutron flux level signals from two neutron detectors, thereby enabling to exactly recognize the local power increase rate in the reactor core. (N.H.)

  9. Reactor protecting device

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  10. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

  11. Qualification of the Darwin code for the studies of the fuel cycle relative to the boiling water reactors; Qualification du formulaire Darwin pour les etudes du cycle du combustible pour les reacteurs a eau bouillante

    Energy Technology Data Exchange (ETDEWEB)

    Allais, V

    1998-03-01

    This thesis was carried out in the framework of fuel cycles studies in partnership with COGEMA; the aim is to determine physics parameters characterising Boiling Reactor Assemblies. Those reactors Firstly distinguish themselves from Pressurised Water Reactor by the boiling of the moderator in the core and secondary by the strong neutronics heterogeneity due to complex design. The diphasic mixture formed is characterised by the void fraction parameter. The loss of information, and neutronic studies characteristics of Boiling Water Reactors led us to make preliminary studies having in view to quantify the void fraction impact on the isotopics evolution. Studies on neutronics influence of assemblies and control rods from the immediate environment allows to define the cluster size to describe. The radial description optimisation with APOLLO-2 is necessary to improve the calculation performance and to reduce the errors coming from the modelization. The following points were studied: pellet radial discretization, clustering of cells characterized by a similar behaviour, options in flux spatial calculation (interface current formalism), self-shielding optimisation (specific to each isotopes). The three dimensional modelization with CRONOS-2 and the simplified accounting of the thermohydraulics / neutronics coupling done by a procedure developed and written during this thesis, allow an evaluation of axial distribution of void fraction, power and burn-up during the irradiation. The comparison with experimental analytic results of complete assembly and pin samples dissolutions allows the qualification of this procedure and confirms the necessity to take into account the void fraction axial variation during the evolution. The application of an automatic coupling with the DARWIN cycle code will allow a precise burnup calculation to be utilized in an industrial procedure. (author)

  12. Physical protection of power reactors

    International Nuclear Information System (INIS)

    Darby, J.L.

    1979-01-01

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

  13. Radiation protection in nuclear reactors

    International Nuclear Information System (INIS)

    El-Ashkar, Mohamed

    2008-01-01

    Full text: People are exposed to ionizing radiation in many different forms: cosmic rays that penetrate earth atmosphere or radiation from soil and mineral resources are natural forms of ionizing radiation. Other forms are produced artificially using radioactive materials for various beneficial applications in medicine, industry and other fields. The greatest concerns about ionizing radiation are tied to its potential health effects and a system of radiation protection has been developed to protect people from harmful radiation. The promotion of radiation protection is one of the International Atomic Energy Agency main activities. Radiation protection concerns the protection of workers, members of public, and patients undergoing diagnosis and therapy against the harmful effects of ionizing radiation. The report covers the responsibility of radiation protection officer in Egypt Second Research Reactor (ETRR-2) in Inshas - Egypt, also presents the protection against ionizing radiation from external sources, including types of radiation, sources of radiation (natural - artificial), and measuring units of dose equivalent rate. Also covers the biological effects of ionizing radiation, personal monitoring and radiation survey instruments and safe transport of radioactive materials. The report describes the Egypt Second Research Reactor (ETRR-2), the survey instruments used, also presents the results obtained and gave a relations between different categories of data. (author)

  14. Reactor protection and shut-down system

    International Nuclear Information System (INIS)

    Klar

    1980-01-01

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

  15. Radiation protection at new reactors

    International Nuclear Information System (INIS)

    Brissaud, A.

    2000-01-01

    The theoretical knowledge and the feedback of operating experience concerning radiations in reactors is now considerable. It is available to the designer in the form of predictive softwares and data bases. Thus, it is possible to include the radiation protection component throughout all the design process. In France, the existing reactors have not been designed with quantified radiation protection targets, although considerable efforts have been made to reduce sources of radiation illustrated by the decrease of the average dose rates (typically a factor 5 between the first 900 MWe and the last 1300 MWe units). The EDF ALARA PROJECT has demonstrated that good practises, radiation protection awareness, careful work organization had a strong impact on operation and maintenance work volume. A decrease of the average collective dose by a factor 2 has been achieved without noticeable modifications of the units. In the case of new nuclear facilities projects (reactor, intermediate storage facility,...), or special operations (such as steam generator replacement), quantified radiation protection targets are included in terms of collective and average individual doses within the frame of a general optimization scheme. The target values by themselves are less important than the application of an optimization process throughout the design. This is because the optimization process requires to address all the components of the dose, particularly the work volume for operation and maintenance. A careful study of this parameter contributes to the economy of the project (suppression of unecessary tasks, time-saving ergonomy of work sites). This optimization process is currently applied to the design of the EPR. General radiation protection provisions have been addressed during the basic design phase by applying general rules aiming at the reduction of sources and dose rates. The basic design optimization phase has mainly dealt with the possibility to access the containment at full

  16. Reactor protection systems for the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Morris, C.R.

    2003-01-01

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

  17. Additional reactor protection system of RBMK-1500

    International Nuclear Information System (INIS)

    1999-01-01

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

  18. Radiation protection at reactors RA and RB

    International Nuclear Information System (INIS)

    Ninkovic, M.

    2003-02-01

    Radiation protection activities at the RA and RB reactors are imposed by the existing legal regulations and international recommendations in this field. This annual report contains five parts which cover the following topics: Radiation safety, dosimetry control and technical radiation protection at reactors RA and RB; Handling of radioactive waste, actions and decontamination; Control of the environment (surroundings of RA and RB reactors) and meteorological measurements; Control of internal contamination and internal exposure; Health control od personnel exposed to radiation. Personnel as well as financial data are part of this report

  19. Nuclear reactor safety protection device

    International Nuclear Information System (INIS)

    Okido, Fumiyasu; Noguchi, Atomi; Matsumiya, Shoichi; Furusato, Ken-ichiro; Arita, Setsuo.

    1994-01-01

    The device of the present invention extremely reduces a probability of causing unnecessary scram of a nuclear reactor. That is, four control devices receive signals from each of four sensors and output four trip signals respectively in a quardruplicated control device. Each of the trip signals and each of trip signals via a delay circuit are inputted to a logical sum element. The output of the logical sum circuit is inputted to a decision of majority circuit. The decision of majority circuit controls a scram pilot valve which conducts scram of the reactor by way of a solenoid coils. With such procedures, even if surge noises of a short pulse width are mixed to the sensor signals and short trip signals are outputted, there is no worry that the scram pilot valve is actuated. Accordingly, factors of lowering nuclear plant operation efficiency due to erroneous reactor scram can be reduced. (I.S.)

  20. Radiological protection in nucleus reactor; Perlindungan radiologi di reaktor nukleus

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The chapter briefly discussed the following subjects: radiological protection problems of reactor 1. in operation 2. types of reactor i.e. power reactors, research reactors, etc. 3. during maintenance and installation of fuels. 4. nuclear fuels.

  1. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

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

  2. Radiation protection in a university TRIGA reactor

    International Nuclear Information System (INIS)

    Tschurlovits, M. . Author

    2004-01-01

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

  3. Reactor protection systems of 500 MWe PHWRs

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

  4. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  5. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

    Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

    1996-01-01

    The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

  6. Safety Analysis for Power Reactor Protection System

    International Nuclear Information System (INIS)

    Eisawy, E.A.; Sallam, H.

    2012-01-01

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

  7. Radiation protection personnel training in Research Reactors

    International Nuclear Information System (INIS)

    Fernandez, Carlos Dario; Lorenzo, Nestor Pedro de

    1996-01-01

    The RA-6 research reactor is considering the main laboratory in the training of different groups related with radiological protection. The methodology applied to several courses over 15 years of experience is shown in this work. The reactor is also involved in the construction, design, start-up and sell of different installation outside Argentina for this reason several theoretical and practical courses had been developed. The acquired experience obtained is shown in this paper and the main purpose is to show the requirements to be taken into account for every group (subjects, goals, on-job training, etc) (author)

  8. Validation of reactor core protection system

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  9. Device for protecting deformations of reactor cores

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Urushihara, Hiroshi.

    1975-01-01

    Object: To provide a fluid pressure cylinder, which is operated according to change in temperature of coolant for a reactor to restrain or release a core, to simply and effectively protect deformation of the core. Structure: A closed fluid pressure cylinder interiorly filled with suitable fluid is disposed in peripherally equally spaced relation in an annular space between a core barrel of a reactor and a reactor vessel. A piston is mounted in fluid-tight fashion in a plurality of piston openings made in the cylinder, the piston being slidably moved according to expansion and contraction of the fluid filled in the cylinder. The piston has a movable frame mounted at the foremost end thereof, the movable frame being moved integral with the piston, and the surface opposite the mount thereof biasing the outermost peripheral surface of the core. (Kamimura, M.)

  10. Fault-tolerant reactor protection system

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1997-01-01

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

  11. Establishing a Radiation Protection Programme for a Research Reactor

    International Nuclear Information System (INIS)

    Abdallah, M. M.

    2014-04-01

    The nature and intensity of radiation from the operation of a research reactor depend on the type of reactor, its design features and its operational history. The protection of workers from the harmful effect of radiation must therefore be of paramount importance to any operating organization of a research reactor. This project report attempts to establish an operational radiation protection programme for a research reactor using the Ghana Research Reactor-1 as a case study. (au)

  12. New technology for reactor protection system of CAREM reactor

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  13. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  14. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

    In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

  15. Reactor protection system refurbishment at Paks

    International Nuclear Information System (INIS)

    Hetzmann, A.; Turi, T.

    1997-01-01

    The history and the milestones of the reactor protection system refurbishment are outlined. During the preparation phase of the refurbishment project, detailed requirements have been set up and specific technical solutions developed. The structure of the project documents prepared during these activities is shown in a figure. The life cycle of the project was divided into four phases: the preparatory phase; the design and manufacturing phase; the installation and commissioning phase; and the operation phase. For all four Paks units a time schedule for implementation was set up. The licensing process is dealt with; the principal license was issued in June 1996. (A.K.)

  16. Radiation protection issues for EPR reactor

    International Nuclear Information System (INIS)

    Miniere, D.; Le Guen, B.; Beneteau, Y.; Le Guen, B.

    2008-01-01

    As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection Radiation Protection aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing Radiation Protection-significant activities has been to include all involved stakeholders - designers, licensee and contractor companies - in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The Radiation Protection target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 Man Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the E.P.R.'s reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), Radiation Protection studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the

  17. ISAT promises fail-safe computer-based reactor protection

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    AEA Technology's ISAT system is a multiplexed microprocessor-based reactor protection system which has very extensive self-monitoring capabilities and is inherently fail safe. It provides a way of addressing software reliability problems that have tended to hamper widespread introduction of computer-based reactor protection. (author)

  18. Protective actions as a factor in power reactor siting

    Energy Technology Data Exchange (ETDEWEB)

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables.

  19. Protective actions as a factor in power reactor siting

    International Nuclear Information System (INIS)

    Gant, K.S.; Schweitzer, M.

    1984-06-01

    This report examines the relationship between a power reactor site and the ease of implementing protective actions (emergency measures a serious accident). Limiting populating density around a reactor lowers the number of people at risk but cannot assure that all protective actions are possible for those who reside near the reactor. While some protective measures can always be taken (i.e., expedient respiratory protection, sheltering) the ability to evacuate the area or find adequate shelter may depend on the characteristics of the area near the reactor site. Generic siting restrictions designed to identify and eliminate these site-specific constraints would be difficult to formulate. The authors suggest identifying possible impediments to protective actions at a proposed reactor site and addressing these problems in the emergency plans. 66 references, 6 figures, 8 tables

  20. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1976-01-01

    Reactor protection systems for nuclear power plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. The paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

  1. Reactor protection system design using micro-computers

    International Nuclear Information System (INIS)

    Fairbrother, D.B.

    1977-01-01

    Reactor Protection Systems for Nuclear Power Plants have traditionally been built using analog hardware. This hardware works quite well for single parameter trip functions; however, optimum protection against DNBR and KW/ft limits requires more complex trip functions than can easily be handled with analog hardware. For this reason, Babcock and Wilcox has introduced a Reactor Protection System, called the RPS-II, that utilizes a micro-computer to handle the more complex trip functions. This paper describes the design of the RPS-II and the operation of the micro-computer within the Reactor Protection System

  2. Extensions and renovations of reactor protection systems

    International Nuclear Information System (INIS)

    Hellmerichs, K.

    1985-01-01

    Increase of requirements by the authorities as to the design of reactor protection systems affected in the last years not only plans being under construction, but also resulted in partly spacious extensions and renovations. While working on the extensions and renovations a lot of problems arose: far-reaching performance of newest guidelines and rules in spite of old plant concepts; partly higher degree of redundancy requirements of the new systems in contrast to the present systems; use of present safeguard systems for new accident countermeasures; designation of priorities between present and new functions, especially in view of fault behaviour of present systems; adaptation of the new I and C equipment to the present signalisation-, operation- and information-arrangements under consideration of the present operational philosophy; spatial incorporation of new equipments; construction as to time without expanding of the planned refuelling phases. Because the KWU has planned and constructed such alterations in nearly 10 plants a lot of experience has been gathered. (author)

  3. Diversity in computerized reactor protection systems

    International Nuclear Information System (INIS)

    Fischer, H.D.; Piel, L.

    1999-01-01

    Based on engineering judgement, the most important measures to increase the independency of redundant trains of a computerized safety instrumentation and control system (I and C) in a nuclear power plant are evaluated with respect to practical applications. This paper will contribute to an objective discussion on the necessary and justifiable arrangement of diversity in a computerized safety I and C system. Important conclusions are: - (i) diverse equipment may be used to control dependent failures only if measures necessary for designing, licensing, and operating a computerized safety I and C system homogeneous in equipment are neither technically nor economically feasible; - (ii) the considerable large operating experience in France with a non-diverse equipment digital reactor protection system does not call for equipment diversity. Although there are no generally accepted methods, the licensing authority is still required to take into account dependent failures in a probabilistic safety analysis; - (ii) the frequency of postulated initiating events implies which I and C functionality should be implemented on diverse equipment. Using non-safety I and C equipment in addition to safety I and C equipment is attractive because its necessary unavailability to control an initiating event in teamwork with the safety I and C equipment is estimated to range from 0.01 to 0.1. This can be achieved by operational experience

  4. RA Research nuclear reactor, Part II: radiation protection at the RA reactor in 1987

    International Nuclear Information System (INIS)

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

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  6. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  8. Improving 200 MW NDHR reactor protection system with GAL devices

    International Nuclear Information System (INIS)

    Shi Mingde; Li Duo; Xie Zhengguo

    1999-01-01

    The emergence of General Array Logic (GAL), a fairly new type of logic devices with the characteristics of user-definable logic functions, have led to a revolutionary change in the design of logical circuits. The improvements of the reactor protection system for the 200 MW nuclear district heating reactor (NDHR) using GAL are covered

  9. Automated reactor protection testing saves time and avoids errors

    International Nuclear Information System (INIS)

    Raimondo, E.

    1990-01-01

    When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

  10. Automated testing of reactor protection instrumentation made easy

    International Nuclear Information System (INIS)

    Iborra, A.; De Marcos, F.; Pastor, J.A.; Alvarez, B.; Jimenez, A.; Mesa, E.; Alsonso, L.; Regidor, J.J.

    1997-01-01

    Maintenance and testing of reactor protection systems is an important cause of unplanned reactor trips. Automated testing is the answer because it minimises test times and reduces human error. The GAMA I system, developed and implemented at Vandellos II in Spain, has the added advantage that it uses visual programming, which means that changing the software does not need specialist programming skills. (author)

  11. Protection of semiconductor converters for controlled bypass reactors

    International Nuclear Information System (INIS)

    Dolgopolov, A. G.; Akhmetzhanov, N. G.; Karmanov, V. F.

    2010-01-01

    Possible ways of protecting thyristor converters in systems for magnetizing 110 - 500 kV controlled bypass reactors during switching and automatic reclosing are examined based on experience with the development of equipment, line tests, and mathematical modelling.

  12. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. The purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing. A demonstration model for protection system of PWR reactor has been designed and built

  13. Model Based Cyber Security Analysis for Research Reactor Protection System

    International Nuclear Information System (INIS)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung; Son, Hanseong

    2013-01-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN

  14. Nuclear Reactor RA Safety Report, Vol. 9, Radiation protection

    International Nuclear Information System (INIS)

    1986-11-01

    Instrumentation for Radiation protection existing at the RA reactor is dating mostly from the period 1957-1959 when the reactor has been built. With some minor exception it was produced in USSR. Radiation protection system was constructed based on specific design project, somewhat modified original USSR project which has been indispensable because of some modification of the building design. During the past 27 years no renewal of the instrumentation was done, only maintenance was performed. Instrumentation consists of old electronic devices which caused difficulties and even prevented regular maintenance because of lack of spare parts. Instrumentation for radiation protection at the RA reactor is classified as follows: centralized dosimetry system; stationary dosimetry instrumentation, movable and personal dosimetry systems. Apart from the scheme of dosimetry instrumentation this volume includes description of radiation protection procedures; protection devices; radiation doses and dose limit data; program for environmental radioactivity control; medical control procedures [sr

  15. Operation monitoring and protection method for nuclear reactor

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi.

    1995-01-01

    In an operation and monitoring method for a PWR-type reactor by using a tetra-sected neutron detector, axial off set is defined by neutron detector signals with respect to an average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core. A departure from nucleate boiling (DNBR) is represented by standardized signals, and the DNBR is calculated by using the axial off set of the average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core, and they are graphically displayed. In addition, a thermal flow rate-water channel coefficient is also graphically displayed, and the DNBR and the thermal flow rate-water channel coefficient are restricted based on the display, to determine an allowable operation range. As a result, it is possible to provide an operation monitoring and protection method for nuclear reactor capable of reducing labors and frequencies for the change of protection system setting in a case of using a tetra-sected neutron detector disposed at the outside and, at the same time, protecting each of DNR and the highest linear power or the thermal water coefficient channel. (N.H.)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  17. Hybrid Plasma Reactor/Filter for Transportable Collective Protection Systems

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-04-06

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

  18. Radiation protection programme for LEU miniature source reactor

    International Nuclear Information System (INIS)

    Beinpuo, Ernest Sanyare Warmann

    2015-02-01

    A radiation protection program has been developed to promote radiation dose reduction. It emphasize radiological protection fundamentals geared at reducing radiation from the application of the research reactor at the reactor center of the National Nuclear Research Institute (NNRI) of the Ghana Atomic Energy Commission. The objectives of the radiation safety program are both to ensure that nuclear scientists and technicians are exposed to a minimum of ionizing radiation and to protect employees and facility users and surrounding community from any potentially harmful effects of nuclear research reactor at GAEC. The primary purpose of the radiation control program is to assure radiological safety of all personnel and the public to guarantee that ionizing radiation arising out of the operations of the Research Reactor at the Reactor Center does not adversely affect personnel, the general public or the environment. This program sets forth polices, regulations, and procedures approved by the Centers Radiation Control Committee. The regulations and procedures outlined in this program are intended to protect all individuals with a minimum of interference in their activities and are consistent with regulations of the Radiation Protection Board (RPB) applicable to ionizing radioactive producing devices. (au)

  19. Radiation protection at the RA Reactor in 1993, RA research reactor, Part

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Sipka, V.; Grsic, Z.

    1993-01-01

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

  20. Reliability analysis of reactor protection systems

    International Nuclear Information System (INIS)

    Alsan, S.

    1976-07-01

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

  1. Design and testing of integrated circuits for reactor protection channels

    International Nuclear Information System (INIS)

    Battle, R.E.; Vandermolen, R.I.; Jagadish, U.; Swail, B.K.; Naser, J.; Rana, I.

    1995-01-01

    Custom and semicustom application-specific integrated circuit design and testing methods are investigated for use in research and commercial nuclear reactor safety systems. The Electric Power Research Institute and Oak Ridge National Laboratory are working together through a cooperative research and development agreement to apply modern technology to a nuclear reactor protection system. Purpose of this project is to demonstrate to the nuclear industry an alternative approach for new or upgrade reactor protection and safety system signal processing and voting logic. Motivation for this project stems from (1) the difficulty of proving that software-based protection systems are adequately reliable, (2) the obsolescence of the original equipment, and (3) the improved performance of digital processing

  2. Reactor protection system with automatic self-testing and diagnostic

    International Nuclear Information System (INIS)

    Gaubatz, D.C.

    1996-01-01

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ''identical'' values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs

  3. Model Based Cyber Security Analysis for Research Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

    2013-07-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

  4. Software reliability and safety in nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor

  5. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  6. The use of process computers in reactor protection systems

    International Nuclear Information System (INIS)

    1973-04-01

    The report contains the papers presented at the LRA information meeting in spring 1972, concerning the use of process computers in reactor protection systems. The main interest was directed at a system conception as proposed from AEG for future BWR-plants. (orig.) [de

  7. Twenty years of Radiology in RP-10 nuclear reactor protection

    International Nuclear Information System (INIS)

    Zapata, Alejandro L.; Ramos, Fernando T.; Arrieta, Rolando W.B.; Vela Mora, Mariano

    2013-01-01

    In this report we present the results about radiation controls during 1990 - 2010, carried out in the Nuclear Reactor RP-10 of the Nuclear Center of Huarangal. These controls and radiological evaluation are of much utility for the radio personnel protection of this one and other reactors, since it allows to compares these variables with respect to the time. From the results obtained in monitoring and radiation controls, we conclude that in no case it has been reached the limits allowed by the Peruvian Regulating Authority. (author)

  8. Reliability analysis of the reactor protection system with fault diagnosis

    International Nuclear Information System (INIS)

    Lee, D.Y.; Han, J.B.; Lyou, J.

    2004-01-01

    The main function of a reactor protection system (RPS) is to maintain the reactor core integrity and reactor coolant system pressure boundary. The RPS consists of the 2-out-of-m redundant architecture to assure a reliable operation. The system reliability of the RPS is a very important factor for the probability safety assessment (PSA) evaluation in the nuclear field. To evaluate the system failure rate of the k-out-of-m redundant system is not so easy with the deterministic method. In this paper, the reliability analysis method using the binomial process is suggested to calculate the failure rate of the RPS system with a fault diagnosis function. The suggested method is compared with the result of the Markov process to verify the validation of the suggested method, and applied to the several kinds of RPS architectures for a comparative evaluation of the reliability. (orig.)

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

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

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

  10. Qualification issues for advanced light-water reactor protection systems

    International Nuclear Information System (INIS)

    Korsah, K.; Clark, R.L.; Antonescu, C.

    1993-01-01

    The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

  11. 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

  12. Reactor protection system including engineered features actuation system

    International Nuclear Information System (INIS)

    Palmaers, W.

    1982-01-01

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

  13. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

    2008-01-01

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  14. Reliability of redundant structures of nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Vojnovic, B.

    1983-01-01

    In this paper, reliability of various redundant structures of PWR protection systems has been analysed. Structures of reactor tip systems as well as the systems for activation of safety devices have been presented. In all those systems redundancy is achieved by means of so called majority voting logic ('r out of n' structures). Different redundant devices have been compared, concerning probability of occurrence of safe as well as unsafe failures. (author)

  15. Study of Physical Protection System at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Ina, I.; Zarina Masood

    2016-01-01

    Physical protection program at PUSPATI TRIGA Reactor (RTP) which is located at Nuklear Malaysia, Bangi Complex has been strengthened and upgraded from time to time to accommodate current situation needs. However, there is always room for improvement. Hence, study have been made to look deeper into physical protection components such as delay systems, external sensors, PPS intruder alarm sensors, use of video system, personnel security or insider threats, access control operation system operation rules and security culture that may need to take into consideration. (author)

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

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

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

  17. System for step-wise accident protection of nuclear reactors

    International Nuclear Information System (INIS)

    Rubek, J.; Kuklik, B.; Bednarik, K.

    1991-01-01

    A system comprising electric switching circuits is proposed for the control of a WWER type reactor shutdown in case of turbine failure or another abnormal situation. The fastest reactor shutdown mode is only resorted to if the pressures in the primary and secondary circuits would otherwise increase above tolerable limits and safety valves would open. The temperature and pressure stress of the nuclear power plant components and fuel is reduced. In this manner, the losses emerging during turbine failures due to false alarms are minimized. The contacts of the system switch if the turbines are relieved to the power of the unit home consumption, if the first or second turbine fails by closing the quick-acting valves, if a signal for blocking the by-pass stations of the operated turbines appears, or if the electric supply of the control system and of the turbo-set protection fails. (M.D.). 1 fig

  18. 10 CFR 73.60 - Additional requirements for physical protection at nonpower reactors.

    Science.gov (United States)

    2010-01-01

    ... nonpower reactors. 73.60 Section 73.60 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION... requirements for physical protection at nonpower reactors. Each nonpower reactor licensee who, pursuant to the... perform their duties. (6) Prior to entry into a material access area, packages shall be searched for...

  19. Development of the signalling circuits for reactor emergency protection systems

    International Nuclear Information System (INIS)

    Volkov, A.V.; Nikiforov, B.N.; Ogon'kov, A.I.; Sychinskij, Yu.L.

    1978-01-01

    Construction of circuits for nuclear reactor emergency protection according to the power level and rate of power rise with the use of integrated microcircuits is discussed. Circuits of relay- and transformer-based logical signaling devices are presented. It is noted that disadvantages of a transformer-based loaical sianaling device are great power consumption (about 300 mW) and slow response limited by the time constant of the output smoothing filter. Further development of circuits under consideration is associated with the employment of new optronic elements intended to replace the transformers

  20. Radiation protection monitoring at the JOYO experimental fast reactor

    International Nuclear Information System (INIS)

    Ouchi, S.; Endo, K.; Susaki, T.

    1979-01-01

    This paper describes the radiation protection monitoring programme for the JOYO experimental fast reactor and some of the health physics problems experienced during the low-power nuclear tests. These include: a detailed description of the centralized radiation monitoring system; the methods and results of the individual monitoring systems; the results of operational monitoring for the handling of new plutonium fuel subassemblies; the evaluation of the external radiation dose rate around the primary coolant system; and the results of an experiment on the thermal dependence of some personnel dose meters. (author)

  1. Preliminary Validation and Verification Plan for CAREM Reactor Protection System

    International Nuclear Information System (INIS)

    Fittipaldi, Ana; Maciel Felix

    2000-01-01

    The purpose of this paper, is to present a preliminary validation and verification plan for a particular architecture proposed for the CAREM reactor protection system with software modules (computer based system).These software modules can be either own design systems or systems based in commercial modules such as programmable logic controllers (PLC) redundant of last generation.During this study, it was seen that this plan can also be used as a validation and verification plan of commercial products (COTS, commercial off the shelf) and/or smart transmitters.The software life cycle proposed and its features are presented, and also the advantages of the preliminary validation and verification plan

  2. Considerations for tritium protection at a fusion reactor

    International Nuclear Information System (INIS)

    Easterly, C.E.

    1981-01-01

    The purpose of this paper is to indicate the general direction of current fusion reactor concepts regarding tritium, and to indicate some options in tritium control strategies. Certain strategies, in addition to providing reduced potential health hazard, afford the potential for engineering alternatives for in-plant tritium control systems. The overall coupling of containment cleanup systems and health protection must continue to develop with increased knowledge of the health effects of different tritium species and the consequent systems options available subsequent to this understanding

  3. Failsafe design criteria for computer based reactor protection systems

    International Nuclear Information System (INIS)

    Keats, A.B.

    1980-01-01

    The design criteria proposed are an extrapolation of the failsafe mode of operation used in the UK in hardwired reactor protection systems. This is achieved by making the operational condition of the reactor dependent upon an energetic state of the protection system components. An important objective of the proposed design criteria is to eliminate, or at least to minimize, the need for a failure-mode-and-effect-analysis (FMEA) of the computer software. This demands some well defined but simple constraints upon the way in which data are stored in the computers, but the objective is achieved almost entirely by hardware properties of the system. The first of these is the systematic use of hardwired test inputs which cause transient excursions into the tripped state in a uniquely ordered but easily recognizable sequence. The second is the use of hardwired pattern recognition logic which generates a dynamic healthy stimulus for the shutdown actuators only in response to the unique sequence formed by the hardwired input signal pattern. It therefore detects abnormal states of any of the system inputs, software errors, wiring errors and hardware failures. This hardwired logic is conceptually simple, failsafe, and is amenable to simple FMEA. (U.K.)

  4. Trip setpoint analysis for the reactor protection system of an advanced integral reactor

    International Nuclear Information System (INIS)

    Yang, Soo Hyung; Kim, Soo Hyung; Chung, Young Jong; Zee, Sung Quun

    2007-01-01

    The trip setpoints for the reactor protection system of a 65-MWt advanced integral reactor have been analyzed through sensitivity evaluations by using the Transients and Setpoint Simulation/System-integrated Modular Reactor code. In the analysis, an inadvertent control rod withdrawal event has been considered as an initiating event because this event results in the worst consequences from the viewpoint of the minimum critical heat flux ratio and its consequences are considerably affected by the trip setpoints. Sensitivity evaluations have been performed by changing the trip setpoints for the ceiling of a variable overpower trip (VOPT) function and the pressure of a high pressurizer pressure trip function. Analysis results show that a VOPT function is an effective means to satisfy the acceptance criteria as the control rod rapidly withdraws: on the other hand, a high pressurizer pressure trip function is an essential measure to preserve the safety margin in the case of a slow withdrawal of the control rod because a reactor trip by a VOPT function does not occur in this case. It is also shown that the adoptions of 122.2% of the rated core power and 16.25 MPa as the trip setpoint for the ceiling of a VOPT function and the pressure of a high pressurizer pressure trip function are good selections to satisfy the acceptance criteria

  5. analysis and implementation of reactor protection system circuits - case study Egypt's 2 nd research reactor-

    International Nuclear Information System (INIS)

    Elnokity, O.E.M.

    2006-01-01

    this work presents a way to design and implement the trip unit of a reactor protection system (RPS) using a field programmable gate arrays (FPGA). instead of the traditional embedded microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the trip unit (TU), which is used in Egypt's 2 nd research reactor ETRR-2. the existing embedded system is built around the STD32 field computer bus which is used in industrial and process control applications. it is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. therefore, the same bus is still used in the proposed design. the state machine of this bus is designed based around its timing diagrams and implemented in VHDL to interface the designed TU circuit

  6. Development of digital power measuring and protecting equipment for SPRR-300 reactor

    International Nuclear Information System (INIS)

    Wang Xuejie; Li Xi'an; Zhu Shilei

    2005-01-01

    A measuring and protecting equipment of reactor power based on Single-Chip Microcomputer is introduced in this paper. The composition of hardware and the major control idea about the software for the equipment are presented. Digitizing the measuring data from nuclear instruments is precondition of reactor control and protection system which would be computerizing, and it is also an application of redundancy and variety of reactor protection system in nuclear measuring instruments. At last the working state is described. (authors)

  7. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures

    International Nuclear Information System (INIS)

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents [sr

  8. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose Luis

    1996-01-01

    Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

  9. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose L.

    2000-01-01

    Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

  10. Using a research reactor to teach practical radiation protection

    International Nuclear Information System (INIS)

    Musilek, A.; Steinhauser, G.

    2010-01-01

    To teach students about the practical handling of radioactive materials and the related radiation protection, it is advantageous to be able to produce radioactive material with specific properties. Through the neutron activation of specific samples, radio-nuclides can be produced that are precisely tailored for particular experiments, both in type of radiation (beta, gamma) as well as in activity and half-life. At the Atominstitut in Vienna, a 250 kW TRIGA Mark II research reactor is used for the production of these nuclides. In this paper, four practical exercises are presented, covering many questions and challenges that occur in radiation protection. The first exercise uses neutron activation of sodium-chloride to cover theoretical aspects of the calculation of dose rates (using dose rate constants) through the activation of Na-23, Cl-35 and Cl-37 (including cross sections, half-life, inverse square law), as well as a practical examination (handling of dose rate meters). The second exercise gives students the opportunity to decontaminate a laboratory after an incident under realistic circumstances. For this exercise, KNO 3 is activated in the reactor. The resulting K-nuclide produces no risk of inadequate decontamination for the laboratory, since the half-life of K-42 is only 12 h. The third exercise is designed to teach students how to deal with unsealed radioactive material by irradiation of ammonium dihydrogenphosphate. In this case, an only-beta-active (P-32) fertilizer is produced, which is applied to plants in subsequent chemical processing. In the following step, the 'quality of this fertilizer' is determined by measuring the absorbed activity of the plant leaves using a GM counter. The fourth exercise is another approach in working with unsealed radioactive material. It simulates the PUREX process to separate uranium from fission products using a liquid - liquid extraction. (authors)

  11. 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

  12. Design of first reactor protection system prototype for C A R E M reactor

    International Nuclear Information System (INIS)

    Azcona, A; Lorenzo, G.; Maciel, F.; Fittipaldi, A

    2006-01-01

    In this paper we present the design of a prototype of the C A R E M Reactor Protection System, which is implemented on a basis of the digital platform T E L E P E R M X S.The proposed architecture for the Reactor Protection System (R P S) has 4 redundant trains composed by a complete set of sensors, a data acquisition computer and a processing computer.The information from the 4 processing computers goes into to a two voting units with a two out of four (2004) logic and its outputs are combined by a final actuation logic with a voting scheme of one out of two (1002).The prototype is implemented with a unique train.The train inputs are simulated by an Automatic Testing Unit.The pre-established test case or procedure results are fed back into the A T U.The choice of the digital platform T E L E P E R M X S for the R P S implementation allows versatility in the design stage and permits the prototype expansion due to its modular characteristic and the software tools flexibility [es

  13. Off-site protective action selection for nuclear reactor accidents

    International Nuclear Information System (INIS)

    Weerakkody, S.D.

    1986-01-01

    A computer program based upon a model using a rational theoretical basis was developed to select appropriate off-site protective actions during nuclear reactor accidents. The special features of this program include (a) introduction of a precursor concept that uses the history of the accident progression to determine the spectrum of potential accident scenarios and estimates of the likelihoods of each accident scenario; (b) use of statistical decision theory and the concept of entropy of a spectrum to select the appropriate protective actions using either the minimax principle or the Bayes action method; and (c) introduction of methods to quantify evacuation travel risks. In order to illustrate the usefulness of the computer program, it was applied at three stages of the Three Mile Island accident scenario. Quantified non-radiological risks of evaluation have been used to establish dose thresholds below which evacuations are not justified. Using the Poisson analysis for evacuation risks and the absolute L-L BEIR model for radiation risk suggests 330 mrems as a reasonable value for this threshold. The usefulness of the program in developing a technical basis to select the size of the plume exposure pathway emergency planning zone (EPZ) is discussed. Quantified evacuation risks, cost, and the current rationale upon which the EPZ is based, justify an EPZ between 5-10 miles for WASH-1400 source-terms

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

    International Nuclear Information System (INIS)

    Teguh-Sulistyo

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  16. Reactor protection system design using application specific integrated circuits

    International Nuclear Information System (INIS)

    Battle, R.E.; Bryan, W.L.; Kisner, R.A.; Wilson, T.L. Jr.

    1992-01-01

    Implementing reactor protection systems (RPS) or other engineering safeguard systems with application specific integrated circuits (ASICs) offers significant advantages over conventional analog or software based RPSs. Conventional analog RPSs suffer from setpoints drifts and large numbers of discrete analog electronics, hardware logic, and relays which reduce reliability because of the large number of potential failures of components or interconnections. To resolve problems associated with conventional discrete RPSs and proposed software based RPS systems, a hybrid analog and digital RPS system implemented with custom ASICs is proposed. The actual design of the ASIC RPS resembles a software based RPS but the programmable software portion of each channel is implemented in a fixed digital logic design including any input variable computations. Set point drifts are zero as in proposed software systems, but the verification and validation of the computations is made easier since the computational logic an be exhaustively tested. The functionality is assured fixed because there can be no future changes to the ASIC without redesign and fabrication. Subtle error conditions caused by out of order evaluation or time dependent evaluation of system variables against protection criteria are eliminated by implementing all evaluation computations in parallel for simultaneous results. On- chip redundancy within each RPS channel and continuous self-testing of all channels provided enhanced assurance that a particular channel is available and faults are identified as soon as possible for corrective actions. The use of highly integrated ASICs to implement channel electronics rather than the use of discrete electronics greatly reduces the total number of components and interconnections in the RPS to further increase system reliability. A prototype ASIC RPS channel design and the design environment used for ASIC RPS systems design is discussed

  17. Safety aspect of digital reactor protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Zen-Ichi

    1998-01-01

    It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

  18. Programming Guidelines for FBD Programs in Reactor Protection System Software

    International Nuclear Information System (INIS)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom; Lee, Jang Su

    2014-01-01

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows

  19. Programming Guidelines for FBD Programs in Reactor Protection System Software

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Se Jin; Lee, Dong Ah; Kim, Eui Sub; Yoo, Jun Beom [Division of Computer Science and Engineering College of Information and Communication, Konkuk University, Seoul (Korea, Republic of); Lee, Jang Su [Man-Machine Interface System team Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Properties of programming languages, such as reliability, traceability, etc., play important roles in software development to improve safety. Several researches are proposed guidelines about programming to increase the dependability of software which is developed for safety critical systems. Misra-c is a widely accepted programming guidelines for the C language especially in the sector of vehicle industry. NUREG/CR-6463 helps engineers in nuclear industry develop software in nuclear power plant systems more dependably. FBD (Function Block Diagram), which is one of programming languages defined in IEC 61131-3 standard, is often used for software development of PLC (programmable logic controllers) in nuclear power plants. Software development for critical systems using FBD needs strict guidelines, because FBD is a general language and has easily mistakable elements. There are researches about guidelines for IEC 61131-3 programming languages. They, however, do not specify details about how to use languages. This paper proposes new guidelines for the FBD based on NUREG/CR-6463. The paper introduces a CASE (Computer-Aided Software Engineering) tool to check FBD programs with the new guidelines and shows availability with a case study using a FBD program in a reactor protection system. The paper is organized as follows.

  20. Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

    International Nuclear Information System (INIS)

    Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

    2012-01-01

    Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

  1. Use of digital computers in the protection system for Savannah River reactors

    International Nuclear Information System (INIS)

    Gimmy, K.L.

    1977-06-01

    Each production reactor at the Savannah River Plant has recently been provided with a protective system using dual digital computers. The dual ''safety computers'' monitor coolant temperature and flow in each of the 600 fuel assemblies in the reactor. The system provides alarms and automatic reactor shutdown (SCRAM) if these variables exceed predetermined setpoints. The system provides the primary protection for unwanted local or general power increase or assembly coolant flow reduction. Standard process control computers are used and all scanning, data output, and protective action are controlled by software prepared by Du Pont

  2. 10 CFR 73.37 - Requirements for physical protection of irradiated reactor fuel in transit.

    Science.gov (United States)

    2010-01-01

    ... fuel in transit. 73.37 Section 73.37 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Physical Protection of Special Nuclear Material in Transit § 73.37 Requirements for physical protection of irradiated reactor fuel in transit. (a) Performance objectives. (1...

  3. Emergency systems and protection equipment of modular steam generators for fast reactors

    International Nuclear Information System (INIS)

    Matal, O.

    The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

  4. Comparison, with regard to safety, between a hard-wired reactor protection system and a computerized protection system. Pt. 1

    International Nuclear Information System (INIS)

    Buettner, W.E.

    1976-07-01

    The study compares a conventional hard-wired dynamic reactor protection system with a computerized protection system. In the comparison, only the unequivocally safety-oriented protection actions are considered. In the first part, the different structures of both systems and the method of verification for their functional safety will be described. In the second part, the mean unavailability in case of demand for both systems under defined conditions will be determined. (orig.) [de

  5. A protection system of low temperature thermo-supply nuclear reactor

    International Nuclear Information System (INIS)

    Jiang Binsen

    1988-09-01

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

  6. 78 FR 50313 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2013-08-19

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Orders; rescission. SUMMARY... the NRC published a final rule, ``Physical Protection of Irradiated Fuel in Transit,'' on May 20, 2013... of Irradiated Reactor Fuel in Transit'' (RIN 3150-AI64; NRC-2009-0163). The final rule incorporates...

  7. French experience in using programmable systems for the control and the protection of nuclear reactors

    International Nuclear Information System (INIS)

    Jover, P.

    1986-01-01

    The paper presents the results obtained in the use of two automated systems important to safety in 1300 MWE French nuclear reactors: the numerical integrated protection system (SPIN) and the logical control system (CONTROBLOC)

  8. A relay rack for a control and protection system for nuclear reactors

    International Nuclear Information System (INIS)

    Miyata, Yasuyuki; Oda, Noriaki; Akiyama, Toyoshi

    1975-01-01

    It is obvious that all the equipment in the various systems that constitute a nuclear power plant must exhibit the highest levels of reliability, but the reactor control and protection system is of vital importance, and thus it requires a particularly thorough approach, incorporating redundancy, independence and separation. The paper describes the functions, construction and specifications of the relay rack - one of the most important items of equipment for reactor control and protection in a generating facility using a pressurized-water reactor - and it gives details of the extent to which these three requirements are satisfied. (author)

  9. Functional safeguards for computers for protection systems for Savannah River reactors

    International Nuclear Information System (INIS)

    Kritz, W.R.

    1977-06-01

    Reactors at the Savannah River Plant have recently been equipped with a ''safety computer'' system. This system utilizes dual digital computers in a primary protection system that monitors individual fuel assembly coolant flow and temperature. The design basis for the (SRP safety) computer systems allowed for eventual failure of any input sensor or any computer component. These systems are routinely used by reactor operators with a minimum of training in computer technology. The hardware configuration and software design therefore contain safeguards so that both hardware and human failures do not cause significant loss of reactor protection. The performance of the system to date is described

  10. Franco-German cooperation for the physical protection of the EPR reactor

    International Nuclear Information System (INIS)

    Jalouneix, J.; Hagemann, A.

    2001-01-01

    This article presents the proceeding that has been followed in the EPR (European pressurized water reactor) project concerning physical protection against malevolent actions and robbery of nuclear materials. Before the different options of the nuclear island were definitely set, a task group had been constituted to examine if these options could hamper the setting of physical protection measures that are required by the legislation of the 2 countries. Another group composed of experts from IPSN/GRS (Institut de Protection et de Surete Nucleaire / Gesellschaft fur Anlagen und Reaktorsicherheit) had the task to define common requirements concerning the physical protection of reactors in Germany and in France. In this framework the EPR project team has prepared a technical document reviewing the different dispositions that have been retained to assure the physical protection of the reactor. (A.C.)

  11. Preventive protection device and method for bottom of reactor pressure vessel

    International Nuclear Information System (INIS)

    Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

    1995-01-01

    In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

  12. Nuclear reactors. Use of the protection system for non-safety purposes (International Electrotechnical Commission Standard Publication 639:1979)

    International Nuclear Information System (INIS)

    Stefanik, J.

    1996-01-01

    This standard applies to the protection system of a nuclear reactor and, more especially, to all interconnections between a reactor protection system (as defined and explained in International Electrotechnical Commission Publication 231 A, first supplement to Publication 231, General Principles of Nuclear Reactor Instrumentation) and all other systems and equipment not part of the protection system, except: a) the physical connection between sensors of the protection system and the physical variables that they monitor, such as for example, thermo wells, moderating medium for neutron sensors, etc.; b) the electrical connection between the protection system and the reactor control rods or other safety mechanism; c) the electrical and pneumatic connections to the power distribution system (mains) and pneumatic supplies that supply power to the protection system. Although many clauses relate to all reactor protection systems, this standard applies mainly to protection systems in nuclear power reactors

  13. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

    International Nuclear Information System (INIS)

    Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  14. Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

    Science.gov (United States)

    Blanchet, David; Fontaine, Bruno

    2017-09-01

    The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

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

    International Nuclear Information System (INIS)

    Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  17. Common cause analysis of the TREAT upgrade reactor protection system

    Energy Technology Data Exchange (ETDEWEB)

    Page, R.J.; Kamis, G.J.; Marbach, R.A.; Mueller, C.J.

    1984-09-01

    A triply redundant reactor scram system (RSS) has been designed for the upgraded TREAT facility. The independent failures reliability goal for the RSS is <10/sup -9/ failures per demand. An independent failures analysis indicated that this goal would be met. In addition, however, recognizing that in heavily redundant systems common-cause failures dominate, a common cause analysis of the TREAT upgrade RSS was done. The objective was to identify those common-cause initiators which could affect the functioning of the RSS, and to subsequently modify the design of the RSS so that the effect was minimized. A number of common-cause initiators were identified which were capable of defeating the triple redundancy feature of the reactor scram system. By means of a systematic analysis of the effect these initiators could have on the system, it was possible to identify seven necessary design and procedural modifications that would greatly reduce the probability of the reactor being run while the RSS was in a faulted condition.

  18. Development of a protection system for research reactor based in Field Programmable Gate Array - FPGA

    International Nuclear Information System (INIS)

    Martins, Roque Hudson da Silva

    2016-01-01

    This study presents a implementation purpose of a protection system for research nuclear reactors by using a programed device FPGA (Field Programmable Gate Array). As well as logic protection method involved on an automatic shutdown (TRIP) of a reactor, that ensure the security on such systems. These new control and operation mechanics are developed to guarantee that the security limits of a power plant are not exceeded, these mechanics can work isolated or in groups to safe guard the security levels. For this implementation to be completed, there will be presented the main aspects and concepts referred to protection systems, mostly about research nuclear reactors, with some applications terms exposed. The system proposed at this paper was developed following the VHDL (Very High Speed Integrated Circuits) hardware describing language, and the Modelsim software from Altera Software to program the automatic turning off routines, and hypothetical simulations for such. The results show that for every software application for supporting nuclear reactors, like security devices, they have to meet the IEC 60880 criteria. This paper have great importance, seeing that nuclear reactor security systems, are a basic element for ensure the reactor security. (author)

  19. A structure for the protection of nuclear-reactor pressurized-vessels against rupture

    International Nuclear Information System (INIS)

    Marcellin, J.-P.; Aubert, Gilles

    1974-01-01

    Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

  20. 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

  1. Research of management information system of radiation protection for low temperature nuclear heating reactor

    International Nuclear Information System (INIS)

    Bai Hongtao; Wang Jiaying; Wu Manxue

    2001-01-01

    Management information system of radiation protection for low temperature reactor uses computer to manage the data of the low temperature nuclear heating reactor radiation monitoring, it saves the data from the front real-time radiation monitoring system, comparing these data with historical data to give the consequence. Also, the system provides some picture in order to show space information at need. The system, based on Microsoft Access 97, consists of nine parts, including radiation dose, environmental data, meteorological data and so on. The system will have value in safely operation of the low temperature nuclear heating reactor

  2. 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

  3. Criteria of the efficiency for radiation protection of tokamak reactor superconducting magnet coils

    International Nuclear Information System (INIS)

    Zimin, S.A.

    1988-01-01

    Factors determining serviceability of the main elements (superconductor, stabilizing conductor, insulation) of superconducting magnet coils for tokamak reactors are discussed. It is suggested that the limiting values of total and specific energy release in the material of superconducting coils, increase in electric resistance of the stabilizing conductor, decrease in the superconductor critical current and damage of the superconducting magnet insulation should be used as criteria of the reactor internal radiation protection efficiency. The conclusion is made that neutron fluence in the magnet coil components considered can be used as a generalized criterion of the first approximation for the evaluation of the protection efficiency

  4. Physical protection of shipments of irradiated reactor fuel

    International Nuclear Information System (INIS)

    Kasun, D.J.

    1979-05-01

    During May 1979 the U.S. Nuclear Regulatory Commission approved for issuance in effective form new interim regulations for strengthening the protection of spent fuel shipments against sabotage and diversion. The new regulations will likely continue in force until the completion of an ongoing research program concerning the response of spent fuel to certain forms of sabotage. At that time the regulations may be rescinded, modified, or made permanent, as appropriate. This report discusses the new regulations and provides a basis on which licensees can develop an acceptable interim program for the protection of spent fuel shipments

  5. Evaluation of liquid metal protection of a limiter/divertor in fusion reactors

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Smith, D.L.

    1988-01-01

    The liquid metal protection concept is proposed mainly to prolong the lifetime of a divertor or a limiter in a fusion reactor. This attractive idea for protection requires studying a wide range of problems associated with the use of liquid-metals in fusion reactors. In this work the protection by liquid-metals has concentrated on predictions of the loss rate of the film to the plasma, the operating surface temperatures required for the film, and the potential tritium inventory requirement. The effect of plasma disruptions on the liquid metal film is also evaluated. Other problems such as liquid metal compatibility with structural materials, magnetic field effects, and the effect of liquid metal contamination on plasma performance are discussed. Three candidate liquid-metals are evaluated, i.e., lithium, gallium, and tin. A wide range of reactor operating conditions valid for both near term machines (INTOR and ITER) and for the next generation commercial reactors (TPSS) are considered. This study has indicated that the evaporation rate for candidate liquid metals can be kept below the sputtering range for reasonable operating temperatures and plasma edge conditions. At higher temperatures, evaporation dominates the losses. Impurity transport calculations indicate that impurities from the plate should not reach the main plasma. One or two millimeters of liquid films can protect the structure from severe plasma disruptions. Depending on the design of the liquid metal protection system, the tritium inventory in the liquid film is predicted to be on the order of a few grams. 16 refs., 5 figs

  6. Nuclear plant-aging research on reactor protection systems

    International Nuclear Information System (INIS)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed

  7. Recent advances in reactor protection and control system technology

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    After a first-generation digital integrated protection system has been installed on all 1300 MWe PWR units in France, a new digital protection system was developed for the 1450 MWe units, using local area networks, fiber optics, Motorola 68000 microprocessors, and a modular design allowing for the design of any system on the basis of around 50 types of standard cards. In 1993, an upgrading program for this equipment was launched in order to reduce costs, in particular software development costs, further improve hardware modularity and facilitate integration and connection to existing equipment. The basic principles of the units are described together with the implementation of computer-aided software engineering (CASE) tools, interfaces with hard-wired equipment, and multiplexed connections. The nuclear instrumentation systems at the Fessenheim and Bugey plants have been renovated with these equipment

  8. Device for the burst protection of nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Daublebsky, P.

    1976-01-01

    The burst protection device has a hood over top and bottom of the pressure vessel with superimposed hinged supports lying in their turn against supporting rings which are connected with each other by vertical bracing. It is proposed to place an intermediate layer between hoods and vertical bracing absorbing thermal stresses, i.e. deforming plastically with gradually increasing pressure, but behaving like a rigid body in the case of shock loads. As a material lead e.g. is proposed. (UWI) [de

  9. 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

  10. Method and practice on safety software verification and validation for digital reactor protection system

    International Nuclear Information System (INIS)

    Li Duo; Zhang Liangju; Feng Junting

    2010-01-01

    The key issue arising from digitalization of reactor protection system for Nuclear Power Plant (NPP) is in essence, how to carry out Verification and Validation (V and V), to demonstrate and confirm the software is reliable enough to perform reactor safety functions. Among others the most important activity of software V and V process is unit testing. This paper discusses the basic concepts on safety software V and V and the appropriate technique for software unit testing, focusing on such aspects as how to ensure test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed herein was successfully used in the work of unit testing on safety software of a digital reactor protection system. (author)

  11. 75 FR 62695 - Physical Protection of Irradiated Reactor Fuel in Transit

    Science.gov (United States)

    2010-10-13

    ... Irradiated Reactor Fuel in Transit AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule. SUMMARY: The... nuclear fuel in transit? H. Why require a telemetric position monitoring system or an alternative tracking... nuclear fuel in transit. The interim final rule added 10 CFR 73.37, ``Requirements for Physical Protection...

  12. Considerations for tritium protection at a fusion reactor

    International Nuclear Information System (INIS)

    Easterly, C.E.

    1982-01-01

    The view on the radiological hazard associated with future fusion power stations as presented in this discussion is rarely supported by reasonably certain or reliably accurate prediction. This fact should not be taken as indicating a major programmatic deficiency. In fact, it is expected that large uncertainty would be present in health effect at the current level of technological development. The details of tritium exposure will be clarified, waiting for the operation of the Tritium System Test Assembly. Once the data base for the TSTA is established, future fusion design can be made based on economic cost/radiation exposure risk benefit. The actual execution of this cost/benefit analysis is complex because three populations are of interest: occupational work force, local population and global population. The knowledge of tritium management must be increased if D-T fusion reactors are to become compatible with the needs of utility companies. In order to exploit the differing hazard between HT and HTO, it is necessary to know much more about the mechanism of uncatalyzed conversion over a wide range of concentration and about the change caused by the variety of potential catalytic sequence in potential tritium leak. (Kako, I.)

  13. Layer protecting the surface of zirconium used in nuclear reactors

    Czech Academy of Sciences Publication Activity Database

    Ashcheulov, Petr; Škoda, R.; Škarohlíd, J.; Taylor, Andrew; Fendrych, František; Kratochvílová, Irena

    2016-01-01

    Roč. 10, č. 1 (2016), 59-65 ISSN 2212-4020 R&D Projects: GA ČR(CZ) GA15-05095S; GA TA ČR TA04020156; GA MŠk LO1409; GA ČR GA13-31783S; GA ČR(CZ) GA14-10279S Institutional support: RVO:68378271 Keywords : plasma enhanced chemical vapor deposition * Raman spectroscopy * SEM * thin polycrystalline diamond film * Zircaloy2 pins protection against oxidation Subject RIV: JF - Nuclear Energetics

  14. Reactor core protection system using a 4-channel microcomputer

    International Nuclear Information System (INIS)

    Mertens, U.

    1982-12-01

    A four channel microcomputer system was fitted in Grafenrheinfeld NPP for local core protection. This system performs continuous on-line monitoring of peak power density, departure from nucleate boiling ratio and fuel duty. The system implements limitation functions with more sophisticated criteria and improved accuracy. The Grafenrheinfeld system points the way to the employment of computer based limitation system, particularly in the field of programming language, demarkation of tasks, commissioning and documentation aids, streamlining of qualification and structuring of the system. (orig.) [de

  15. A reactor core/containment status evaluation flowchart for determining protective actions in emergencies

    International Nuclear Information System (INIS)

    Glissman, M.A.

    1988-01-01

    In the event of an emergency at a power reactor station, there might not be adequate time or sufficient data to fully assess radiological implications and make protective action recommendations based on projected population exposures. Thus, decision-making guidance is needed that is based on readily available plant indicators, not just on time-consuming dose calculations. In the United States, this guidance must be compatible with the recommended by the Nuclear Regulatory Commission and the Environmental Protection Agency, and it must include predetermined, measurable, site-specific parameters for assessing conditions in the reactor core and containment. The preparation of this real time guidance calls for the selection of suitable parameters and the determination of the values for these parameters that will correspond to different levels of protective action. This process is illustrated in this paper by selecting parameters and determining appropriate values for constructing a Core/Containment Status Evaluation Flowchart for an example power plant

  16. Environmental protection problems from the standpoint of regeneration of fast neutron reactor fuel

    International Nuclear Information System (INIS)

    Gedeonov, L.I.; Lazarev, L.N.; Suprunenko, A.N.

    The discussion of the problem of environmental protection is based on two principles: a strict observance of legislatively established standards for permissible concentrations of radionuclides in objects of the environment and for dose loads for the population; all possible steps to reduce the contamination to a level justified in practice. Environmental protection steps are considered from the points of view of a systematic analysis. A survey of the environmental protection system near sources of radioactive discharges is given. The basic interactions and feedbacks are indicated. Characteristics differentiating the discharges of the fuel cycle of fast neutron breeder reactors from discharges of the slow neutron cycle are discussed. It is shown that it is necessary to study the overall regional and global interactions of discharges of the atomic power industry. The characteristics of situations at nuclear fuel cycle facilities of fast neutron reactors are discussed. The necessity of additional technical steps to prevent accidents and eliminate their effects if they take place is emphasized

  17. A top priority problem of national radiation protection - proper disposal of research reactor spent fuel

    International Nuclear Information System (INIS)

    Marinkovic, N.; Matausek, M.V.; Jovic, V.

    1997-01-01

    The paper presents basic facts about RA research reactor at the Vinca Institute. The present state of the RA reactor spent fuel storage pool appears to be a serious safety and radiological problem, which must be solved urgently, independent of the decision about the future status of the reactor itself. The following paragraphs describe current activities on improving storage conditions of the research reactor RA spent fuel. Activities performed so far, concerning identification and improvement of the spent fuel storage conditions are presented. These are verification of radiation protection measures, radiological and chemical analyses, visual inspection and photographing, safety analyses and nuclear criticality studies.A project for long-term solution of the research reactor spent fuel storage is proposed. In order to minimise further corrosion and establish strict control of all the relevant technological parameters of the utility, improvement of conditions for disposal of the fuel in the existing storage, is foreseen in the first phase. New dry storage for long-term storing of the spent fuel should be built during the second phase of the project. Particular attention is paid to the activities related to radiation protection and waste treatment, starting from standard monitoring and control, radiological analyses, regulations and legislation, to complicated handling of high level radioactive waste. (authors)

  18. Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

    International Nuclear Information System (INIS)

    Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

    2012-01-01

    The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

  19. New digital control and power protection system of VR 1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Juoeickova, M.

    2005-01-01

    The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

  20. Utilization of a statistical procedure for DNBR calculation and in the survey of reactor protection limits

    International Nuclear Information System (INIS)

    Pontedeiro, A.C.; Camargo, C.T.M.; Galetti, M.R. da Silva.

    1987-01-01

    A new procedure is applied to Angra 1 NPP, which is related to DNBR calculations, considering the design parameters statistically: Improved Thermal Design Procedure (ITDP). The ITDP application leads to the determination of uncertainties in the input parameters, the sensitivity factors on DNBR. The DNBR limit and new reactor protection limits. This was done to Angra 1 with the subchannel code COBRA-IIIP. The analysis of limiting accident in terms of DNB confirmed a gain in DNBR margin, and greater operation flexibility of the plant, decreasing unnecessary trips of the reactor. (author) [pt

  1. Unavailability Analysis of the Reactor Core Protection System using Reliability Block Diagram

    International Nuclear Information System (INIS)

    Shin, Hyun Kook; Kim, Sung Ho; Choi, Woong Suk; Kim, Jae Hack

    2006-01-01

    The reactor core of nuclear power plants needs to be monitored for the early detection of core abnormal conditions to protect plants from a severe accident. The core protection calculator system (CPCS) has been provided to calculate the departure from nucleate boiling ratio (DNBR) and the local power density (LPD) based on measured parameters of reactor and coolant system. The original CPCS for OPR 1000 has been designed and implemented based on the concurrent 3205 computer system whose components are obsolete. The CPCS based on Westinghouse Common-Q system has recently been implemented for the Shin-Kori Nuclear Power Plant, Units 1 and 2(SKN 1 and 2). An R and D project has been launched to develop new core protection system called as RCOPS (Reactor Core Protection System) with the partnership of KOPEC and Doosan Heavy Industries and Construction Co. RCOPS is implemented on the HFC-6000 safety class programmable logic controller (PLC). In this paper, the reliability of RCOPS is analyzed using the reliability block diagram (RBD) method. The calculated results are compared with that of the CPCS for SKN 1 and 2

  2. Protection method and protection device for liquid supply channel for nuclear reactor

    International Nuclear Information System (INIS)

    Sato, Masanori; Fujimoto, Sachiko

    1998-01-01

    In the present invention, thermal stresses exerted on feedwater pipelines and a feedwater nozzle portion in a LBR type reactor are reduced to improve integrity of a reactor, suppress addition of facilities and reliably reduce thermal stresses by a simple structure. A connection pipe channel is formed between the upper side of a horizontal portion of a pipeline for feeding water to a reactor pressure vessel and a vertical portion of the feedwater pipeline. A transferring pump is disposed in the midway of the connection pipe channel for sucking supernatant water in the horizontal portion and rendering it to join with water in the vertical portion. When supply of water is stagnated, high temperature water in the horizontal portion is transferred by the action of the transfer pump to the low temperature water in the vertical portion to join them. With such procedures, the water supplied to the horizontal portion in the feedwater pipeline is flown to suppress occurrence of heat-stratification phenomenon thereby enabling to reduce the temperature difference. (T.M.)

  3. Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

    Varde, P. V.; Lee, D. Y.; Han, J. B.

    2003-03-01

    A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

  4. Comparison of radioactive doses after the last protection layer insight the reactor structure for Russian VVER-1000 and German PWR-1300 reactors

    International Nuclear Information System (INIS)

    Rahimi, A.; Mansourshaiflu, N.; Alizadeh, M. R.

    2004-01-01

    In pressurized reactors (VVER and PWR), various protections layers are used for reducing the output core doses. At any protection layer, some amount of neutron and gamma doses is reduced. In this project the axial flux of neutron and gamma beams have been evaluated at various protection layers in the operation state the German PWR-1300 and Russian VVER-1000 reactors by the MCNP computer code. For the purpose of effective use of the MCNP code and assuring its correct performance about of fluxed beams common and series of scientific answers and bench marks should be considered and the results obtained by the MCNP code, be compared with this answers. Then by using appropriate method, for reducing the flux variants of neutron and gamma beams at various protection layers of German PWR-1300 and Russian VVER-1000 reactors of the operation state of both reactors have been accelerated. In this projects, bench marks are computations and numbers existing in PSAR's present at Bushehr nuclear power plant. At the end, by using the results obtained and the standard doses, the time which a person can have work activity at the reactor wall (after the last protection layer), was compared for the operation status of the German PWR-1300 and Russian VVER-1000 reactors

  5. CHANGE IN DEFORMATION PROPERTIES MODELING OF CONCRETE IN PROTECTIVE STRUCTURES OF NUCLEAR REACTOR BY IONIZING RADIATION

    Directory of Open Access Journals (Sweden)

    E. K. Agakhanov

    2016-01-01

    Full Text Available The necessity of studying the effect impact of elementary particles impact on the strength and deformation materials properties used in protective constructions nuclear reactors and reactor technology has been stipulated. A nuclear reactor pressure vessel from prestressed concrete, combining the functions of biological protection is to be considered. The neutron flux problem distribution in the pressure vessel of a nuclear reactor has been solved. The solution is made in axisymmetric with the finite element method using a flat triangular finite element. Computing has been conducted in Matlab package. The comparison with the results has been obtained using the finite difference method, as well as the graphs of changes under the influence of radiation exposure and the elastic modulus of concrete radiation deformations have been constructed. The proposed method allows to simulate changes in the deformation properties of concrete under the influence of neutron irradiation. Results of the study can be used in the calculation of stress-strain state of structures, taking into account indirect heterogeneity caused by the physical fields influence.

  6. Development and Reliability Analysis of HTR-PM Reactor Protection System

    International Nuclear Information System (INIS)

    Li Duo; Guo Chao; Xiong Huasheng

    2014-01-01

    High Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM) digital Reactor Protection System (RPS) is a dedicated system, which is designed and developed according to HTR-PM NPP protection specifications. To decrease the probability of accident trips and increase the system reliability, HTR-PM RPS has such features as a framework of four redundant channels, two diverse sub-systems in each channel, and two level two-out-of-four logic voters. Reliability analysis of HTR-PM RPS is based on fault tree model. A fault tree is built based on HTR-PM RPS Failure Modes and Effects Analysis (FMEA), and special analysis is focused on the sub-tree of redundant channel ''2-out-of-4'' logic and the fault tree under one channel is bypassed. The qualitative analysis of fault tree, such as RPS weakness according to minimal cut sets, is summarized in the paper. (author)

  7. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  8. Electromagnetic drive of the control and protection system of a nuclear reactor

    International Nuclear Information System (INIS)

    Zav'yalova, G.I.

    1983-01-01

    The design and operating principle of an electromagnetic drive with a linear synchronous reaction motor are described. At the present time, electromagnetic control mechanisms using linear electric motors are finding increasingly widespread application as drives for the control and protection system of nuclear reactors. In these drives there is a functional mergence of the electromagnetic mechanism with the final control element; these drives, therefore, have advantages over electromechanical drives

  9. Research and Development of Protection OPC server for China advanced research reactor digital monitoring system

    International Nuclear Information System (INIS)

    Jia Yuwen; Xu Qiguo

    2012-01-01

    OPC server was developed as I/O driver to communicate the digital monitoring system of China Advanced Research Reactor iFIX and protection system. The framework and working principle of the OPC server were researched, and an effective method was developed to resolve the special communication protocol. After commissioning and testing, the results show that this method is reliable and stable, makes the system easy to configure, and can reduce the complexity of the system. (authors)

  10. Optimum supervision intervals and order of supervision in nuclear reactor protective systems

    International Nuclear Information System (INIS)

    Kontoleon, J.M.

    1978-01-01

    The optimum inspection strategy of an m-out-of-n:G nuclear reactor protective system with nonidentical units is analyzed. A 2-out-of-4:G system is used to formulate a multi-variable optimization problem to determine (a) the optimum order of supervision of the units and (b) the optimum supervision intervals between units. The case of systems with identical units is a special case of the above. Numerical results are derived using a computer algorithm

  11. The protection system of nuclear reactors. Example of a very high reliability system. Its evolution

    International Nuclear Information System (INIS)

    Weill, J.

    1980-06-01

    The present state of reactor protection is described and mention is made of certain evolutionary trends towards completely automated systems which either help the operator to take decisions in the event of an accident or take and execute these decisions for him. To do so, the use of models and recourse to complex data processing systems is necessary. This ensemble reflects an evolution of the reliability of the equipment towards that of the software [fr

  12. The Automatic Test Features of the IDiPS Reactor Protection System

    International Nuclear Information System (INIS)

    Hur, Seop; Kim, Dong-Hoon; Hwang, In-Koo; Lee, Cheol-Kwon; Lee, Dong-Young

    2007-01-01

    The reactor protection system (RPS) is designed to minimize a propagation of abnormal or accident conditions of nuclear power plants. A digital RPS (Integrated Digital Protection System (IDiPS) RPS) is being developed in the Korea Nuclear Instrumentation and Control System (KNICS) R and D project. To make good use of the advantages of the digital technology, it is necessary to improve the reliability and availability of a system through automatic test features including an on-line testing, a self-diagnostics, an auto calibration, etc. This paper summarizes the system test strategy and the automatic test features of the IDiPS RPS

  13. Environment report 1990 of the Federal Minister for the Environment, Nature Protection and Reactor Safety

    International Nuclear Information System (INIS)

    1990-01-01

    The 'Environment Report 1990' describes the environmental situation in the Federal Republic of Germany; draws a balance of environmental policy measures taken and introduced; gives information on future fields of action in environmental policy. The 'Environment Report 1990' also deals with the 'Environment Expert Opinion 1987', produced by the board of experts on environmental questions. It contains surveys of the following sectors: Protection against hazardous materials air pollution abatement, water management, waste management, nature protection and preservation of the countryside, soil conservation, noise abatement, radiation protection, reactor safety. A separate part of the 'Environment Report 1990' deals with the progress made in 'interdisciplinary fields' (general law on the protection of the environment, instruments of environmental policy, environmental information and environmental research, transfrontier environmental policy). (orig./HP) [de

  14. 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.

  15. Dynamic response of aircraft impact of a reactor building with protective shell on independent foundation

    International Nuclear Information System (INIS)

    Constantopoulos, I.V.; Vardanega, C.; Attalla, I.

    1981-01-01

    Aircraft impact loading can penalize significantly the design of the equipment in a conventional containment building. An alternative scheme was developed in an attempt to reduce the aircraft impact response. A preliminary study was carried out to investigate the feasibility of the alternative scheme. This study was made in such perspective and for the purpose of comparing the response to aircraft impact of a standard reactor building, to that of a reactor building having an independently founded outer shell. In the second scheme, the outer shell is meant to receive the aircraft impact, so that the load will be transmitted to the reactor building internals only by way of the structure-soil-structure system. In both cases, the aircraft impact was postulated to occur on a linear single degree of freedom oscillator which modeled, approximately, the plastification of the impact area. The soil was considered as a half-space with properties corresponding to a medium stiff soil, and modeled by lumped soil springs and dashpots. The reactor internals, inner shell and protective outer shell were modeled with beam elements and concentrated inertias. In modeling the coupled system, soil-structure interaction and structure-to-structure interaction through the soil were represented by a global stiffness matrix corresponding to the three degrees the freedom of each foundation, i.e. horizontal, vertical and rocking. (orig./HP)

  16. Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

    International Nuclear Information System (INIS)

    Parkes, Alison; Saratsopoulos, John; Deura, Michael; Kenny, Pat

    2008-01-01

    The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

  17. The overpressure protection for the chemical reactors: the batch-size approach

    International Nuclear Information System (INIS)

    Dellavedova, M.; Gigante, L.; Lunghi, A.; Pasturenzi, C.; Cardillo, P.; Gerosa, N.P.; Rota, R.

    2008-01-01

    Small and medium enterprises (SMEs) main feature is to run batch and semi-batch processes, working on job orders. They generally have multi propose reactors with an emergency relief system (ERS) already installed. These are normally sized when the reactor is designed, assuming as worst incidental scenario a single phase vapour flow generated by a fire developed outside the apparatus. These assumptions can lead to a big underestimation of the vent area if the actual flow is two-phase and besides generated by a runaway reaction. ERS sizing is particularly hazardous and complex for small mills, as for example fine chemicals and pharmaceutical companies. These factories have usually narrow financial and personal resources, moreover they often use fast processes turnovers. In many cases a complete safety study or the replacement of the ERS is not possible and it can lead to not sustainable costs. The batch-size approach is focused on discontinuous process conditions: aim of this approach is to find the reactor fill level that can lead a vapour single phase flow whether an incident occurs, this condition is considered safe that the ERS installed on the reactor can protect the plant from explosions [it

  18. Computer based systems for fast reactor core temperature monitoring and protection

    International Nuclear Information System (INIS)

    Wall, D.N.

    1991-01-01

    Self testing fail safe trip systems and guardlines have been developed using dynamic logic as a basis for temperature monitoring and temperature protection in the UK. The guardline and trip system have been tested in passive operation on a number of reactors and a pulse coded logic guardline is currently in use on the DIDO test reactor. Acoustic boiling noise and ultrasonic systems have been developed in the UK as diverse alternatives to using thermocouples for temperature monitoring and measurement. These systems have the advantage that they make remote monitoring possible but they rely on complex signal processing to achieve their output. The means of incorporating such systems within the self testing trip system architecture are explored and it is apparent that such systems, particularly that based on ultrasonics has great potential for development. There remain a number of problems requiring detailed investigation in particular the verification of the signal processing electronics and trip software. It is considered that these problems while difficult are far from insurmountable and this work should result in the production of protection and monitoring systems suitable for deployment on the fast reactor. 6 figs

  19. Reliability analysis of protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

    Varde, P. V.; Choi, J. G.; Lee, D. Y.; Han, J. B.

    2003-04-01

    Reliability analysis was carried out for the protection system of the Korean Advanced Pressurized Water Reactor - APR 1400. The main focus of this study was the reliability analysis of digital protection system, however, towards giving an integrated statement of complete protection reliability an attempt has been made to include the shutdown devices and other related aspects based on the information available to date. The sensitivity analysis has been carried out for the critical components / functions in the system. Other aspects like importance analysis and human error reliability for the critical human actions form part of this work. The framework provided by this study and the results obtained shows that this analysis has potential to be utilized as part of risk informed approach for future design / regulatory applications

  20. Method for performing diversity and defense-in-depth analyses of reactor protection systems

    International Nuclear Information System (INIS)

    Preckshot, G.G.

    1994-12-01

    The purpose of this NUREG is to describe a method for analyzing computer-based nuclear reactor protection systems that discovers design vulnerabilities to common-mode failure. The potential for common-mode failure has become an important issue as the software content of protection systems has increased. This potential was not present in earlier analog protection systems because it could usually be assumed that common-mode failure, if it did occur, was due to slow processes such as corrosion or premature wear-out. This assumption is no longer true for systems containing software. It is the purpose of the analysis method described here to determine points of a design for which credible common-mode failures are uncompensated either by diversity or defense-in-depth

  1. Stability of the lithium waterfall first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion (ICF) reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived which predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  2. Stability of the lithium ''WATERFALL'' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Abel-Khalik, S.I.; Paul, D.D.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular ''waterfall'' of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet breakup length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  3. Stability of the lithium 'waterfall' first wall protection concept for inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Esser, P.D.; Paul, D.D.; Abdel-Khalik, S.I.

    1981-01-01

    Uncertainties regarding the feasibility of using an annular waterfall of liquid lithium to protect the first wall in inertial confinement fusion reactor cavities have prompted a theoretical investigation of annular jet stability. Infinitesimal perturbation techniques are applied to an idealized model of the jet with disturbances acting upon either or both of the free surfaces. Dispersion relations are derived that predict the range of disturbance frequencies leading to instability, as well as the perturbation growth rates and jet break-up length. The results are extended to turbulent annular jets and are evaluated for the lithium waterfall design. It is concluded that inherent instabilities due to turbulent fluctuations will not cause the jet to break up over distances comparable to the height of the reactor cavity

  4. Modelling of reactor control and protection systems in the core simulator program GARLIC

    International Nuclear Information System (INIS)

    Beraha, D.; Lupas, O.; Ploegert, K.

    1984-01-01

    For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

  5. Device for protecting the containment vessel dome of a nuclear reactor

    International Nuclear Information System (INIS)

    Allain, A.; Filloleau, E.; Mulot, P.

    1976-01-01

    A device is disclosed for protecting the dome of a nuclear reactor containment vessel against the upward displacement of the concrete shield slab of said reactor and the resultant effects of tilting of an equipment unit mounted on the shield slab at the periphery of said slab, wherein said device comprises: (1) means for separating the equipment unit into two sections consisting of an upper section and a lower section, said lower section being rigidly fixed to said shield slab and said means being actuated by the upward displacement of said slab, (2) a system for vertical rectilinear guiding of said upper section within the containment vessel, and (3) rigid mechanical components which provide a coupling between the aforesaid upper and lower sections of the equipment unit and exert on said upper section under the action of the tilting motion of said lower section a thrust which causes the upward displacement of said upper section

  6. Application Software for the Cabinet Operator Module of the Reactor Protection System

    International Nuclear Information System (INIS)

    Lee, Hyun-Chul; Jung, Hae-Won; Lee, Sung-Jin; Koo, Young-Ho; Kim, Seong-Tae; Kwak, Tae-Kil; Jin, Kyo-Hong

    2006-01-01

    A reactor protection system (RPS) plays the roles of generating the reactor trip signal and the engineered safety features (ESF) actuation signal when the monitored plant processes reach predefined limits. A Korean project group, so-called KNICS (Korean Nuclear I and C System), is developing a new digitalized RPS and the Cabinet Operator Module (COM) of the RPS which is used for the RPS integrity testing and monitoring by equipment operators. A flat panel display (FPD) with a touch screen capability is provided as a main user interface for the RPS. This paper shows the application software developed for the COM FPD. Equipment operators can monitor the status of the RPS and carry out various tests to verify system functions by means of the application software. A qualified hardware and software development environment are used to develop the application software

  7. A coincidencd logic reactor protection system with automatic and permanent testing

    International Nuclear Information System (INIS)

    Tricornot.

    1978-01-01

    Within the context of sodium-cooled fast reactors, the CGEE Alsthom enterprise, under consulting for Novatome, has been engaged with the development of an specific protection system for emergency shutdown situations. System described has been conceived on several stages according to the general organization shown on figure I of the Annex, in such a manner that the exigences and recommendations from the safety regulatory authorities are respected and, at the same time, it is assured a significant reactor operation availability without an spureous rod drop. As an example, a selection of principles, rules and criteria currently applied to the development of a system of this kind is reminded. (J.E. de C.)

  8. Radioactivity, radiation protection and monitoring during dismantling of light-water reactors

    International Nuclear Information System (INIS)

    Hummel, L.; Zech, J.B.

    2005-01-01

    Based on the radioactivity inventory in the systems and components of light-water reactors observed during operation, the impact of actions during plant emptying after the conclusion of power operation and possible subsequent long-term safe enclosure concerning the composition of the nuclide inventory of the plant to be dismantled will be described. Derived from this will be the effects on radioactivity monitoring in the plant, physical radiation protection monitoring, and the measured characterization of the residual materials resulting from the dismantling. The impact of long-term interim storage will also be addressed in the discussion. The talk should provide an overview of the interrelationships between source terms, decay times and the radioactivity monitoring requirements of the various dismantling concepts for commercial light-water reactors. (orig.)

  9. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations. Published on September 24, 2012

    International Nuclear Information System (INIS)

    Couturier, Jean; Bruna, Giovanni; Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Hache, Georges

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  10. Results of neutron physics analyses of WWER-440 cores with modified reactor protection and control systems

    International Nuclear Information System (INIS)

    Lehmann, M.; Pecka, M.; Rocek, J.; Zalesky, K.

    1993-12-01

    Detailed results are given of neutron physics analyses performed to assess the efficiency and acceptability of modifications of the WWER-440 core protection and control system; the modifications have been proposed with a view to increasing the proportion of mechanical control in the compensation of reactivity effects during reactor unit operation in the variable load mode. The calculations were carried out using the modular MOBY-DICK macrocode system together with the SMV42G36 library of two-group parametrized diffusion constants, containing corrections which allow new-design WWER-440 fuel assemblies to be discriminated. (J.B). 37 tabs., 18 figs., 5 refs

  11. Safety assessment principles for reactor protection systems in the United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Philp, W

    1990-07-01

    The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems.

  12. Safety assessment principles for reactor protection systems in the United Kingdom

    International Nuclear Information System (INIS)

    Philp, W.

    1990-01-01

    The duty of Nuclear Installations Inspectorate (NII) is to see that the appropriate standards are developed, achieved and maintained by the plant operators, and to monitor and regulate the safety of the plant by means of its powers under the licence. It does not issue standards or codes of practice for NPPs, but it requires each plant operator to develop its own safety criteria and requirements. The following relevant issues are described: NII assessment principles and societal risks; principles and guidance for the assessment of rector protection systems; assessment of reactor shutdown systems

  13. Development of digital plant protection system for Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Suk-Joon Park

    1998-01-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author)

  14. Development of digital plant protection system for Korean Next Generation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suk-Joon [NSSS Engineering and Development Division, Korea Power Engineering Company, Taejon (Korea, Republic of)

    1998-10-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author) 8 refs, 4 figs, 3 tabs

  15. Cyber Security Analysis by Attack Trees for a Reactor Protection System

    International Nuclear Information System (INIS)

    Park, Gee-Yong; Lee, Cheol Kwon; Choi, Jong Gyun; Kim, Dong Hoon; Lee, Young Jun; Kwon, Kee-Choon

    2008-01-01

    As nuclear facilities are introducing digital systems, the cyber security becomes an emerging topic to be analyzed and resolved. The domestic and other nation's regulatory bodies notice this topic and are preparing an appropriate guidance. The nuclear industry where new construction or upgrade of I and C systems is planned is analyzing and establishing a cyber security. A risk-based analysis for the cyber security has been performed in the KNICS (Korea Nuclear I and C Systems) project where the cyber security analysis has been applied to a reactor protection system (RPS). In this paper, the cyber security analysis based on the attack trees is proposed for the KNICS RPS

  16. Radiological protection issues arising during and after the Fukushima nuclear reactor accident

    International Nuclear Information System (INIS)

    González, Abel J; Akashi, Makoto; Sakai, Kazuo; Yonekura, Yoshiharu; Boice Jr, John D; Chino, Masamichi; Homma, Toshimitsu; Ishigure, Nobuhito; Kai, Michiaki; Kusumi, Shizuyo; Lee, Jai-Ki; Menzel, Hans-Georg; Niwa, Ohtsura; Yamashita, Shunichi; Weiss, Wolfgang

    2013-01-01

    Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with ‘contamination’ of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of

  17. Radiological protection issues arising during and after the Fukushima nuclear reactor accident.

    Science.gov (United States)

    González, Abel J; Akashi, Makoto; Boice, John D; Chino, Masamichi; Homma, Toshimitsu; Ishigure, Nobuhito; Kai, Michiaki; Kusumi, Shizuyo; Lee, Jai-Ki; Menzel, Hans-Georg; Niwa, Ohtsura; Sakai, Kazuo; Weiss, Wolfgang; Yamashita, Shunichi; Yonekura, Yoshiharu

    2013-09-01

    Following the Fukushima accident, the International Commission on Radiological Protection (ICRP) convened a task group to compile lessons learned from the nuclear reactor accident at the Fukushima Daiichi nuclear power plant in Japan, with respect to the ICRP system of radiological protection. In this memorandum the members of the task group express their personal views on issues arising during and after the accident, without explicit endorsement of or approval by the ICRP. While the affected people were largely protected against radiation exposure and no one incurred a lethal dose of radiation (or a dose sufficiently large to cause radiation sickness), many radiological protection questions were raised. The following issues were identified: inferring radiation risks (and the misunderstanding of nominal risk coefficients); attributing radiation effects from low dose exposures; quantifying radiation exposure; assessing the importance of internal exposures; managing emergency crises; protecting rescuers and volunteers; responding with medical aid; justifying necessary but disruptive protective actions; transiting from an emergency to an existing situation; rehabilitating evacuated areas; restricting individual doses of members of the public; caring for infants and children; categorising public exposures due to an accident; considering pregnant women and their foetuses and embryos; monitoring public protection; dealing with 'contamination' of territories, rubble and residues and consumer products; recognising the importance of psychological consequences; and fostering the sharing of information. Relevant ICRP Recommendations were scrutinised, lessons were collected and suggestions were compiled. It was concluded that the radiological protection community has an ethical duty to learn from the lessons of Fukushima and resolve any identified challenges. Before another large accident occurs, it should be ensured that inter alia: radiation risk coefficients of potential

  18. International guidelines for fire protection at nuclear installations including nuclear fuel plants, nuclear fuel stores, teaching reactors, research establishments

    International Nuclear Information System (INIS)

    The guidelines are recommended to designers, constructors, operators and insurers of nuclear fuel plants and other facilities using significant quantities of radioactive materials including research and teaching reactor installations where the reactors generally operate at less than approximately 10 MW(th). Recommendations for elementary precautions against fire risk at nuclear installations are followed by appendices on more specific topics. These cover: fire protection management and organization; precautions against loss during construction alterations and maintenance; basic fire protection for nuclear fuel plants; storage and nuclear fuel; and basic fire protection for research and training establishments. There are numerous illustrations of facilities referred to in the text. (U.K.)

  19. Evaluation of mean time between forced outage for reactor protection system using RBD and failure rate

    International Nuclear Information System (INIS)

    Lee, D. Y.; Park, J. H.; Hwang, I. K.; Cha, K. H.; Choi, J. K.; Lee, K. Y.; Park, J. K.

    2001-01-01

    The design life of nuclear power plants (NPPs) under recent construction is about fifty to sixty years. However, the duration that equipments of control systems operate without failures is at most five to ten years. Design for diversity and adequate maintenance strategy are required for NPP protection system in order to use the control equipment which has shorter life time than the design life of NPP. Fault Tree Analysis (FTA) technique, which has been applied to Probabilistics Safety Analysis (PSA), has been introduced to quantitatively evaluate the reliability of NPP I and C systems. The FTA, however, cannot properly consider the effect of maintenance. In this work, we have reviewed quantitative reliability evaluation techniques using the reliability block diagram and failure rates and applied it to the evaluation of mean time between forced outage for reactor protection system

  20. Protective interior wall and attaching means for a fusion reactor vacuum vessel

    International Nuclear Information System (INIS)

    Phelps, R.D.; Upham, G.A.; Anderson, P.M.

    1988-01-01

    A protective wall for the interior surface of a fusion reactor vessel wall is described comprising: an array of plates, each plate of the array including a main body section, a pair of edge sections bent at an angle with respect to the main body section, and a pair of flange-like end sections each having protruding sections with cut-aways therein, the protruding sections of the flange-like end sections extending in a direction substantially parallel to the main body section; and means operatively associated with the protruding sections of the flange-like end sections of the plates for mounting the array of plates to an associated vessel wall to be protected

  1. A bustling academic reactor creates challenges and opportunities in the area of physical protection

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.; Standley, V.

    2001-01-01

    The 250 kW TRIGA research reactor is located at the Atominstitut Vienna, Austria, only a few subway stations from the city centre of Vienna. Its main purpose is the training of university students in the field of nuclear engineering and radiation protection as well as in radiochemistry and neutron- and solid-state physics. The existing facility is visited during a normal academic year by about 300 persons per day falling into seven different categories including fully employed staff, students occasionally visiting a seminar, and IAEA personnel from all over the world. These different groups have to be accounted for daily and are separated into different categories in view of security and physical protection. (author)

  2. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

    International Nuclear Information System (INIS)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

    2014-10-01

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  3. Enhancing the functionality of reactor protection systems to provide diagnostic and monitoring information: The ISATTM approach

    International Nuclear Information System (INIS)

    Baldwin, J.A.; Rowe, B.J.; Jones, C.D.

    1996-01-01

    The ISAT TM architecture has been successfully implemented as the Single Channel Trip System (SCTS), part of the primary protection system of Nuclear Electric's Dungeness 'B' Advanced Gas-Cooled Reactors. The system is the first computer-based protection system licensed on a UK civil nuclear reactor. The system provides protection against single channel faults resulting in high coolant gas outlet temperature. The SCTS was designed to output data at several points in the system to an Ethernet to allow checks to be made on the operation of parts of the protection system and the system as a whole. In order to monitor the performance of this shutdown system a PC based monitoring system was developed to take input as data from the Ethernet, check its integrity and then analyze the data to provide information of the state of the system and subsystems. The SCTS monitor was basically intended to alert the operator to any fault on the safety system and indicate its source, provide a diagnosis of the cause of any trip initiated by the safety system, and log the occurrences of these incidents for later inspection. The intention was also to provide accurate real-time information on the thermocouple readings and to decrease the effort required to maintain the safety system. This paper will describe briefly the development of the ISAT TM monitoring system: how its requirements were arrived at, and how the design, code and testing were carried out to ensure approval for this application. It will then go on to report how the ISAT TM monitor has performed during its time in service: how more functionality has been added over and above its original requirements. Features of additional monitors for the SCTS and other ISAT TM systems will also be described. (author). 2 refs, 5 figs

  4. Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Louison, R.; Boardman, C.E.

    1981-01-01

    The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events

  5. The implementation and evaluation of physical protection system of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Vaz, Antonio Carlos Alves

    2016-01-01

    The September 11, 2001 terrorist attacks in New York, the accident at the Fukushima nuclear power plant on March 2011 and the recent attacks in Paris on November 2015 are examples of events that justify the efforts of the International Agency of Energy Atomic - IAEA to improve security at nuclear facility. The Brazilian government has been collaborating with this project and investing resources to improve the Physical Protection System - PPS of the nuclear research reactor system, technically is associated with the elements of detection, delay and response. The PPS is an integrated system of people, equipment and procedures used to protect nuclear facilities and radioactive sources against threat, theft or sabotage. The PPS works to avoid, to mitigate or to minimize the consequences caused by these actions. This study evaluates the PPS of the reactor, identifying the vulnerabilities and suggesting ways to improve the system effectiveness. The analyses were based on the methodology developed by Sandia National Laboratories´ security experts in Albuquerque - USA, allowing the system evaluation through hypothetical and probabilistic analyzes; identifying threats, determining the targets and analyzing the possible adversaries paths. From the methodology adopted was obtained the value around 40% for PE indicator, which shows the need to improve the system to minimizing the vulnerabilities. (author)

  6. Safety-technical comparison of a hard-wired dynamic reactor protection system with a computerized alternative. Pt. 2

    International Nuclear Information System (INIS)

    Buettner, W.E.

    1978-03-01

    The investigation presented here compares a conventional hard-wired dynamic reactor protection system with a computerized alternative. For both systems the spurious trip probability is determined, i.e. the probability of a false release of one or more protection actions due to failures. The mean unavailability is also determined for those not distinctly safety-oriented protection actions of the computerized protection system which are released via the open circuit current principle. This is because system breakdowns prevent the release of those protection actions in case of demand. The advantages and disadvantages of either type of system are viewed against each other. (orig.) [de

  7. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  8. Possibilities of achieving non-positive void reactivity effect in fast sodium-cooled reactors with increased self-protection

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Yu.A.; Morozov, A.G.; Orlov, V.V.; Slesarev, I.S.; Subbotin, S.A.

    1989-01-01

    The problems of self-protection inhancement for the liquid-metal cooled fast reactors with intra-assembly heterogeneity of the core are studied. Possible approaches to arrangement of such reactors with various powers characterized by high levels of coolant natural circulation, minimum reactivity changes during fuel burn-up and non-positive void effect of reactivity are found. 10 refs.; 11 figs

  9. Fail-safe design criteria for computer-based reactor protection systems

    International Nuclear Information System (INIS)

    Keats, A.B.

    1980-01-01

    The increasing quantity and complexity of the instrumentation required in nuclear power plants provides a strong incentive for using on-line computers as the basis of the control and protection systems. On-line computers using multiplexed sampled data are already well established but their application to nuclear reactor protection systems requires special measures to satisfy the very high reliability which is demanded in the interests of safety and availability. Some existing codes of practice relating to segregation of replicated subsysttems continue to be applicable and lead to division of the computer functions into two distinct parts. The first computer, referred to as the Trip Algorithm Computer may also control the multiplexer. Voting on each group of status inputs yielded by the trip algorithm computers is performed by the Vote Algorithm Computer. The conceptual disparities between hardwired reactor-protection systems and those employing computers also rise to a need for some new criteria. An important objective of these criteria, minimising the need for a failure-mode-and-effect-analysis of the computer software, but is achieved almost entirely by 'hardware' properties of the system: the systematic use of hardwired test inputs which cause excursions of the trip algorithms into the tripped state in a uniquely ordered but easily recognisable sequence, and the use of hardwired 'pattern recognition logic' which generates a dynamic 'healthy' stimulus for the shutdown actuators only in response to the unique sequence generated by the hardwired input signal pattern. The adoption of the proposed design criteria ensure not only failure-to-safety in the hardware but the elimination, or at least minimisation, of the dependence on the correct functioning of the computer software for the safety system. (auth)

  10. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  11. Twenty years of Radiology in RP-10 nuclear reactor protection; Veinte anos de proteccion radiologica en el reactor nuclear RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Zapata, Alejandro L.; Ramos, Fernando T.; Arrieta, Rolando W.B.; Vela Mora, Mariano, E-mail: lzapata@ipen.gob.pe, E-mail: framos@ipen.gob.pe, E-mail: rarrieta@ipen.gob.pe, E-mail: mvela@ipen.gob.pe [Instituto Peruano de Energia Nuclear (IPEN), Lima (Peru)

    2013-07-01

    In this report we present the results about radiation controls during 1990 - 2010, carried out in the Nuclear Reactor RP-10 of the Nuclear Center of Huarangal. These controls and radiological evaluation are of much utility for the radio personnel protection of this one and other reactors, since it allows to compares these variables with respect to the time. From the results obtained in monitoring and radiation controls, we conclude that in no case it has been reached the limits allowed by the Peruvian Regulating Authority. (author)

  12. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

    1994-04-01

    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

  13. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    International Nuclear Information System (INIS)

    Korsah, K.; Clark, R.L.; Wood, R.T.

    1994-04-01

    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I ampersand C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems' environmental qualification and functional reliability. To bound the problem of new I ampersand C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I ampersand C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I ampersand C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software

  14. Physical protection of shipments of irradiated reactor fuel; Interim guidance. Regulatory report

    International Nuclear Information System (INIS)

    1980-06-01

    During May, 1979, the U.S. Nuclear Regulatory Commission approved for issuance in effective form new interim regulations for strengthening the protection of spent fuel shipments against sabotage and diversion. The new regulations were issued without benefit of public comment, but comments from the public were solicited after the effective date. Based upon the public comments received, the interim regulations were amended and reissued in effective form as a final interim rule in May, 1980. The present document supersedes a previously issued interim guidance document, NUREG-0561 (June, 1979) which accompanied the original rule. This report has been revised to conform to the new interim regulations on the physical protection of shipments of irradiated reactor fuel which are likely to remain in effect until the completion of an ongoing research program concerning the response of spent fuel to certain forms of sabotage, at which time the regulations may be rescinded, modified or made permanent, as appropriate. This report discusses the amended regulations and provides a basis on which licensees can develop an acceptable interim program for the protection of spent fuel shipments

  15. Protected plutonium breeding by transmutation of minor actinides in fast breeder reactor

    International Nuclear Information System (INIS)

    Meiliza, Yoshitalia; Saito, Masaki; Sagara, Hiroshi

    2008-01-01

    The improvement of proliferation resistance properties of Pu and the burnup characteristics of fast breeder reactor (FBR) had been studied by utilizing minor actinides (MAs) to produce more 238 Pu from 237 Np and 241 Am through neutron capture reaction. The higher the 238 Pu content in the fuel, the higher the proliferation resistance of the fuel would be owing to the natural characteristics of 238 Pu with high decay heat and high neutron production. The present paper deals with the assessment of passive measure against nuclear material proliferation by focusing on improving the inherent proliferation barrier of discharged Pu from an FBR. Results showed that 5% MA doping to the blanket of an FBR gives as high as 17-19% 238 Pu, which could be seen as a significant improvement of the proliferation properties of Pu. Moreover, additional 5% ZrH 2 , together with 5% MA doping to the blanket, could enhance the 238 Pu fraction much more (22-24%). With an assumption of protected Pu whose 238 Pu isotopic fraction is more than 12%, the present paper revealed that protected Pu could be produced more than the Pu consumed (protected Pu breeding) through incineration in an FBR with doping of a minimum 3% MAs or (2% MAs+5% ZrH 2 ) to the blanket. (author)

  16. Reactor safety and radiation protection. Draft of the BMU deparmental budget 16 of the 1998 German federal budget

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    The expenditures earmarked for reactor safety and radiation protection in the 1998 budget of the German Federal Ministry for the Environment, Nature Conservation, and Reactor Safety (BMU) total DM 101 million. The expenditures of the German Federal Office for Radiation Protection (BfS) are to amount to a total of DM 579 million. These are the figures included in departmental budget 16 of the 1998 federal budget, which was discussed by the Federal Parliament in September 1997. The atw compilation singles out a number of significant items of the departmental budget. (orig.) [de

  17. Optimisation of radiation protection for the new european pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Miniere, D.; Beneteau, Y.; Le Guen, Bernard

    2008-01-01

    Full text: As part of the EPR (European Pressurized Reactor) project being deployed at Flamanville, EDF has pro actively made the decision to focus on radiation protection (RP) aspects right from the start of the design phase, as it has done with nuclear safety. The approach adopted for managing RP-significant activities has been to include all involved stake holders -designers, licensee and contractor companies- in the three successive phases, starting with a survey among workers and designers, followed by a proposal review, and finally ending with the decision-making phase entrusted to an ALARA committee. The RP target set by EDF for this new reactor is to engage in an effort of continuous improvement and optimisation, through benchmarking with the best performing plants of the fleet. The collective dose target is currently set at 0.35 man.Sv/year per unit. In addition to other aspects, efforts will focus on shortening the duration of the highest-dose jobs, with a new challenge being set for work performed in the reactor building during normal operations, the aim being to improve plant availability. The plan is for work to be performed 7 days prior to shutting down the reactor and 3 days afterwards, in order to make logistical arrangements for forthcoming jobs. Without this reduction, the estimated drop is currently 4.5% of annual dose. For this purpose, two areas have been set up in the EPR 's reactor building: one no-go area for containing leaks from the primary circuit, and one accessible area for normal operations, separated from the no-go area by purpose-built ventilation equipment and facilities. To offer protection against radioactive flux (neutrons and high energy), RP studies have resulted in the installation of a concrete floor and of nuclear shielding at the outlets of primary circuit pipes. Steam generator bunkers and pumps have also been reinforced. All these measures will ensure that the accessible area can be posted as a green area (dose rate < 25

  18. 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

  19. Development of a user interface style guide for the reactor protection system cabinet operator module

    International Nuclear Information System (INIS)

    Lee, Hyun-Chul; Lee, Dong-Young; Lee, Jung-Woon

    2004-01-01

    The reactor protection system (RPS) plays the roles of generating the reactor trip signal and the engineered safety features (ESF) actuation signal when the monitored plant processes reach the predefined limits. A Korean project group is developing a new digitalized RPS and the Cabinet Operator Module (COM) of the RPS is used for the RPS integrity testing and monitoring by an equipment operator. A flat panel display (FPD) with a touch screen capability is provided as a main user interface for the RPS operation. To support the RPS COM user interface design, actually the FPD screen design, we developed a user interface style guide because the system designer could not properly deal with the many general human factors design guidelines. To develop the user interface style guide, various design guideline gatherings, a walk-though with a video recorder, guideline selection with respect to user interface design elements, determination of the properties of the design elements, discussion with system designers, and a conversion of the properties into the screen design were carried out. This paper describes the process details and the findings in the course of the style guide development. (Author)

  20. Design and implementation of STD32-BUS based reactor protection trip unit on FPGA imbaby

    International Nuclear Information System (INIS)

    Mahmoud, I.; Elnokity, O.A.; Refai, M.K.

    2007-01-01

    This paper presents a way to design and implement the Trip Unit of a Reactor Protection System (RPS) using a Field Programmable Gate Arrays (FPGA). Instead of the traditional embedded Microprocessor based interface design method, a proposed tailor made FPGA based circuit is built to substitute the Trip Unit (TL1) existing in Egypt's 2' ' Research reactor ETRR-2. The existing embedded system is built around the STD32 field Computer Bus which used in industrial and process control applications. It is modular, rugged, reliable, and easy-to-use and is able to support a large mix of I/O cards and to easily change its configuration in the future. Therefore, the state machine of this bus is extracted from its timing diagrams and implemented in VHDL to interface the designed TU circuit. The proposed designed circuit implemented using ALTERA EPF10K10LC84-3 chip replaces the Single Board Computer which have the embedded SAY program of the TU providing the same integrated HAV and SAV functions implemented in FPGA Chip housed in an printed circuit board, which uses the same shape and specifications of STD32 boards. H/W implementation of both TU and STD32 Bus in VHDL addresses the issues of safety and reusability

  1. Protection of spent aluminum-clad research reactor fuels during extended wet storage

    International Nuclear Information System (INIS)

    Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A.

    2013-01-01

    Aluminum-clad spent nuclear fuel from research reactors (RR) is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: (a) the preparation and characterization of hydrotalcite (HTC) coatings; (b) the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; (c) the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature (HT) bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. (author)

  2. Corrosion and protection of spent Al-clad research reactor fuel during extended wet storage

    International Nuclear Information System (INIS)

    Ramanathan, Lalgudi V.

    2009-01-01

    A variety of spent research reactor fuel elements with different fuel meats, geometries and 235 U enrichments are presently stored under water in basins throughout the world. More than 90% of these fuels are clad in aluminum (Al) or its alloy and are susceptible to corrosion. This paper presents an overview of the influence of Al alloy composition, galvanic effects (Al alloy/stainless steel), crevice effects, water parameters and synergism between these parameters as well as settled solids on the corrosion of typical Al alloys used as fuel element cladding. Pitting is the main form of corrosion and is affected by water conductivity, chloride ion content, formation of galvanic couples with rack supports and settled solid particles. The extent to which these parameters influence Al corrosion varies. This paper also presents potential conversion coatings to protect the spent fuel cladding. (author)

  3. Separation review program for reactor protection system and engineered safeguard systems in a nuclear power plant

    International Nuclear Information System (INIS)

    Lamb, F.J.; Walrod, B.E.

    1980-01-01

    This review program is utilized during the design of a nuclear power plant to insure separation between interdiscipline design for the Reactor Protection System (RPS) and Engineered Safeguard Systems (ESS). Color coded transparent drawings of the RPS and ESS are produced by each discipline. The separation is then reviewed by overlaying drawings of different disciplines on a light table. When this inspection shows that RPS or ESS elements have less than the established minimum separation, an analysis is performed to determine what, if any, design revision is necessary to insure proper separation. ''Hazard'' drawings are also made for determination of each type of potential hazard in each area of the plant. The review is a continuing process as the design progresses and is revised by any discipline. 5 refs

  4. Human Factors Support in the Design and Evaluation of the Reactor Protection System Cabinet Operator Module

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Lee, Jung Woon

    2005-01-01

    A Korean project group, KNICS, is developing a new digitalized reactor protection system (RPS) and the developed system will be packaged into a cabinet with several racks. The cabinet of the RPS is used for the RPS function testing and monitoring by maintenance operators and is equipped with a flat panel display (FPD) with a touch screen capability as a main user interface for the RPS operation. This paper describes the human factors activities involved in the development process of the RPS: conceptual design, design guidance, and evaluation. The activities include a functional requirements analysis and task analysis, user interface style guide for the RPS cabinet operator module (COM), and a human factors evaluation through an experiment and questionnaires

  5. Determination of the protection set-points lines for the Angra-1 reactor core

    International Nuclear Information System (INIS)

    Furieri, E.B.

    1980-03-01

    In this work several thermo-hidraulic calculation were performed to obtain Protection set-points lines for the Angra-1 reactor core in order to compare with the values presented by the vendor in the FSAR. These lines are the locus of points where DNBR min = 1,3 and power = 1,18 x P nominal as a function of ΔT m and T m , the temperature difference and the average coolant temperature between hot and cold legs. A computation scheme was developed using COBRA-IIIF as a subroutine of a new main program and adding new subroutines in order to obtain the desired DNBR. The solution is obtained through a convergentce procedure using parameters estimated in a sensivity study. (author) [pt

  6. Guideline for Bayesian Net based Software Fault Estimation Method for Reactor Protection System

    International Nuclear Information System (INIS)

    Eom, Heung Seop; Park, Gee Yong; Jang, Seung Cheol

    2011-01-01

    The purpose of this paper is to provide a preliminary guideline for the estimation of software faults in a safety-critical software, for example, reactor protection system's software. As the fault estimation method is based on Bayesian Net which intensively uses subjective probability and informal data, it is necessary to define formal procedure of the method to minimize the variability of the results. The guideline describes assumptions, limitations and uncertainties, and the product of the fault estimation method. The procedure for conducting a software fault-estimation method is then outlined, highlighting the major tasks involved. The contents of the guideline are based on our own experience and a review of research guidelines developed for a PSA

  7. Cyber Security Analysis by Attack Trees for a Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gee-Yong; Lee, Cheol Kwon; Choi, Jong Gyun; Kim, Dong Hoon; Lee, Young Jun; Kwon, Kee-Choon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    As nuclear facilities are introducing digital systems, the cyber security becomes an emerging topic to be analyzed and resolved. The domestic and other nation's regulatory bodies notice this topic and are preparing an appropriate guidance. The nuclear industry where new construction or upgrade of I and C systems is planned is analyzing and establishing a cyber security. A risk-based analysis for the cyber security has been performed in the KNICS (Korea Nuclear I and C Systems) project where the cyber security analysis has been applied to a reactor protection system (RPS). In this paper, the cyber security analysis based on the attack trees is proposed for the KNICS RPS.

  8. Refurbishment of the reactor protection system at Paks NPP. The refurbishment process

    International Nuclear Information System (INIS)

    Turi, T.; Katics, B.

    1998-01-01

    The Reactor Protection System Refurbishment Project had an extensive preparation period in Paks started in 1992. During this preparation a large volume of the basic engineering tasks has been performed and as a result a contract for implementation of a three-train digital RPS on the four Units was concluded with Siemens in September, 1996. According to that contract the first refurbished Unit will be commissioned in 1999 followed by a further Unit in each succeeding year. This paper introduces the process of the refurbishment, overview of the V and V activities, introduce the architecture, summarize the main design principles and outlines the additional tasks to be performed together with the RPS design. (author)

  9. United States Department of Energy's reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    International Nuclear Information System (INIS)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage

  10. Radiation protection at the RA reactor in 1984, Part II c Radioactivity control in the environment of the RA reactor, Meteorology measurements

    International Nuclear Information System (INIS)

    Grsic, Z.; Zaric, M.; Nikolic, R.; Stevanovic, M.

    1984-01-01

    During 1984, meteorology measurements were continued as a part of the environmental control of the Vinca Institute. This report covers the period from December 1983 - November 1984. Part of the meteorology measurements and data analysis is adapted to the needs of the Institute, i.e. RA reactor and some Laboratories. The objective of these activities is forming the data base for solving everyday and special problems related to control, protection and safety of Institute environment

  11. Technical improvement for the output drive unit of the reactor protection system in QNPP

    International Nuclear Information System (INIS)

    Jiang Zuyue

    1995-11-01

    For improving the reliability of the output drive unit of the reactor protection system in Qinshan NPP, the former design of this part was improved and researched on the problem appeared during the commissioning and operation under the conditions of narrow process space of cabinets and unchanged overall arrangement: (1) The output relay modules was redesigned to unify the relay specification to improve the versatility, and also to improve the pin's contact by means of welding them directly on the printed circuit boards and to make the modules detachable by connectors instead of previously non-detachable. Th modules were connected in series by both power supply line and ground line which were finally connected at same point respectively, so that other protection signals can still be output correctly when a single module is removed. (2) The relay drive circuit was also redesigned for working in on-off state instead of in amplification to minimize the power consumption. On the other hand, the CMOS buffers were taken to couple the CMOS circuits to the TTL circuits. The actuating time for the new shutdown relay was decreased from the former 35 ms to 5 ms, the actuating time for the engineered safety feature drive signal relay was decreased from 10 ms to 6 ms after the above-mentioned improvements, the reliability of the RPS is remarkably improved and a great economic benefit is obtained. (4 refs., 3 figs., 2 tabs.)

  12. Evaluation of the physical protection system of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Vaz, Antonio C.A.; Conti, Thadeu das N.

    2013-01-01

    The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

  13. Evaluation of the physical protection system of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vaz, Antonio C.A.; Conti, Thadeu das N., E-mail: acavaz@ipen.br, E-mail: tnconti@yahoo.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    The '09/11' in New York and the accident at the Fukushima power plant are two events that served as worldwide reference to review some aspects of the Physical Protection System (PPS) in nuclear areas. The nuclear research reactor IEA-R1 has followed this new world order and improved the protection systems that are directly related to detection (CCTV, sensors, alarms, etc), delay (turnstile, gates, barriers, etc) and response (communication systems, response force, etc), for operation against malicious act, seeking always to avoid or minimize any possibility of threat, theft and sabotage. These actions were performed to prevent and to mitigate the consequence on the environment, economy and society from damages caused by natural hazard, as well. This study evaluates the PPS of the IEA-R1 regarding the weaknesses, strengths,and impacts of the changes resulting from the system implanted. The analyses were based on methodology developed by security experts from SANDIA National Laboratories in Texas - U.S.A, allowing the evaluation of the system through probabilistic and hypothetical analysis. (author)

  14. On implementation of the self-protection principle to the reactors with fast-resonance neutron spectrum

    International Nuclear Information System (INIS)

    Kuznetsov, V.V.; Morozov, A.G.; Slesarev, I.S.; Alekseev, P.N.; Zverkov, Yu.A.; Subbotin, S.A.

    1990-01-01

    The calculational substantiation of SWPR posessing inherent physical properties of self-protection against possible accidents is given. A variety of approaches to the layout of these reactors have been found, the possible level of their fuel utilization characteristics is analyzed. 12 refs.; 14 figs.; 6 tabs

  15. Software Unit Testing during the Development of Digital Reactor Protection System of HTR-PM

    International Nuclear Information System (INIS)

    Guo Chao; Xiong Huasheng; Li Duo; Zhou Shuqiao; Li Jianghai

    2014-01-01

    Reactor Protection System (RPS) of High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) is the first digital RPS designed and to be operated in the Nuclear Power Plant (NPP) of China, and its development process has receives a lot of concerns around the world. As a 1E-level safety system, the RPS has to be designed and developed following a series of nuclear laws and technical disciplines including software verification and validation (software V&V). Software V&V process demonstrates whether all stages during the software development are performed correctly, completely, accurately, and consistently, and the results of each stage are testable. Software testing is one of the most significant and time-consuming effort during software V&V. In this paper, we give a comprehensive introduction to the software unit testing during the development of RPS in HTR-PM. We first introduce the objective of the testing for our project in the aspects of static testing, black-box testing, and white-box testing. Then the testing techniques, including static testing and dynamic testing, are explained, and the testing strategy we employed is also introduced. We then introduce the principles of three kinds of coverage criteria we used including statement coverage, branch coverage, and the modified condition/decision coverage. As a 1E-level safety software, testing coverage needs to be up to 100% mandatorily. Then we talk the details of safety software testing during software development in HTR-PM, including the organization, methods and tools, testing stages, and testing report. The test result and experiences are shared and finally we draw a conclusion for the unit testing process. The introduction of this paper can contribute to improve the process of unit testing and software development for other digital instrumentation and control systems in NPPs. (author)

  16. Development of an Advanced Digital Reactor Protection System Using Diverse Dual Processors to Prevent Common-Mode Failure

    International Nuclear Information System (INIS)

    Shin, Hyun Kook; Nam, Sang Ku; Sohn, Se Do; Chang, Hoon Seon

    2003-01-01

    The advanced digital reactor protection system (ADRPS) with diverse dual processors has been developed to prevent common-mode failure (CMF). The principle of diversity is applied to both hardware design and software design. For hardware diversity, two different types of CPUs are used for the bistable processor and local coincidence logic (LCL) processor. The Versa Module Eurocard-based single board computers are used for the CPU hardware platforms. The QNX operating system and the VxWorks operating system were selected for software diversity. Functional diversity is also applied to the input and output modules, and to the algorithm in the bistable processors and LCL processors. The characteristics of the newly developed digital protection system are described together with the preventive capability against CMF. Also, system reliability analysis is discussed. The evaluation results show that the ADRPS has a good preventive capability against the CMF and is a highly reliable reactor protection system

  17. Development of monitoring, control and protection systems of nuclear power reactors in the USSR: Status and trends

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.; Mikhailov, M.N.; Prozorov, V.K.; Chudin, A.G.

    1992-01-01

    In 1989-90 the Supervision Body approved the new safety regulations for nuclear power plants. The operating plants do not always completely satisfy them and a great amount of work to develop operating power units up to the conditions required is necessary. This paper describes briefly the main changes made in monitoring, control and protection systems of nuclear power reactors to increase the reactor safety. The following fields are presented in the paper: General status of the NPP control and safety systems and instrumentation in USSR; sensors; electronic equipment; actuators improvement; qualification tests. (author). 3 figs

  18. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR and PP)

    International Nuclear Information System (INIS)

    Moses, David Lewis

    2011-01-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR and PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR and PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR and PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR and PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet

  19. Trans-Uranium Doping Utilization for Increasing Protected Plutonium Proliferation of Small Long Life Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suud, Zaki [Nuclear and Biophysics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Suzuki, Mitsutoshi [Japan Atomic Energy Agency, Nuclear Non-proliferation Science and Technology Center, 2-4 Shirane Shirakata, Tokai-mura, Ibaraki, 319-1195 (Japan)

    2009-06-15

    Scientific approaches are performed by adopting some methodologies in order to increase a material 'barrier' in plutonium isotope composition by increasing the even mass number of plutonium isotope such as Pu-238, Pu-240 and Pu-242. Higher difficulties (barrier) or more complex requirement for peaceful use of nuclear materials, material fabrication and handling and isotopic enrichment can be achieved by a higher isotopic barrier. Higher barrier which related to intrinsic properties of plutonium isotopes with even mass number (Pu-238, Pu-240 and Pu-242), in regard to their intense decay heat (DH) and high spontaneous fission neutron (SFN) rates were used as a parameter for improving the proliferation resistance of plutonium itself. Pu-238 has relatively high intrinsic characteristics of DH (567 W/kg) and SFN rate of 2660 n/g/s can be used for making a plutonium characteristics analysis. Similar characteristics with Pu-238, other even mass number of plutonium isotopes such as Pu-240 and Pu-242 have been shown in regard to SFN values. Those even number mass of plutonium isotope contribute to some criteria of plutonium characterization which will be adopted for present study such as IAEA, Pellaud and Kessler criteria (IAEA, 1972; Pellaud, 2002; and Kessler, 2004). The study intends to evaluate the trans-uranium doping effect for increasing protected plutonium proliferation in long-life small reactors. The development of small and medium reactor (SMR) is one of the options which have been adopted by IAEA as future utilization of nuclear energy especially for less developed countries (Kuznetsov, 2008). The preferable feature for small reactors (SMR) is long life operation time without on-site refueling and in the same time, it includes high proliferation resistance feature. The reactor uses MOX fuel as driver fuel for two different core types (inner and outer core) with blanket fuel arrangement. Several trans-uranium doping and some doping rates are evaluated

  20. A RESEARCH ON SEAMLESS PLATFORM CHANGE OF REACTOR PROTECTION SYSTEM FROM PLC TO FPGA

    Directory of Open Access Journals (Sweden)

    JUNBEOM YOO

    2013-08-01

    Full Text Available The PLC (Programmable Logic Controller has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems. Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array. Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

  1. 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)

  2. Generic assessment procedures for determining protective actions during a reactor accident

    International Nuclear Information System (INIS)

    1997-08-01

    This manual provides the tools, procedures and data needed to evaluate the consequences of a nuclear accident occurring at a nuclear power plant throughout all phases of the emergency before, during and after a release of radioactive material. It is intended for use by on-site and off-site groups responsible for evaluating the accident consequences and making recommendations for the protection of the plant personnel, the emergency workers and the public. The scope of this manual is restricted to the technical assessment of radiological consequences. It does not address the emergency response infrastructure requirements, nor does it cover the emergency management aspects of accident assessment (e.g. reporting, staff qualification, shift replacement, and procedure implementation). The procedures and methods in this manual were developed based on a number of assumptions concerning the design and operation of the nuclear power plant and national practices. Therefore, this manual must be reviewed as part of the planning process to match the potential accidents, local conditions, national criteria and other unique characteristics of an area or nuclear reactor where it may be used. Refs, figs, tabs

  3. Irradiation of quench protection diodes at cryogenic temperatures in a nuclear research reactor

    International Nuclear Information System (INIS)

    Hagedorn, D.; Schoenbacher, H.; Gerstenberg, H.

    1996-01-01

    Within the framework of the Large Hadron Collider (LHC) R ampersand D programme, CERN and the Department of Physics E21 of the Technical University Munich have established a collaboration to carry out irradiation experiments at liquid helium and liquid nitrogen temperatures on epitaxial diodes for the superconducting magnet protection. Small diode samples of 10 mm wafer diameter from two different manufacturers were submitted to doses of up 50 kGy and neutron fluences up to 1015 n/cm 2 and the degradation of the electrical characteristics was measured versus dose. During irradiation the diodes were submitted to current pulse annealing and after irradiation to thermal annealing. After exposure some diodes show a degradation in forward voltage drop of up to 600 % which, however, can be reduced to about 15 % - 20 % by thermal annealing. The degradation at liquid helium temperature is very similar to the degradation at liquid nitrogen temperature. These degradations of electrical characteristics during the short term irradiation in a nuclear reactor are compared with degradations during long term irradiation in an accelerator environment at liquid nitrogen temperature

  4. A Research on Seamless Platform Change of Reactor Protection System From PLC to FPGA

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Junbeom; Lee, Jonghoon [Konkuk Univ., Seoul (Korea, Republic of); Lee, Jangsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-08-15

    The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea.

  5. A Research on Seamless Platform Change of Reactor Protection System From PLC to FPGA

    International Nuclear Information System (INIS)

    Yoo, Junbeom; Lee, Jonghoon; Lee, Jangsoo

    2013-01-01

    The PLC (Programmable Logic Controller) has been widely used to implement real-time controllers in nuclear RPSs (Reactor Protection Systems). Increasing complexity and maintenance cost, however, are now demanding more powerful and cost-effective implementation such as FPGA (Field-Programmable Gate Array). Abandoning all experience and knowledge accumulated over the decades and starting an all-new development approach is too risky for such safety-critical systems. This paper proposes an RPS software development process with a platform change from PLC to FPGA, while retaining all outputs from the established development. This paper transforms FBD designs of the PLC-based software development into a behaviorally-equivalent Verilog program, which is a starting point of a typical FPGA-based hardware development. We expect that the proposed software development process can bridge the gap between two software developing approaches with different platforms, such as PLC and FPGA. This paper also demonstrates its effectiveness using an example of a prototype version of a real-world RPS in Korea

  6. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors

    International Nuclear Information System (INIS)

    Alvarez G, G.

    1991-01-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  7. A Behavior-Preserving Translation From FBD Design to C Implementation for Reactor Protection System Software

    International Nuclear Information System (INIS)

    Yoo, Junbeom; Kim, Euisub; Lee, Jangsoo

    2013-01-01

    Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation

  8. A Behavior-Preserving Translation From FBD Design to C Implementation for Reactor Protection System Software

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Junbeom; Kim, Euisub [Konkuk Univ., Seoul (Korea, Republic of); Lee, Jangsoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-08-15

    Software safety for nuclear reactor protection systems (RPSs) is the most important requirement for the obtainment of permission for operation and export from government authorities, which is why it should be managed with well-experienced software development processes. The RPS software is typically modeled with function block diagrams (FBDs) in the design phase, and then mechanically translated into C programs in the implementation phase, which is finally compiled into executable machine codes and loaded on RPS hardware - PLC (Programmable Logic Controller). Whereas C Compilers are fully-verified COTS (Commercial Off-The-Shelf) software, translators from FBDs to C programs are provided by PLC vendors. Long-term experience, experiments and simulations have validated their correctness and function safety. This paper proposes a behavior-preserving translation from FBD design to C implementation for RPS software. It includes two sets of translation algorithms and rules as well as a prototype translator. We used an example of RPS software in a Korean nuclear power plant to demonstrate the correctness and effectiveness of the proposed translation.

  9. Formal verification and validation of the safety-critical software in a digital reactor protection system

    International Nuclear Information System (INIS)

    Kwon, K. C.; Park, G. Y.

    2006-01-01

    This paper describes the Verification and Validation (V and V) activities for the safety-critical software in a Digital Reactor Protection System (DRPS) that is being developed through the Korea nuclear instrumentation and control system project. The main activities of the DRPS V and V process are a preparation of the software planning documentation, a verification of the software according to the software life cycle, a software safety analysis and a software configuration management. The verification works for the Software Requirement Specification (SRS) of the DRPS consist of a technical evaluation, a licensing suitability evaluation, a inspection and traceability analysis, a formal verification, and preparing a test plan and procedure. Especially, the SRS is specified by the formal specification method in the development phase, and the formal SRS is verified by a formal verification method. Through these activities, we believe we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the nuclear safety-critical software in a DRPS. (authors)

  10. Study on human-factors-engineering properties of reactor maintenance workers with protection suits, (2). Basic research on various biological characteristics in reactor maintenance simulation tests

    Energy Technology Data Exchange (ETDEWEB)

    Yoshino, K; Ishii, K; Nakasa, H [Central Research Inst. of Electric Power Industry, Tokyo (Japan); Shigeta, S

    1980-11-01

    To ensure the safety of reactor maintenance workers and to reduce the radiation exposure through the enhancement of labor efficiency, it is needed to evaluate quantitatively work-stress levels of workers with radiation-protection suits. This paper presents the results of reactor-maintenance simulation tests in which the relationship between the work stress and biological characteristics is investigated for 5 pinds of model works done by testees without protection suits in an artificial climate chamber. Major results obtained are: (1) the selected model works are mostly evaluated to be relatively heavy through the measurement of RMR (Relative Metabolic Rate). (2) biological characteristics such as heart rate and respiratory volume under the model works have close relationship to RMR which is the cumulative quantity in relatively long time, and then they may become the real-time indicator for the work stress level. (3) such biological characteristics are greatly affected by the high-temperature work-environment which is often seen in workers with protection suits.

  11. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  12. RA Research nuclear reactor, Part II - radiation protection at the RA nuclear reactor in 1984; Istrazivacki nuklearni reaktor RA - Deo II - zastita od zracenja kod nuklearnog reaktora RA u 1984. godini

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M; Ajdacic, N; Zaric, M; Vukovic, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-12-15

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

  13. Preliminary Validation and Verification Plan for CAREM Reactor Protection System; Modelo de Plan Preliminar de Validacion y Verificacion para el Sistema de Proteccion del Reactor CAREM

    Energy Technology Data Exchange (ETDEWEB)

    Fittipaldi, Ana; Felix, Maciel [Comision Nacional de Energia Atomica, Centro Atomico Bariloche (Argentina)

    2000-07-01

    The purpose of this paper, is to present a preliminary validation and verification plan for a particular architecture proposed for the CAREM reactor protection system with software modules (computer based system).These software modules can be either own design systems or systems based in commercial modules such as programmable logic controllers (PLC) redundant of last generation.During this study, it was seen that this plan can also be used as a validation and verification plan of commercial products (COTS, commercial off the shelf) and/or smart transmitters.The software life cycle proposed and its features are presented, and also the advantages of the preliminary validation and verification plan.

  14. Pilot program: NRC severe reactor accident incident response training manual: Public protective actions: Predetermined criteria and initial actions

    International Nuclear Information System (INIS)

    Martin, J.A. Jr.; McKenna, T.J.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

    1987-02-01

    This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Public Protective Actions - Predetermined Criteria and Initial Actions is the fourth in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume reviews public protective action criteria and objectives, their bases and implementation, and the expected public response. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

  15. Radiation protection at the RA Reactor in 1988, Part -2, Annex 2c, Environmental radioactivity control, meteorology measurements

    International Nuclear Information System (INIS)

    Grsic, Z.; Zaric, M.; Adamovic, M.; Stevanovic, M.

    1988-01-01

    During 1988, meteorology measurements were continued as a part of the environmental control of the Vinca Institute. This report covers the period from November 1984 - November 1985. Part of the meteorology measurements and data analysis is adapted to the needs of the Institute, i.e. RA reactor and some Laboratories. The objective of these activities is forming the data base for solving everyday and special problems related to control, protection and safety of Institute environment [sr

  16. Case study on the use of PSA methods: Assessment of technical specifications for the reactor protection system instrumentation

    International Nuclear Information System (INIS)

    1992-10-01

    This case study presents a methodology for the probabilistic evaluation of alternative plant technical specifications regarding system surveillance frequencies and out-of-service times. The methodology is applied to the reactor protection systems of a 4 loop BWR-RESAR-3S type nuclear power plant. The effect of the statistical characteristics of the system on the relative comparison of various sets of technical specifications is examined through sensitivity studies and an uncertainty analysis. Refs, figs and tabs

  17. Design of reactor protection systems for HTR plants generating electric power and process heat problems and solutions

    International Nuclear Information System (INIS)

    Craemer, B.; Dahm, H.; Spillekothen, H.G.

    1982-06-01

    The design basis of the reactor protection system (RPS) for HTR plants generating process heat and electric power is briefly described and some particularities of process heat plants are indicated. Some particularly important or exacting technical measuring positions for the RPS of a process heat HTR with 500 MWsub(th) power (PNP 500) are described and current R + D work explained. It is demonstrated that a particularly simple RPS can be realized in an HTR with modular design. (author)

  18. High-temperature gas-cooled reactor steam-cycle/cogeneration lead plant. Plant Protection and Instrumentation System design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Plant Protection and Instrumentation System provides plant safety system sense and command features, actuation of plant safety system execute features, preventive features which maintain safety system integrity, and safety-related instrumentation which monitors the plant and its safety systems. The primary function of the Plant Protection and Instrumentation system is to sense plant process variables to detect abnormal plant conditions and to provide input to actuation devices directly controlling equipment required to mitigate the consequences of design basis events to protect the public health and safety. The secondary functions of the Plant Protection and Instrumentation System are to provide plant preventive features, sybsystems that monitor plant safety systems status, subsystems that monitor the plant under normal operating and accident conditions, safety-related controls which allow control of reactor shutdown and cooling from a remote shutdown area

  19. A PLC platform-independent structural analysis on FBD programs for digital reactor protection systems

    International Nuclear Information System (INIS)

    Jung, Sejin; Yoo, Junbeom; Lee, Young-Jun

    2017-01-01

    Highlights: • FBD has been widely used to implement safety-critical software for PLC-based systems. • The safety-critical software should be developed strictly with safety programming guidelines. • There are no argued rules that have specific links to higher guidelines NUREG/CR-6463 PLC platform-independently. • This paper proposes a set of rules on the structure of FBD programs with providing specific links to higher guidelines. • This paper also provides CASE tool ‘FBD Checker’ for analyzing the structure of FBD. - Abstract: FBD (function block diagram) has been widely used to implement safety-critical software for PLC (programmable logic controller)-based digital nuclear reactor protection systems. The software should be developed strictly in accordance with safety programming guidelines such as NUREG/CR-6463. Software engineering tools of PLC vendors enable us to present structural analyses using FBD programs, but specific rules pertaining to the guidelines are enclosed within the commercial tools, and specific links to the guidelines are not clearly communicated. This paper proposes a set of rules on the structure of FBD programs in accordance with guidelines, and we develop an automatic analysis tool for FBD programs written in the PLCopen TC6 format. With the proposed tool, any FBD program that is transformed into an open format can be analyzed the PLC platform-independently. We consider a case study on FBD programs obtained from a preliminary version of a Korean nuclear power plant, and we demonstrate the effectiveness and potential of the proposed rules and analysis tool.

  20. Very High-Temperature Reactor (VHTR) Proliferation Resistance and Physical Protection (PR&PP)

    Energy Technology Data Exchange (ETDEWEB)

    Moses, David Lewis [ORNL

    2011-10-01

    This report documents the detailed background information that has been compiled to support the preparation of a much shorter white paper on the design features and fuel cycles of Very High-Temperature Reactors (VHTRs), including the proposed Next-Generation Nuclear Plant (NGNP), to identify the important proliferation resistance and physical protection (PR&PP) aspects of the proposed concepts. The shorter white paper derived from the information in this report was prepared for the Department of Energy Office of Nuclear Science and Technology for the Generation IV International Forum (GIF) VHTR Systems Steering Committee (SSC) as input to the GIF Proliferation Resistance and Physical Protection Working Group (PR&PPWG) (http://www.gen-4.org/Technology/horizontal/proliferation.htm). The short white paper was edited by the GIF VHTR SCC to address their concerns and thus may differ from the information presented in this supporting report. The GIF PR&PPWG will use the derived white paper based on this report along with other white papers on the six alternative Generation IV design concepts (http://www.gen-4.org/Technology/systems/index.htm) to employ an evaluation methodology that can be applied and will evolve from the earliest stages of design. This methodology will guide system designers, program policy makers, and external stakeholders in evaluating the response of each system, to determine each system's resistance to proliferation threats and robustness against sabotage and terrorism threats, and thereby guide future international cooperation on ensuring safeguards in the deployment of the Generation IV systems. The format and content of this report is that specified in a template prepared by the GIF PR&PPWG. Other than the level of detail, the key exception to the specified template format is the addition of Appendix C to document the history and status of coated-particle fuel reprocessing technologies, which fuel reprocessing technologies have yet to be

  1. Nuclear Reactor RA Safety Report, Vol. 9, Radiation protection; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 9, Zastita od zracenja

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    Instrumentation for Radiation protection existing at the RA reactor is dating mostly from the period 1957-1959 when the reactor has been built. With some minor exception it was produced in USSR. Radiation protection system was constructed based on specific design project, somewhat modified original USSR project which has been indispensable because of some modification of the building design. During the past 27 years no renewal of the instrumentation was done, only maintenance was performed. Instrumentation consists of old electronic devices which caused difficulties and even prevented regular maintenance because of lack of spare parts. Instrumentation for radiation protection at the RA reactor is classified as follows: centralized dosimetry system; stationary dosimetry instrumentation, movable and personal dosimetry systems. Apart from the scheme of dosimetry instrumentation this volume includes description of radiation protection procedures; protection devices; radiation doses and dose limit data; program for environmental radioactivity control; medical control procedures. [Serbo-Croat] Instrumentacija za zastitu od zracenja koja danas postoji na reaktoru RA najvecim delom potice iz perioda 1957-1959 kada je reaktor gradjen. Sa malim izuzetcima instrumentacija je proizvedena u SSSR. Sistem za zastitu izveden je na osnovu posebnog projekta koji predstavlja modifikaciju originalnog projekta koja je bila neophodna usled modifikovanog gradjevinskog projekta. U proteklom periodu od 27 godina instrumentacija nije obnavljana vec je vrseno samo odrzavanje. Instrumentacija je izradjena u danas prevazidjenoh tehnoligiji (elektronske cevi) sto je otezavalo i skoro onemogucavalo normalno odrzavanje. Instrumentacija za zastitu od zracenja na reaktoru RA moze se podeliti na tri dela: centralizovani dozimetrijski sistem; stacionarna dozimetrijska instrumentacija; prenosna i licna dozimetrijska instrumentacija. Pored seme sistema dozimetrijske kontrole ova knjiga sadrzi opis

  2. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  3. Operational experiences in radiation protection in fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    Meenakshisundaram, V.; Rajagopal, V.; Santhanam, R.; Baskar, S.; Madhusoodanan, U.; Chandrasekaran, S.; Balasundar, S.; Suresh, K.; Ajoy, K.C.; Dhanasekaran, A.; Akila, R.; Indira, R.

    2008-01-01

    The Compact Reprocessing facility for Advanced fuels in Lead cells (CORAL), situated at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam is a pilot plant to reprocess the mixed carbide fuel, for the first time in the world. Reprocessing of fuel with varying burn-ups up to 155 G Wd/t, irradiated at Fast Breeder Test Reactor (FBTR), has been successfully carried out at CORAL. Providing radiological surveillance in a fuel reprocessing facility itself is a challenging task, considering the dynamic status of the sources and the proximity of the operator with the radioactive material and it is more so in a fast reactor fuel reprocessing facility due to handling of higher burn-up fuels associated with radiation fields and elevated levels of fissile material content from the point of view of criticality hazard. A very detailed radiation protection program is in place at CORAL. This includes, among others, monitoring the release of 85 Kr and other fission products and actinides, if any, through stack on a continuous basis to comply with the regulatory limits and management of disposal of different types of radioactive wastes. Providing radiological surveillance during the operations such as fuel transport, chopping and dissolution and extraction cycle was without any major difficulty, as these were carried out in well-shielded and high integrity lead cells. Enforcement of exposure control assumes more importance during the analysis of process samples and re-conversion operations due to the presence of fission product impurities and also since the operations were done in glove boxes and fume hoods. Although the radiation fields encountered in process area were marginally higher, due to the enforcement of strict administrative controls, the annual exposure to the radiation workers was well within the regulatory limit. As the facility is being used as test bed for validation of prototype equipment, periodic inspection and maintenance of components such as centrifuge

  4. Solution to the incompatibility between reactor protection logic and turbine shot logic. Scram by high pressure; Solucion a la incompatibilidad entre logica de proteccion de reactor y logica de disparo de turbina. SCRAM por alta presion

    Energy Technology Data Exchange (ETDEWEB)

    Ramos Q, R.; Santiago F, C.; Gonzalez P, G., E-mail: ruben.ramos01@cfe.gob.mx [Comision Federal de Electricidad, Central Nuclear Laguna Verde, Subgerencia de Ingenieria, Carretera Cardel-Nautla Km. 42.5, Alto Lucero, Veracruz (Mexico)

    2013-10-15

    The nuclear power plant of Laguna Verde carried out the Modernization and Increase of Extended Power Project in its two Units (2005-2011). This modernization included to the electro-hydraulic control system of the main turbine, replacing an ana logical system by one digital (Digital Electro-hydraulic Control - DEHC) whose functions are of controlling the reactor pressure in the different operation ways as wells as of controlling the velocity and load of the main turbine. Also, it has protections that are related with diverse plant systems, as the Reactor Protection Systems (RPS). During the tests stage was realized a programmed load rejection, which Reactor Scram should cause when being presented the shot of main turbine. However, the logic of the RPS was inhibited due to the quick response of the new control DEHC, propitiating a condition of non prospective plant and, in consequence, the Reactor Scram happened for another protection of the RPS. (author)

  5. Multiple implementation of a reactor protection code in PHI2, PASCAL, and IFTRAN on the SIEMENS-330 computer

    International Nuclear Information System (INIS)

    Gmeiner, L.; Lemperle, W.; Voges, U.

    1978-01-01

    In safety related computer applications, as in the case of a reactor protection system considered here, mostly multi-computer systems are necessary for reasons of reliability and availability. The hardware structure of the protection system and the software requierements derived from it are explained. In order to study the effects of diversified programming of the three computers the protection codes were implemented in the languages IFTRAN, PASCAL, and PHI2. According to the experience gained diversified programming seems to be a proper means to prevent identical programming errors in all three computers on one hand and to detect ambiguities of the specification on the other. During all of the experiment the errors occurring were recorded in detail and at the moment are being evaluated. (orig./WB) [de

  6. Aircraft-crash-protected steel reactor building roof structure for the European market

    International Nuclear Information System (INIS)

    Posta, B.A.; Kadar, I.; Rao, A.S.

    1996-01-01

    This paper recommends the use of all steel roof structures for the reactor building of European Boiling Water Reactor (BWR) plants. This change would make the advanced US BWR designs more compatible with European requirements. Replacement of the existing concrete roof slab with a sufficiently thick steel plate would eliminate the concrete spelling resulting from a postulated aircraft crash, potentially damaging the drywell head or the spent fuel pool

  7. Programming of computers for the protection system for Savannah River reactors

    International Nuclear Information System (INIS)

    Finley, R.H.

    1977-06-01

    The monitoring requirements for the SRP Safety Computers are shown. These fast response times coupled with the large number of analog inputs to be scanned imposed stringent program requirements. The system consists of two separate computers, each with its own inputs to monitor half the reactor positions. Either computer can provide the minimum required monitoring. The desired redundant monitoring is provided when both computers are on-line. If both computers are off-line, the reactor is automatically shut down

  8. Plant protection system optimization studies to mitigate consequences of large breaks in the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Khayat, M.I.; March-Leuba, J.

    1993-01-01

    This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the advanced neutron source (ANS) reactor. The ANS is a new basic and applied research facility based on a powerful steady-state research reactor that provides beams of neutrons for measurements and experiments in the field of material science and engineering, biology, chemistry, material analysis, and nuclear science. To achieve the high neutron fluxes for these state-of-the-art experiments, the ANS design has a very high power density core (330 MW fission with an active volume of 67.6 ell) surrounded by a large heavy-water reflector, where most neutrons are moderated. This design maximizes the number of neutrons available for experiments but results in a low heat capacity core that creates unique challenges to the design of the plant protection system

  9. Protection set-points lines for the reactor core and considerations about power distribution and peak factors

    International Nuclear Information System (INIS)

    Furieri, E.B.

    1981-01-01

    In order to assure the reactor core integrity during the slow operational transients (power excursion above the nominal value and the high coolant temperature), the formation of a steam film (DNB-Departure from Nucleate Boiling) in the control rods must be avoided. The protection set points lines presents the points where DNBR (relation between critical heat flux-q sub(DNB) and the local heat flux-q' sub(local) is equal to 1.30, corrected by peak factors and uncertainty in function of ΔTr and T sub(R), respectively coolant elevation and medium coolant temperature in reactor pressure vessel. The curve set-points were determined using a new version of COBRA-IIIF (CUPRO) computer code, implemented with new subroutines and linearized convergence scheme. Pratical results for Angra-1 core were obtained and its were compared with the results from the fabricator. (E.G.) [pt

  10. Reactor accidents and how to protect oneself from them. Emergency measures. Schutz bei Atomunfaellen. Vorbereitet sein auf den Notfall

    Energy Technology Data Exchange (ETDEWEB)

    Gerosa, K

    1986-01-01

    The Chernobyl reactor accident has sharpened our sense of nuclear energy risks and dangers. Much as we hope it to never occur again in the future we know that there is no way of guaranteeing this to be the case. What is to be done in case of another reactor accident with radioactive radiation threatening to destroy us and the environment. Emergency measures protecting and saving our lives and health cannot be taken but with sufficient information at hand. The book informs about measures to be taken under emergency conditions. Nuclear energy and its alternatives, preventive measures, adequate nutrition and advice for pregnant women and for children are among the further subjects dealt with.

  11. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  12. Note on protection offered by ventilation systems in the event of a nuclear reactor accident

    International Nuclear Information System (INIS)

    Wayland, J.R.; McGrath, P.E.

    1976-05-01

    A brief review of the protection offered by natural and forced ventilation systems in buildings to an atmospheric release of radioactive material is given. The protection to be gained by using the internal ventilation system is estimated

  13. Automation of the radiation protection monitoring system in the RP-10 reactor

    International Nuclear Information System (INIS)

    Anaya G, Olgger; Castillo Y, Walter; Ovalle S, Edgar

    2002-01-01

    During the reactor operation, it is necessary to carry out the radiological control in the different places of the reactor, in periodic form and to take a registration of these values. For it the radioprotection official, makes every certain periods, settled down in the procedures, to verify and to carry out the registration of those values in manual form of each one of the radiation monitors. For this reason it was carried out the design and implementation of an automatic monitoring system of radioprotection in the reactor. In the development it has been considered the installation of a acquisition data system for 27 radiation gamma monitors of the type Geiger Mueller, installed inside the different places of the reactor and in the laboratories where they are manipulated radioactive material, using as hardware the FieldPoint for the possessing and digitalization of the signs which are correspondents using the communication protocol RS-232 to a PC in which has settled a program in graphic environment that has been developed using the tools of the programming software LabWindows/CVI. Then, these same signs are sent 'on line' to another PC that is in the Emergency Center of Coordination to 500 m of the reactor, by means of a system of radiofrequency communication. (author)

  14. Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

    International Nuclear Information System (INIS)

    Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

    2009-10-01

    The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and foreseen by for instance the

  15. Evaluation Of Fire Safety And Protection At PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ahmad Nabil Ab Rahim; Alfred Sanggau Ligam; Nurhayati Ramli; Mohd Fazli Zakaria; Naim Syauqi Hamzah; Phongsakorn Prak; Mohammad Suhaimi Kassim; Zarina Masood

    2014-01-01

    Fire hazard is one of many risks that can affect the safety operation of PUSPATI TRIGA Reactor. Reactor building in Malaysian Nuclear Agency was built in 1980s and the fire system has been introduced since then. The evaluation of the fire safety system at this time is important to ensure the efficiency of fire prevention, fighting and mitigation task that probably occurs. This evaluation involves with the fire fighting system and equipment, integrity of the system from the perspective of management and equipment, fire fighting procedure and fire fighting response team. (author)

  16. Method and device for the passive protection of a nuclear reactor

    International Nuclear Information System (INIS)

    Cachera, P.C.

    1976-01-01

    Conventional fuel elements within the core of a nuclear reactor and especially a fast reactor are at least partly replaced by ''safety elements'' each formed by a stack of fissile fuel pellets enclosed in a can. Each pellet is provided with a central orifice so as to form an axial flow duct of sufficiently large cross-sectional area to ensure that the portion of fuel which is liable to melt as a result of a neutron-flux excursion flows under gravity to the bottom of the fuel element and has the effect of reducing the reactivity without damaging the fuel can

  17. Protection walls and other means used in everyday work on the Vinca RA Reactor

    International Nuclear Information System (INIS)

    Milosevic, M.; Ninkovic, M.

    1964-10-01

    Work with radioactive materials requires special protection of the personnel. Special attention has been paid to this problem because the time allowed for work on a problem depends on the protection provided. The paper gives a short review of the means and methods of protection against irradiation and contamination, it also describes some personal and technical protection means used in specific working conditions. A special description is given of the technical means of radiation protection (protection against free beams): heavy bricks (iron and sand), water and iron shields, plugs for beam cutting. Experimental data on the efficiency of these means in moderating the radiation by gamma rays and thermal neutrons are given. (All measurements of the efficiency of the protection means have been carried out under the real conditions, that is to say conditions under which these measurements are usually made, so the data obtained completely respond to dosimetry demands) (author)

  18. Software verification and validation methodology for advanced digital reactor protection system using diverse dual processors to prevent common mode failure

    International Nuclear Information System (INIS)

    Son, Ki Chang; Shin, Hyun Kook; Lee, Nam Hoon; Baek, Seung Min; Kim, Hang Bae

    2001-01-01

    The Advanced Digital Reactor Protection System (ADRPS) with diverse dual processors is being developed by the National Research Lab of KOPEC for ADRPS development. One of the ADRPS goals is to develop digital Plant Protection System (PPS) free of Common Mode Failure (CMF). To prevent CMF, the principle of diversity is applied to both hardware design and software design. For the hardware diversity, two different types of CPUs are used for Bistable Processor and Local Coincidence Logic Processor. The VME based Single Board Computers (SBC) are used for the CPU hardware platforms. The QNX Operating System (OS) and the VxWorks OS are used for software diversity. Rigorous Software Verification and Validation (V and V) is also required to prevent CMF. In this paper, software V and V methodology for the ADRPS is described to enhance the ADRPS software reliability and to assure high quality of the ADRPS software

  19. A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods

    International Nuclear Information System (INIS)

    Benko, Pedro Luiz

    1997-01-01

    A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

  20. 10 CFR 140.12 - Amount of financial protection required for other reactors.

    Science.gov (United States)

    2010-01-01

    .... Determine the population of each such minor civil division (according to the same census or later data... of the estimated distance to the nearest mile from the reactor to the geographic center of the minor... pursuant to the formula and other provisions of this section: Provided, That in no event shall the amount...

  1. Reactor protection system software test-case selection based on input-profile considering concurrent events and uncertainties

    International Nuclear Information System (INIS)

    Khalaquzzaman, M.; Lee, Seung Jun; Cho, Jaehyun; Jung, Wondea

    2016-01-01

    Recently, the input-profile-based testing for safety critical software has been proposed for determining the number of test cases and quantifying the failure probability of the software. Input-profile of a reactor protection system (RPS) software is the input which causes activation of the system for emergency shutdown of a reactor. This paper presents a method to determine the input-profile of a RPS software which considers concurrent events/transients. A deviation of a process parameter value begins through an event and increases owing to the concurrent multi-events depending on the correlation of process parameters and severity of incidents. A case of reactor trip caused by feedwater loss and main steam line break is simulated and analyzed to determine the RPS software input-profile and estimate the number of test cases. The different sizes of the main steam line breaks (e.g., small, medium, large break) with total loss of feedwater supply are considered in constructing the input-profile. The uncertainties of the simulation related to the input-profile-based software testing are also included. Our study is expected to provide an option to determine test cases and quantification of RPS software failure probability. (author)

  2. Fire protection at the Fast Flux Test Facility (a sodium cooled test reactor)

    International Nuclear Information System (INIS)

    Bell, J.R.

    1980-01-01

    For purposes of this presentation, fire protection at the FFTF is subdivided into two catagories; protection for non-sodium areas and protection for areas containing sodium. Fire protection systems and philosophies for non-sodium areas at the FFTF are very similar to those used at conventional power plants being constructed throughout the country. They follow, essentially, the NRC rules and guidelines and ANSI 59.4 Generic Requirements for Light Water Nuclear Power Plant Fire Protection. The FFTF with its support facilities have their own water system comprised of a looped 8'' and 10'' underground distribution system, three 1500 GPM fire pumps and three ground level storage tanks totaling 736,000 gallons with 420,000 reserved for fire protection. Fire hydrants are enclosed with hose houses outfitted for use by the Emergency Response Team (ERT). Fire prevention systems for sodium areas of the FFTF are also described

  3. Enhancing the functionality of reactor protection systems to provide diagnostic and monitoring information: The ISAT{sup TM} approach

    Energy Technology Data Exchange (ETDEWEB)

    Baldwin, J A; Rowe, B J [AEA Technology, Winfrith (United Kingdom); Jones, C D [Nuclear Electric Ltd., Kent (United Kingdom). Dungeness ` B` Power Station

    1997-12-31

    The ISAT{sup TM} architecture has been successfully implemented as the Single Channel Trip System (SCTS), part of the primary protection system of Nuclear Electric`s Dungeness `B` Advanced Gas-Cooled Reactors. The system is the first computer-based protection system licensed on a UK civil nuclear reactor. The system provides protection against single channel faults resulting in high coolant gas outlet temperature. The SCTS was designed to output data at several points in the system to an Ethernet to allow checks to be made on the operation of parts of the protection system and the system as a whole. In order to monitor the performance of this shutdown system a PC based monitoring system was developed to take input as data from the Ethernet, check its integrity and then analyze the data to provide information of the state of the system and subsystems. The SCTS monitor was basically intended to alert the operator to any fault on the safety system and indicate its source, provide a diagnosis of the cause of any trip initiated by the safety system, and log the occurrences of these incidents for later inspection. The intention was also to provide accurate real-time information on the thermocouple readings and to decrease the effort required to maintain the safety system. This paper will describe briefly the development of the ISAT{sup TM} monitoring system: how its requirements were arrived at, and how the design, code and testing were carried out to ensure approval for this application. It will then go on to report how the ISAT{sup TM} monitor has performed during its time in service: how more functionality has been added over and above its original requirements. Features of additional monitors for the SCTS and other ISAT{sup TM} systems will also be described. (author). 2 refs, 5 figs.

  4. Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Pilgrim Nuclear Power Station

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Pilgrim Nuclear Power Station. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  5. Dosimetry and technical radiation protection at the RA reactor in 1979, Report on the project 'Operation and maintenance of the RA reactor'

    International Nuclear Information System (INIS)

    Ninkovic, M. et al.

    1979-12-01

    This report include the analysis of the dosimetry and technical radiation protection results collected during 1979. The first part shows the data about the fundamental exposure to radiation and the statistical review of the total number of measurements. The following are included as well: measured values of radioactive gases and aerosol contents in air; contamination level of surfaces, clothes and uncovered parts of the staff bodies. Analysis of the personnel exposure is presented in the second part of the report. It was stated that the maximum individual external dose was 9.5 mGy, and the exposure dose was 1/5 lower than the annual dose limit for all the exposed personnel. It was estimated that that the internal contamination is negligible compared to the external contamination. This statement was based on the frequency of tasks in the contaminated zones, undertaken control and protection measures, as well as on the evaluation of the expected internal contamination. Comparative evaluation of the occupational exposures is given for the past five years. This showed that the exposure in 1979 was approximately on the same level as the mean value during past four years. The third part of the report covers the numerical data about the quantity of the collected radioactive waste, total size of contaminated and decontaminated surfaces and number of decontaminated objects. A brief analysis of the accidents occurred during this year under regular reactor operating conditions, maintenance and repair is given at the end. It was stated that there has been no accident that would cause exposure higher than the prescribed limits or any significant contamination of the working space or environment. The more detailed analyses concerned with radiation protection aspects of the irregularities noticed on the nuclear fuel elements were described in the Annex. It was concluded that the mentioned problems did not cause any leaks. This was significant for planning and manipulating the

  6. French experience in the programmed systems for nuclear reactor control and protection

    International Nuclear Information System (INIS)

    Jover, P.

    1986-03-01

    The analysis of incidents during the start-up of the first nuclear power plant 1300 MWe has made possible to obtain good performances evaluation of the two computerized control and protection systems: the protection system (SPIN) and the logic control system (CONTROBLOC). The results of this experiment have shown that the objectives have been attained [fr

  7. Radiation protection at the RA Reactor in 1994. Part 2, Annex 1, Control of the working environment, dosimetry and radiation protection

    International Nuclear Information System (INIS)

    Pavlovic, R.; Kalinic, S.

    1994-01-01

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. It is stated that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel [sr

  8. The correspondence concerning fire protection regulation for operating reactors (separation flame test of unpurified cables)

    International Nuclear Information System (INIS)

    Hasegawa, Takayasu; Miyakoshi, Hirohisa; Goto, Masami

    2013-01-01

    Nuclear power plants are taking fire protection measures taking into account past findings about the effects of fire by the demonstration test in order to maintain the safety of nuclear power plant in the event of a fire. The objective of the demonstration test described in this paper is to obtain advanced knowledge about over current fire of unqualified cable to be applied to fire protection measures. (author)

  9. Radiation protection at the RA Reactor in 1993, Part II, Decontamination and actions, collection of liquid effluents and solid radioactive waste

    International Nuclear Information System (INIS)

    Mandic, M.; Vukovic, Z.; Lazic, S.; Plecas, I.; Voko, A.

    1993-01-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

  10. Radiation protection at the RA Reactor in 1998, Part 2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste

    International Nuclear Information System (INIS)

    Mandic, M.; Vukovic, Z.; Bacic, S.; Plecas, I.

    1998-01-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

  11. Radiation protection at the RA Reactor in 1995, Part -2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste

    International Nuclear Information System (INIS)

    Mandic, M.; Vukovic, Z.; Lazic, S.; Plecas, I.; Voko, A.

    1995-01-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

  12. Radiation protection at the RA Reactor in 1989, Part -2, Decontamination, collection of treatment of fluid and solid radioactive waste, Annex 3

    International Nuclear Information System (INIS)

    Mandic, M.; Vukovic, Z.; Plecas, I.; Knezevic, Lj.; Lazic, S.; Bacic, S.

    1989-01-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [sr

  13. Taking into account radiation protection for the EPR (European pressurized water reactor) design

    International Nuclear Information System (INIS)

    Michoux, X.

    2005-01-01

    For a designer, the taking into account of radiation protection for the EPR design is based on several thrusts which concern different scopes as choice of materials, checking of design's options, layout of components and systems able to contain radioactivity in different states of operation (i.e.: pressurizer, tanks, actives systems separated from non actives systems), or the optimization of shielding according to the estimated maintenance during outage or during power operation. The EPR method used for radiation protection studies is close to the safety method (use of dose gauge, demonstration of radiation protection, works with high stake regarding the radiation protection studied in priority, parametric studies with use of one field Radiation protection...). Results of this method place EPR in a satisfactory progress compared to the best existing nuclear plants, regarding collective doses and privileging the most exposed workers. This method has also induced on the EPR Project the choice of working during power operation in order to obtain shorts outages, scrupulously respecting security rules, radiation protection and human factor. (author)

  14. Optimal Protection of Reactor Hall Under Nuclear Fuel Container Drop Using Simulation Methods

    Directory of Open Access Journals (Sweden)

    Králik Juraj

    2014-12-01

    Full Text Available This paper presents of the optimal design of the damping devices cover of reactor hall under impact of nuclear fuel container drop of type TK C30. The finite element idealization of nuclear power plant structure is used in software ANSYS. The steel pipe damper system is proposed for dissipation of the kinetic energy of the container free fall in comparison with the experimental results. The probabilistic and sensitivity analysis of the damping devices was considered on the base of the simulation methods in program AntHill using the Monte Carlo method.

  15. First-wall-coating candidates for ICF reactor chambers using dry-wall protection only

    International Nuclear Information System (INIS)

    Sink, D.A.

    1983-01-01

    Twenty pure metals were considered as potential candidates for first-wall coatings of ICF reactor chambers. Seven were found to merit further consideration based on the results of computer-code calculations of figures-of-merit. The seven are rhenium, iridium, molybdenum, chromium, tungsten, tantalum, and niobium (listed in order of decreasing values of figures-of-merit). The calculations are based on mechanical, thermal, and vacuum vaporization engineering constraints. A number of alloys of these seven metals are suggested as additional candidates

  16. Extension and update of the reliability data for fire protection equipment in German light water reactors; Ergaenzung und Aktualisierung von Zuverlaessigkeitskenngroessen fuer Brandschutzeinrichtungen in deutschen Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Forell, Burkhard; Einarsson, Svante

    2016-12-15

    Within the BMUB project 3610R01370 ''Extension and update of reliability data for fire protection equipment in German light water reactors'' plant-specific as well as generic failure rates for the technical reliability of active fire protection features in German nuclear power plants have been calculated. Based on results of previous projects, in this project observation times of components were updated and extended and additional components and functions were assessed. Now, the data evaluated results from a total of six German reference plants with seven reactor units.

  17. Research and engineering application of coordinated instrumentation control and protection technology between reactor and steam turbine generator on nuclear power plant

    International Nuclear Information System (INIS)

    Sun Xingdong

    2014-01-01

    The coordinated instrumentation control and protection technology between reactor and steam turbine generator (TG) usually is very significant and complicated for a new construction of nuclear power plant, because it carries the safety, economy and availability of nuclear power plant. Based on successful practice of a nuclear power plant, the experience on interface design and hardware architecture of coordinated instrumentation control and protection technology between reactor and steam turbine generator was abstracted and researched. In this paper, the key points and engineering experience were introduced to give the helpful instructions for the new project. (author)

  18. Development and validation of a nuclear data and calculation system for Superphenix with steel reflectors; Developpement et qualification d`un formulaire adapte a superphenix avec reflecteurs

    Energy Technology Data Exchange (ETDEWEB)

    Bosq, J Ch

    1998-11-09

    This thesis concerns the definition and the validation of the ERANOS neutronic calculation system for steel reflected fast reactors. The calculation system uses JEF2.2 evaluated nuclear data, the ECCO cell code and the BISTRO and VARIANT transport codes. After a description of the physical phenomena induced by the existence of the these sub-critical media, an inventory of the past studies related to steel reflectors is reported. A calculational scheme taking into account the important physical phenomena (strong neutronic slowing-down, presence of broad resonances of the structural materials and spatial variation of the spectrum in the reflector) is defined. This method is validated with the TRIPOLI4 reference Monte-Carlo code. The use of this upgraded calculation method for the analysis of the part of the CIRANO experimental program devoted to the study of steel reflected configurations leads to discrepancies between the calculated and measured values. These remaining discrepancies obtained for the reactivity and the fission rate traverses are due to inaccurate nuclear data for the structural materials. The adjustment of these nuclear data in order to reduce these discrepancies id demonstrated. The additional uncertainty associated to the integral parameters of interest for a nuclear reactor (reactivity and power distribution) induced by the replacement of a fertile blanket by a steel reflector is determined for the Superphenix reactor and is proved to be small. (author) 86 refs.

  19. Radiation protection at the RA Reactor in 1985, Part -2, Annex 1, Radioactivity control of working environment, dosimetry

    International Nuclear Information System (INIS)

    Ninkovic, M.; Bjelanovic, J.; Minincic, Z.; Komatina, R.; Raicevic, J.

    1985-01-01

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted to 8.2 mSV during past 10 months. Individual exposures for 7/10 of the personnel were less than 1/10 of the annual permissible exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. The last part analyzes accidents occurred at the reactor during 1985. It was found that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel [sr

  20. Reorganization of the radiologic protection of the nuclear reactor RA-0 for the next starting up at Cordoba National University

    International Nuclear Information System (INIS)

    Martin, H.R.; Chautemps, N.A.; Rumis, D.A.

    1991-01-01

    Due to the fulfillment to the tasks for the new starting up of the RA-0 Nuclear Reactor situated at the National University of Cordoba, it was necessary to plan and organize the service of Radiologic Protection to meet the future requirements in normal operation. The special characteristics that an installation of this type has in the university field, required special attention for making the university staff become aware in the working proceedings to follow up in normal conditions, such as the case of emergency that would originate in the installation. The training of the teaching and non teaching staff of the National University of Cordoba, the adjusting of the installations, the obtention of dosimetry and measurement equipment and the implementation of a monitor system of the staff were the main tasks confronted for the reorganization of the sector. (Author) [es

  1. Enhancement the physical protection system of the WWR-SM reactor at Institute of Nuclear Physics of Academy of Science of the Republic of Uzbekistan

    International Nuclear Information System (INIS)

    Karabaev, Kh.Kh.; Rakhimbaev, A.T.; Rakhmanov, A.B.; Salikhbaev, U.S.; Yuldashev, B.S.

    2004-01-01

    Full text: Joining of the Republic of Uzbekistan to Non-Proliferation Treaty required the revision of nuclear fuel protection system and radioactive sources from illegal access in all stages of work with nuclear materials. One of the primary technical actions of ensuring non-proliferation of nuclear materials is physical protection. The project was worked out on upgrading and enhancement of the physical protection of the reactor building. In cooperation with Sandia National Laboratory and support of the Department of Energy (DOE) USA The first stage of the physical protection upgrading provided for fresh fuel protection: - the new fresh fuel storage room was built and equipped with the modern control and detection system, - the reactor building was equipped with detection devices and access control, - the central alarm station (CAS) has been built and equipped with computer control and observing system, - code access system has been implemented. The first stage of upgrading of physical protection system was accomplished for 4 months, and put into operation in 1996. The second stage of physical protection system modernization included the construction of the second barrier of the physical protection, equipping it with observation and control devices and also extension of the CAS. The perimeter around the reactor building was cleaned from trees, bushed and in a short time a two-fence barrier was erected. The access control point provided the secured intensified control of the access to the reactor territory. The physical protection system was supplied with equipment for safeguard and TV observation of perimeter, access control to the territory of the reactor: - the CAS was extended and computer observation control system was upgraded, - the badge station has been constructed, equipped and set up, - all doors, windows, reactor hall gate have been replaced by strengthened metal ones, - uninterruptible power supply (UPS) and diesel-generator have been installed, - the

  2. Tasks related to increase of RA reactor exploitation and experimental potential, 01. Designing the protection chamber in the RA reactor hall for handling the radioactive experimental equipment (I-II) Part II, Vol. II

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    This second volume of the project for construction of the protection chamber in the RA reactor hall for handling the radioactive devices includes the technical description of the chamber, calculation of the shielding wall thickness, bottom lead plate, horizontal stability of the chamber, cost estimation, and the engineering drawings

  3. Process and device for the protection of steam-raising units, particularly of nuclear reactors

    International Nuclear Information System (INIS)

    Beyer, W.; Wieling, N.; Stellwag, B.

    1986-01-01

    To protect the housing against corrosion by chemical conditioning of the feedwater, the redox potential of the feedwater and the corrosion potential of at least one pipe of the pipe bundle is continuously determined during operation of the steam raising unit. With potentials indicating the danger of corrosion, the quality of the secondary water can be improved by suitable measures. (orig./HP) [de

  4. Analysis of the impact of an aircraft crash on underground concrete ducts with protective slab at reactor buildings

    International Nuclear Information System (INIS)

    Kotulla, B.; Hansson, V.

    1977-01-01

    In this paper different types of idealization for a dynamic analysis of underground concrete ducts with protective slab are discussed and compared. Ducts between reactor and control building of a nuclear power plant are to be designed for loadings produced by an aircraft crash. These ducts have a height of about three to four meters and are two to eight meters wide. They are designed with a protective slab about 1.5 m in thickness at ground level and with an intermediate layer of earth of about one meter in thickness. An analysis has to take into account the combined effects of a protective slab with a relatively thin intermediate layer of earth and the underlaying duct and layer of soil with the nonlinear behavior of concrete due to cracking. For describing this behavior two types of idealization were made. One type is a continuum type calculation which describes the slab, the soil and the duct by finite elements. In the other type of idealization a model consisting of springs and lumped masses is used. The protective slab and the intermediate layer of earth may be described as a plate on elastic foundation. The behavior of the cracked part of the plate and the part of earth layer beneath and loads transferred to the uncracked part of the slab and the surrounding soil may be described by parallel springs. Spring and mass of this part of the model have to take into account the cracking of the upper slab which leads to a nonlinear characteristic of the spring. In addition the location of the loading in relation to the duct has to be considered. The duct may be described by a beam on elastic foundation which is loaded locally. From this model representative mass and spring have to be determined

  5. Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation in Eastern Europe

    Energy Technology Data Exchange (ETDEWEB)

    Dassen, Lars van; Delalic, Zlatan; Ekblad, Christer; Keyser, Peter; Turner, Roland; Rosengaard, Ulf; German, Olga; Grapengiesser, Sten; Andersson, Sarmite; Sandberg, Viviana; Olsson, Kjell; Stenberg, Tor

    2009-10-15

    The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral assistance to Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in various projects financed by the European Union. The purpose of this project-oriented report is to provide the Swedish Government and other funding agencies as well as other interested audiences in Sweden and abroad with an encompassing understanding of our work and in particular the work performed during 2008. the activities are divided into four subfields: Nuclear waste management; Reactor safety; Radiation safety and emergency preparedness; and, Nuclear non-proliferation. SSM implements projects in the field of spent nuclear fuel and radioactive waste management in Russia. The problems in this field also exist in other countries, yet the concentration of nuclear and radioactive materials are nowhere higher than in north-west Russia. And given the fact that most of these materials stem from the Cold War era and remain stored under conditions that vary from 'possibly acceptable' to 'wildly appalling' it is obvious that Sweden's first priority in the field of managing nuclear spent fuel and radioactive waste lies in this part of Russia. The prioritisation and selection of projects in reactor safety are established following thorough discussions with the partners in Russia and Ukraine. For specific guidance on safety and recommended safety improvements at RBMK and VVER reactors, SSM relies on analyses and handbooks established by the IAEA in the 1990s. In 2008, there were 16 projects in reactor safety. SSM implements a large number of projects in the field of radiation protection and emergency preparedness. The activities are at a first glance at some distance from the activities covered and

  6. Operating experience feedback report -- turbine-generator overspeed protection systems: Commercial power reactors. Volume 11

    International Nuclear Information System (INIS)

    Ornstein, H.L.

    1995-04-01

    This report presents the results of the US Nuclear Regulatory Commission's Office for Analysis and Evaluation of Operational Data (AEOD) review of operating experience of main turbine-generator overspeed and overspeed protection systems. It includes an indepth examination of the turbine overspeed event which occurred on November 9, 1991, at the Salem Unit 2 Nuclear Power Plant. It also provides information concerning actions taken by other utilities and the turbine manufacturers as a result of the Salem overspeed event. AEOD's study reviewed operating procedures and plant practices. It noted differences between turbine manufacturer designs and recommendations for operations, maintenance, and testing, and also identified significant variations in the manner that individual plants maintain and test their turbine overspeed protection systems. AEOD's study provides insight into the shortcomings in the design, operation, maintenance, testing, and human factors associated with turbine overspeed protection systems. Operating experience indicates that the frequency of turbine overspeed events is higher than previously thought and that the bases for demonstrating compliance with NRC's General Design Criterion (GDC) 4, Environmental and dynamic effects design bases, may be nonconservative with respect to the assumed frequency

  7. Computerized reactor protection and safety related systems in nuclear power plants. Proceedings of a specialists' meeting. Working material

    International Nuclear Information System (INIS)

    1998-01-01

    Though the majority of existing control and protection systems in nuclear power plants use old analogue technology and design philosophy, the use of computers in safety and safety related systems is becoming a current practice. The Specialists Meeting on ''Computerized Reactor Protection and Safety Related Systems in Nuclear Power Plants'' was organized by IAEA (jointly by the Division of Nuclear Power and the Fuel Cycle and the Division of Nuclear Installation Safety), in co-operation with Paks Nuclear Power Plant in Hungary and was held from 27-29 October 1997 in Budapest, Hungary. The meeting focused on computerized safety systems under refurbishment, software reliability issues, licensing experiences and experiences in implemented computerized safety and safety related systems. Within a meeting programme a technical visit to Paks NPP was organized. The objective of the meeting was to provide an international forum for the presentation and discussion on R and D, in-plant experiences in I and C important to safety, backfits and arguments for and reservations against the digital safety systems. The meeting was attended by 70 participants from 16 countries representing NPPs and utility organizations, design/engineering, research and development, and regulatory organizations. In the course of 4 sessions 25 technical presentations were made. The present volume contains the papers presented by national delegates and the conclusions drawn from the final general discussion

  8. Decision making on population protection in a large-scale radioactive contamination following a nuclear reactor accident

    International Nuclear Information System (INIS)

    Konstantinov, Yu. O.

    1993-01-01

    Since the first years of development of nuclear power the most serious attention has been given to the planning of measures of population protection in the event of a radioactive release to atmosphere from a nuclear reactor. In the 60s 'Criteria for urgent decision making in the event of an accidental radioactive release into the environment' were developed in the USSR. When substantiating numerical values of potential radiation doses reasoning the implementation of countermeasures, specific conditions of emergency situations, characteristics of countermeasures and the real possibilities of timely dosimetric estimation of the situation were considered. The 'Criteria' were designed for urgent decision making at an early stage, in the first hours and days following the emergency. After the start of the Chernobyl accident on April 26, 1986, decisions on measures of protection of the population living in proximity to the site of the accident, including relocation of residents of the town of Pripyat on May 27, 1986, were taken on the basis of this document, as well as decisions for iodine prophylaxis and for relocation of other settlements within the 30 km zone. The decisions were taken by the result of the estimation prediction of the radiation situation which showed a possibility of an excess of criteria levels by external gamma radiation and by inhalation of radioiodine

  9. Environmental qualification and functional issues for microprocessor-based reactor protection systems

    International Nuclear Information System (INIS)

    Korsah, K.; Kisner, R.; Wood, R.T.; Antonescu, C.

    1992-01-01

    Issues of obsolescence and lack of intrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I ampersand C) systems with digital and microprocessor-based systems. This movement away from analog can be expected to increase in advanced light-water reactors (ALWRs), which will make extensive use of fiber optic transmission, multiplexing techniques, and microprocessor-based technology. Although these technologies have several advantages and, in fact, have been in widespread use in the non-nuclear industry for several years, their application to safety-related systems in nuclear power plants raises key issues relating to the systems' environmental and functional reliability. For example, does the new hardware introduce additional system aging degradation mechanisms that could adversely impact the safety of the plant? Do the systems introduce the possibility of new and different malfunction scenarios or increase the probability of common-mode failures that could reduce the reliability of the safety system?. Are current environmental qualification standards adequate for microprocessor-based I ampersand C systems? Accordingly in 1991 the Nuclear Regulatory Commission (NRC) initiated the qualification of advanced Instrumentation and Control Systems program at ORNL to investigate issues that may arise with the use of advanced digital I ampersand C in ALWRs. The results of our studies to date are summarized in this paper

  10. NKS/RAK-2. Protection against radioactive release in reactor accidents

    International Nuclear Information System (INIS)

    Lindholm, I.

    1995-01-01

    The work scope of RAK-2 project is divided into three subprojects: 1. Severe accident phenomenology. 2. Computerized accident management. 3. Reactors in Nordic surroundings. All three subprojects are ongoing. The project work on three subareas is in general progressing according to the time schedule and budget. The construction of melt jet breakup test facility at Kungliga Tekniska Hoegskolan (KTH) has been delayed due to complexity of the test arrangement and due to meeting the necessary safety requirements connected to tests mixing water and high temperature melts. Because of the delay in melt jet break up tests a slight redirection of the KTH work for NKS was taken. The present KTH work concentrates on theoretical studies of melt pool behavior in the lower head and on theoretical/experimental studies on core melt discharge from the pressure vessel failure. It is expected that single drop melt-water interaction experiments to study the thermal fragmentation phenomenon will begin in very early 1996. The recriticality studies are well underway, but the work is proposed to continue in 1996 to get more analyses carried out. (au)

  11. Burst protection for reactor pressure vessels using a hinged support bearing

    International Nuclear Information System (INIS)

    Michel, E.; Maritsch, F.

    1976-01-01

    The invention deals with a simplification of the design and manufacture and the way of controlling a hinged support bearing used as burst protection. The pure pressure load of the, e.g., 32 hinged supports distributed along the circumference of the pressure vessel head is achieved in the braced state with little control effort by a pure rotating motion caused pneumatically or hydraulically. The hinged supports are inclined by about 45 0 upwards/outwards in the braced state and with their cap-shaped head and foot are selflocking by pivoted between a supporting structure, firmly connected with the building, and a fishing ring. (TK) [de

  12. Clinch River breeder reactor sodium fire protection system design and development

    International Nuclear Information System (INIS)

    Foster, K.W.; Boasso, C.J.; Kaushal, N.N.

    1984-01-01

    To assure the protection of the public and plant equipment, improbable accidents were hypothesized to form the basis for the design of safety systems. One such accident is the postulated failure of the Intermediate Heat Transfer System (IHTS) piping within the Steam Generator Building (SGB), resulting in a large-scale sodium fire. This paper discusses the design and development of plant features to reduce the consequences of the accident to acceptable levels. Additional design solutions were made to mitigate the sodium spray contribution to the accident scenario. Sodium spill tests demonstrated that large sodium leaks can be safely controlled in a sodium-cooled nuclear power plant

  13. Vulnerability Analysis for Physical Protection System at Hypothetical Facility of a Different Type Reactor

    International Nuclear Information System (INIS)

    Jung, Won-Moog; Kim, Jung-Soo; Kim, Jae-Kwang; Yoo, Ho-Sik; Kwak, Sung-Ho; Jang, Sung-Soon

    2007-01-01

    Since the 9/11 event in the U.S.A, International terror possibilities has been increased for nuclear facilities including nuclear power plants(NPPs). It is necessary to evaluate the performance of an existing physical protection system(PPS) at nuclear facilities based on such malevolent acts. Detection, delay, and response elements are all important to PPS. They are used for the analysis and evaluation of a PPS and its effectiveness. Methods are available to analyze a PPS and evaluate its effectiveness. Sandia National Laboratory(SNL) in the U.S.A developed a System Analysis of Vulnerability to Intrusion (SAVI) computer code for evaluating the effectiveness of PPS against outsider threats. This study presents the vulnerability analysis of the PPS at hypothetical facility of a different type using SAVI code that the basic input parameters are from PPS of a different type. For analysis, first, the site-specific Adversary Sequence Diagrams(ASDs) of the PPS is designed. It helps to understand the functions of the existing PPS composed of physical areas and Protection Elements(PEs). Then, the most vulnerable path of an ASD as a measure of effectiveness is determined. The results in the analysis can compare with the most vulnerable paths at a different type

  14. Protection system for minimizing the consequences of a flow blockage incident at a pool-type research reactor

    International Nuclear Information System (INIS)

    de Vries, J.W.; van Dam, H.; Gysler, G.

    1990-01-01

    Safety analysis activities were performed for the HOR, a pool-type research reactor with plate-type fuel elements and a maximum licensed power of 3 MW. Following internationally accepted guidelines, a wide variety of possible process disturbances has been considered. For the HOR the most aggravating accident conditions could result from a sudden flow blockage of cooling channels. If this event occurs in the high power density region of the core, a decrease of the hot channel flow either causes flow reversal or prompts burnout. Unless the reactor is scrammed in time, the fuel plates will heat up rapidly and local melting will occur with possible propagation of voiding and burnout to adjacent channels. In the analysis, melting of the cladding has been considered by using a simplified model approach. The number of voided coolant channels, as well as the propagation rate of fuel plates reaching locally the melting temperature, were calculated for different conditions of operation. In order to reduce the risk of a fuel melt accident occurring at the HOR, the protection system features a special design option. The system recognizes cooling channel voiding by detection of a sudden decrease of neutron flux. In the present work, it has been shown that a flow blockage incident can be detected in the early stages of development. Also, in accordance with the results of experimental tests, it can be concluded that in many cases melting of fuel plates will be effectively prevented. If such an accident occurs on a very fast time scale, at least the radiological consequences are significantly mitigated by preventing propagation, thus limiting the number of molten fuel plates

  15. Radiation Protection Practices during the Helium Circulator Maintenance of the 10 MW High Temperature Gas-Cooled Reactor-Test Module (HTR-10

    Directory of Open Access Journals (Sweden)

    Chengxiang Guo

    2016-01-01

    Full Text Available Current radiation protection methodology offers abundant experiences on light-water reactors, but very few studies on high temperature gas-cooled reactor (HTR. To fill this gap, a comprehensive investigation was performed to the radiation protection practices in the helium circulator maintenance of the Chinese 10 MW HTR test module (HTR-10 in this paper. The investigation reveals the unique behaviour of HTR-10’s radiation sources in the maintenance as well as its radionuclide species and presents the radiation protection methods that were tailored to these features. Owing to these practices, the radioactivity level was kept low throughout the maintenance and only low-level radioactive waste was generated. The quantitative analysis further demonstrates that the decontamination efficiency was over 89% for surface contamination and over 34% for γ dose rate and the occupational exposure was much lower than both the limits of regulatory and the exposure levels in comparable literature. These results demonstrate the effectiveness of the reported radiation protection practices, which directly provides hands-on experience for the future HTR-PM reactor and adds to the completeness of the radiation protection methodology.

  16. Refurbishing the reactor protection systems of VVER-440/230 and VVER-1000/320 nuclear power plants with exclusively digital IandC systems

    International Nuclear Information System (INIS)

    Martin, M.

    1997-01-01

    The refurbishment of reactor protection systems of nuclear power plants is based on two sets of requirements: engineering aspects such as performance, qualification and licensing, as well as interfaces to other systems; and cost-benefit relationships, ease of service and maintenance as well as installation during scheduled outages. A number of WWER-440 and WWER-1000 nuclear plants have announced their intention to refurbish their protection systems. Since 1994, these plants have been placing orders with Siemens for new protection systems, including the neutron flux monitoring system utilizing the advanced system TELEPERM XS. This exclusively digital IandC system provides an excellent foundation for the remaining plant service life

  17. Technical evaluation report on the monitoring of electric power to the reactor protection system for the Nine Mile Point Nuclear Station, Unit 1 (Docket No. 50-220)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1984-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Nine Mile Point Nuclear Station, Unit 1. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  18. Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Brunswick Steam Electric Plant, Units 1 and 2

    International Nuclear Information System (INIS)

    Selan, J.C.

    1982-01-01

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Brunswick Steam Electric Plant, Units 1 and 2. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications with time delays verified by GE, will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources

  19. Flow protection trip limits operational charge-discharge facility -- C Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van Wormer, F.W.

    1958-09-19

    Because of wide variations in the venturi throat pressure, well beyond the panellit gage trip range, that occur during the sequence of operational charge-discharge, the panellit gage cannot be included in the scram safety circuit during the period of time that charge- discharge operations are being performed. In its stead, the function of the panellit gage is replaced in an overlapping manner by a tube inlet pressure monitor that is equipped with high and low pressure trip mechanisms that may be included in the scram safety circuit during the time that the panellit gage must be by-passed. The tube inlet pressure monitor is then used to provide the protection from unstable flow that is normally obtained with the panellit gage. This memorandum describes the manner in which the tube inlet pressure monitor trip points are to be determined and used.

  20. Bursting-protection configuration for cylindrical steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Mutzl, J.

    1979-01-01

    The bursting-protection jacket consists of cylinder courses, being joined together in axial direction, and of a bottom and a cover, being connected by means of axial prestressing tendons. For absorption and transmission of the steam generator weight and the bursting forces the bottom consists of a conical shell, tapered towards the side of the steam generator, and a support ring supporting the bottom circle of the cone. This support ring is built in sandwich construction and is connected with the axial tendons. The conical shell may be reinforced by radial ribs. If a primary coolant pump is built in there is provided for a rocking bearing between its pump casing flange and the bottom. (DG) [de

  1. Tritium conversion and its influence on personnel protection at a fusion reactor

    International Nuclear Information System (INIS)

    Easterly, C.E.; Phillips, J.E.

    1980-01-01

    Tritium gas is less hazardous than tritiated water. The difference appears to be on the order of 10 4 rather than the previously used figure of 10 2 . With an additional factor of 10 2 for protective clothing, a potential difference of 10 6 in the relative HTO to HT hazard results. The mechanisms for conversion of HT to HTO are not fully known but models presented strongly link the presence of H and OH radicals with the ultimate formation of water. Therefore a hindrance of this conversion may be possible by using specific quencher materials. The maintenance of tritium in the gaseous form then allows for a wider variety of tritium management schemes

  2. Reliability Assessment Method of Reactor Protection System Software by Using V and Vbased Bayesian Nets

    International Nuclear Information System (INIS)

    Eom, H. S.; Park, G. Y.; Kang, H. G.; Son, H. S.

    2010-07-01

    Developed a methodology which can be practically used in quantitative reliability assessment of a safety c ritical software for a protection system of nuclear power plants. The base of the proposed methodology is V and V being used in the nuclear industry, which means that it is not affected with specific software development environments or parameters that are necessary for the reliability calculation. Modular and formal sub-BNs in the proposed methodology is useful tool to constitute the whole BN model for reliability assessment of a target software. The proposed V and V based BN model estimates the defects in the software according to the performance of V and V results and then calculate reliability of the software. A case study was carried out to validate the proposed methodology. The target software is the RPS SW which was developed by KNICS project

  3. Development and experimental qualification of the new safety-criticality CRISTAL package; Developpement et qualification experimentale du nouveau formulaire de surete-criticite Cristal

    Energy Technology Data Exchange (ETDEWEB)

    Mattera, Ch

    1998-11-01

    This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA 93) and the newest accurate codes (APOLLO 2, MORET 4, TRIPOLI 4). The CRISTAL system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route. To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OECD working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author)

  4. Development and experimental testing of the new safety-criticality Cristal package; Developpement et qualification experimentale du nouveau formulaire de surete-criticite Cristal

    Energy Technology Data Exchange (ETDEWEB)

    Mattera, Ch

    1998-11-10

    This thesis is concerned with Criticality-Safety studies related to the French Nuclear Fuel Cycle. We first describe the steps in the nuclear fuel cycle and the specific characteristics of these studies compared with those performed in Reactor Physics. In order to respond to the future requirements of the French Nuclear Program, we have developed a new package CRISTAL based on a recent cross sections library (CEA93) and the newest accurate codes (APOLLO2, MORET4, TRIPOLI4). The cristal system includes two calculations routes: a design route which will be used by French Industry (COGEMA/SGN) and a reference route.) To transfer this package to the French industry, we have elaborated calculation schemes for fissile solutions, dissolver media, transport casks and storage pools. Afterwards, these schemes have been used for the CRISTAL experimental validation. We have also contributed to the CRISTAL experimental database by reevaluating a French storage pool experiment: the CRISTO II experiment. This revaluation has been submitted to the OCDE working group in order that this experiment can be used by international criticality safety engineers to validate calculations methods. This work represents a large contribution to the recommendation of accurate calculation schemes and to the experimental validation of the CRISTAL package. These studies came up to the French Industry expectations. (author) 70 refs.

  5. Impact of the impurity seeding for divertor protection on the performance of fusion reactors

    Science.gov (United States)

    Siccinio, Mattia; Fable, Emiliano; Angioni, Clemente; Saarelma, Samuli; Scarabosio, Andrea; Zohm, Hartmut

    2017-10-01

    A 0D divertor and scrape-off layer (SOL) model has been coupled to the 1.5D core transport code ASTRA. The resulting numerical tool has been employed for various parameter scans in order to identify the most convenient choices for the operation of electricity producing fusion devices with seeded impurities for the divertor protection. In particular, the repercussions of such radiative species on the main plasma through the fuel dilution have been taken into account. The main result we found is that, when the limits on the maximum tolerable divertor heat flux are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, i.e. no improvement would descend from a further increase of R or BT once the maximum has been reached. This occurrence appears as an intrinsic physical limit for all devices where a radiative SOL is needed to deal with the power exhaust. Furthermore, the relative importance of the different power loss channels (e.g. hydrogen radiation, charge exchange, perpendicular transport and impurity radiation), through which the power entering the SOL is dissipated before reaching the target plate, is investigated with our model.

  6. Methods and criteria for evaluation of nuclear reactor fire protection alternatives and modifications

    International Nuclear Information System (INIS)

    Levinson, S.H.

    1982-01-01

    The objective of this work is to develop a methodology for the evaluation of a fire protection system in a nuclear power plant and demonstrate the feasibility of encoding this method in a computer program. A Monte Carlo simulation has been developed; it is divided into the four phases of a fire scenario: ignition, detection, suppression and propagation. The ignition model consists of probabilistically determining at what location within the zone a fire will occur. The detection model is divided into two components. THe first is the automatic detection model, which calculates the fire's physical symptoms and compares them against the threshold values of the detectors specified for the zone to determine a time-to-detection. The second part is the human detection model; this evaluates the time required for a human to observe and report a fire. If detection is successful, the suppression mode determines if the fire is effectively extinguished, and if so, the time required to do so. This model is also divided into an automatic and human component. The propagation model is embedded in a deterministic, control-volume computer code which calculates the fire scenario history. A computer program, FIRES, is described which supports the developed models. FIRES is an interactive graphics package providing a simple means of establishing the many input parameters. In addition to allowing parameter values to be easily set or modified, the graphics provides a convenient display mode for the results of a simulation

  7. Public protection strategies in the event of a nuclear reactor accident: multicompartment ventilation model for shelters

    International Nuclear Information System (INIS)

    Aldrich, D.C.; Ericson, D.M. Jr.

    1978-01-01

    A multicompartment ventilation model has been presented for the calculation of airborne radioactive material concentrations internal to structures. The model was used to estimate the potential effectiveness of sheltering in reducing the dose due to inhaled radionuclides. The sensitivity of the model to parameter values and protection strategies was discussed. Using ''best estimate'' values for the model parameters, this analysis indicated that sheltered individuals received a reduction of 35 percent in the dose from inhaled radionuclides. Larger reductions would be possible if lower values of the ventilation rate n, could be achieved by either tighter building construction or emergency sealing of openings in the structure. Such emergency means could include taping windows, placing wet paper over cracks, etc. Further analysis indicated that the strategy of opening doors and windows, turning on ventilating systems, etc., in an attempt to ''air-out'' the structure after the cloud of radioactive material had passed will most likely not contribute significantly to reduction in dose due to inhaled radionuclides unless very low initial ventilation rates are achieved. Although the available data did not allow quantitative predictions of dose reductions afforded by basements or other appropriately sealed-off rooms, preliminary analysis indicated qualitatively that they could be significant

  8. Radiation protection at the RA Reactor in 1984, Part -2, Annex 2a: Radioactivity control of the RA reactor environment - atmospheric precipitations, dust, water, soil, plants, fruit.

    International Nuclear Information System (INIS)

    Ajdacic, N.; Martic, M.; Jovanovic, J.

    1984-01-01

    Control of radioactivity in the biosphere in the vicinity of the RA reactor is part of the radioactivity control done regularly for the whole territory of the Vinca institute. During 1984 control was conducted according to the plan. According to the measured data no significant changes have been found in the surroundings of the RA reactor. All the analysed samples have followed the activity values of the precipitations

  9. Radiological protection of the staff during the decommissioning operations of the Romanian VVR-S research reactor

    International Nuclear Information System (INIS)

    Ene, D.C.

    2002-01-01

    Dose rate estimates for periods of 100 days and 6, 10, 25, 100 years after the shut down of the Romanian VVR-S reactor are presented in this paper for some foreseen decommissioning activities which include: i) cutting the water pipe in the pump room and the reactor sealing operations; ii) extracting reactor components; and iii) handling and dismantling the internal structures taken of from the reactor. For the reactor components extracted from the reactor, the considered calculation points were placed in the central plan of the items, on the surface and at distances from the surface which correspond to +0.2m, +1m, +2m, +8m, and +10m. Time dependence of the resulted dose rates are presented and discussed. Qualitative comparison with the measured values from other VVR-S reactors is done. The obtained results assist to develop working procedures that must be observed during the decommissioning activities. (author)

  10. A Study on Quantitative Assessment of Design Specification of Reactor Protection System Software Using Bayesian Belief Networks

    International Nuclear Information System (INIS)

    Eom, H. S.; Kang, H. G.; Chang, S. C.; Park, G. Y.; Kwon, K. C.

    2007-02-01

    This report propose a method that can produce quantitative reliability of safety-critical software for PSA by making use of Bayesian Belief Networks (BBN). BBN has generally been used to model the uncertain system in many research fields. The proposed method was constructed by utilizing BBN that can combine the qualitative and the quantitative evidence relevant to the reliability of safety-critical software, and then can infer a conclusion in a formal and a quantitative way. A case study was also carried out with the proposed method to assess the quality of software design specification of safety-critical software that will be embedded in reactor protection system. The V and V results of the software were used as inputs for the BBN model. The calculation results of the BBN model showed that its conclusion is mostly equivalent to those of the V and V expert for a given input data set. The method and the results of the case study will be utilized in PSA of NPP. The method also can support the V and V expert's decision making process in controlling further V and V activities

  11. Upgrading Planning and Executive Strategy for Reactor Protection System and Relative Equipment in Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jiang Zuyue

    2010-01-01

    Qinshan Nuclear Power Plant (QNPP) is the first nuclear power plant in China which completed the reactor protection system (RPS) upgrading with new digital safety instrumentation and control (I and C) platform instead of original analog system. At the same time,the nuclear instrumentation system (NIS) was upgraded with the same digital I and C platform. For adapting QNPP's actual engineering situation,the upgrading planning was taken by comprehensively investigating current development and application of digital safety I and C platform in the worldwide scope and by reviewing plant's original systems operation history. The project executive strategy-QNPP's leading role with necessary overseas cooperation and internal technical supports as great as possible, was determined. Some significant factors might influence and restrict the RPS and relative equipment upgrading executive actions in an operating NPP were analyzed.Finally, the engineering feasibility was briefly assessed to recognize the anticipated issues and difficulties and to prepare the relative solutions in advance for the purpose of ensuring the RPS upgrading objectives completely realized. (authors)

  12. Dosimetry and radiation protection at the RA reactor in 1972; Dozimetrija i zastita kod reaktora RA u 1972. godini

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1973-07-01

    Dosimetry results collected within radiation protection of the RA reactor during this year are presented. Neutron and gamma radiation data were measured at characteristic control points. Statistical review of the total number of measurement is given as well. The report includes contents of radioactive gasses and aerosols in the air, as well as the contamination data of surfaces, clothes and uncovered body parts of the personnel. Particular accident which occurred during dismantling of the experimental channel containing the capsule with the new fuel element was analysed. This accident occurred at the RA reactor at the beginning of this year. It was found that that the maximum individual external dose was 2.2 R, and that only one individual was exposed to this dose. About 15% of the personnel was exposed to doses between 1 and 2 R, the remaining 85% was exposed to doses less than 1 R. Base on the frequency of activities undertaken in the contaminated regions, safety and control measures and expected internal exposure of the personnel, it was evaluated that the internal exposure could be neglected compared to the external exposure of the personnel. Prikazani su rezultati sakupljani u toku godine u okviru dozimetrijske kontrole i zastite kod reaktora. Dati su podaci o nivoima neutrona i gama zracenja na karakteristicnim kontrolnim mestima, kao i statisticki pregledi ukupnog broja merenja. Navedeni su rezultati merenja sadrzaja radioaktivnih gasova i aerosola u vazduhu, kao i stepena kontaminacije povrsina, odece i otkrivenih delova tela radnog osoblja. Analiziran je specifican akcident koji se odigrao na reaktoru pocetkom godine, pri demontazi eksperimentalnog kanala sa kapsulom novog gorivog elementa. Na kraju, izlozena je analiza ozracivanja radnog osoblja. Konstatovano je da je maksimalna individualna doza spoljasnjeg ozracivanja bila 2,2 (R), i da je ovoj dozi bilo izlozeno samo jedno lice. Oko 15% osoblja bilo je izlozeno dozama izmedju 1 i 2 (R), a ostalih 85

  13. Data acquisition, monitoring and diagnostic system for predictive control and protection of rotating components of IEAR-1 reactor by vibration analysis

    International Nuclear Information System (INIS)

    Serra, Reynaldo Cavalcanti; Tecco, Dorival Goncalves

    1996-01-01

    This work presents the vibration and temperature data acquisition, monitoring and diagnostic systems, recently installed in the primary circuit, secondary circuit and emergency generator of the IEA-R1 reactor at IPEN during the course of the first power elevation tests to 5MW. It incorporates a series of routines for equipment configuration, interactive automatic monitoring , data processing and documentation/storage without the exposure of operators in the radiological protection areas. (author)

  14. Behavior of the future LHC magnet protection diodes irradiated in a nuclear reactor at 4.6 K with intermediate annealing

    International Nuclear Information System (INIS)

    Berland, V.; Hagedorn, D.; Gerstenberg, H.

    1996-01-01

    In the framework of the LHC project at CERN, the effects of radiation on the electrical characteristics of epitaxial diodes for superconducting magnet protection were studied. The diodes were exposed to an irradiation dose up to 50 kGy and a neutron fluence of 10 15 n/cm 2 with intermediate thermal annealing each 10 kGy dose steps in the Technical University of Munich reactor at 4.6 K

  15. Radiation protection at the RA Reactor in 1985, Part -3, Removal and treatment of radioactive effluents for the needs of RA reactor

    International Nuclear Information System (INIS)

    Plecas, I.; Vukovic, Z.; Kostadinovic, A.

    1985-01-01

    Contaminated water originates from: hot cells, heavy water distillation device, storage pools for cooling and cutting of fuel elements, water biological shield of the reactor. During 1985 5 m of contaminated water was released from hot cells in the VR-pool containing highly radioactive waste water with 60 Co and trace amount of fission products [sr

  16. Radiation protection at the RA Reactor in 1985, Part -4, decontamination and treatment of solid radioactive materials for the needs of RA reactor

    International Nuclear Information System (INIS)

    Plecas, I.; Vukovic, Z.; Blagojevic, R.; Kostadinovic, A.

    1985-01-01

    This report describes the activity of the decontamination and treatment team for the needs of the RA reactor, its equipment, working conditions, methods for decontamination, means of decontamination, type and quantity of decontaminated surfaces, number of decontaminated objects, quantity of collected radioactive solid wastes, their packaging, transport to the storage place and topography od radiation field in the storage during 1985 [sr

  17. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

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

  18. Reactor protection device

    Energy Technology Data Exchange (ETDEWEB)

    Shida, T; Hirose, M

    1977-01-19

    Purpose: To prevent abrupt increase or decrease in the recycling flow rate by comparing output signals from controllers in each of the loops in the recycling flow rate control system to lock the positions of fluid coupling scooping pipes or flow control valves corresponding to the groups generating abnormal signals. Constitution: The recycling flow rate is controlled by the r.p.m. of a motor directly coupled with a recycling pump and the value of r.p.m. is in proportion to the generator frequency varied with the sliding operation of the fluid coupling in MG set. The sliding operation of the fluid coupling is adjusted by a scooping pipe driver. When the device is set to automatic operation, the output signal of the main controller is delivered to the recycling flow rate control system, the output signal of which is input to respective scooping pipe drivers. The loop output signals are supplied to an adder where the deviation signal between both of them are detected and the scooping pipe is locked if the set value is exceeded.

  19. Reactor protection device

    International Nuclear Information System (INIS)

    Shida, Toichi; Hirose, Masao.

    1977-01-01

    Purpose: To prevent abrupt increase or decrease in the recycling flow rate by comparing output signals from controllers in each of the loops in the recycling flow rate control system to thereby lock the positions of fluid coupling scooping pipes or flow control valves corresponding to the groups generating abnormal signals. Constitution: The recycling flow rate is controlled by r.p.m. of a motor directly coupled with a recycling pump and the value of r.p.m. is in proportion to the generator frequency varied with the sliding operation of the fluid coupling in MG set. The sliding operation of the fluid coupling is adjusted by a scooping pipe driver. When the device is set to automatic operation, the output signal of the main controller is delivered to the recycling flow rate control system, the output signal of which is input to respective scooping pipe drivers. The loop output signals are supplied to an adder where the deviation signal between both of them are detected and the scooping pipe is locked if the set value is exceeded. (Yoshino, M.)

  20. Radiation protection at the RA Reactor in 1986, Part -2, Annex 2a, Radioactivity control of the RA reactor environment (atmospheric precipitations, dust, water, soil, plants, fruit...)

    International Nuclear Information System (INIS)

    Ajdacic, N.; Martic, M.; Jovanovic, J.

    1986-01-01

    Control of radioactivity in the biosphere in the vicinity of the RA reactor is part of the radioactivity control done regularly for the whole territory of the Vinca institute (report by the same authors included in this Annex). During 1986 control was conducted according to the plan until May 1, 1986 when a dramatic increase of the precipitations and all other samples from the biosphere was recorded. According to the measured data no significant changes have been found in the surroundings of the RA reactor, until April 29 1986. Since then more detailed control was conducted, the number of samples was increased, apart from standard measuring procedure of total beta activity measurements, gamma spectrometry of all samples was applied. High activity level of the following nuclides was found: Iodine, cerium,cesium, tellurium, ruthenium, barium, lanthanum, etc. As an example activity of ?1?3?1 I in one sample was 564±5 kBq/m 2 [sr

  1. Radiation protection at the RA reactor in 1984, Part III Removal of the liquid radioactive effluents for the needs of the RA reactor

    International Nuclear Information System (INIS)

    Mandic, M.; Plecas, I.; Vukovic, Z.; Knezevic, Lj.; Jankovic, O.; Kostadinovic, A.; Mihailovic, B.

    1984-01-01

    Contaminated water originates from: hot cells, heavy water distillation device, storage pools for cooling and cutting of fuel elements, water biological shield of the reactor. During 1984, 400 liters of water contaminated by 60 Co was treated. Most recent measurements showed that the VR-1 pool contains 280 m 3 of effluents having specific activity of 3.3 10 4 Bq/ml, and VR-2 contains 30 m 3 with specific activity of 4 10 3 Bq/ml

  2. Inspection methods for physical protection Task III review of other agencies' physical security activities for research reactors

    International Nuclear Information System (INIS)

    In Task I of this project, the current Nuclear Regulatory Commission (NRC) position-on physical security practices and procedures at research reactors were reviewed. In the second task, a sampling of the physical security plans was presented and the three actual reactor sites described in the security plans were visited. The purpose of Task III is to review other agencies' physical security activities for research reactors. During this phase, the actions, procedures and policies of two domestic and two foreign agencies other than the NRC that relate to the research reactor community were examined. The agencies examined were: International Atomic Energy Agency; Canadian Atomic Energy Control Board; Department of Energy; and American Nuclear Insurers

  3. Plant protection system optimization studies to mitigate consequences of large breaks in the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khayat, M.I.; March-Leuba, J.

    1993-01-01

    This paper documents some of the optimization studies performed to maximize the performance of the engineered safety features and scram systems to mitigate the consequences of large breaks in the primary cooling system of the Advanced Neutron Source (ANS) Reactor

  4. Experimental and computational analysis of the hot water layer for the radiological protection in swimming pool reactor

    International Nuclear Information System (INIS)

    Ribeiro, Rogerio.

    1995-01-01

    Pool reactors are research reactors, which allow easy access to the core and rare simple to operate. Reactors of this kind operating at power levels higher than about one megawatt need a hot water layer at the surface of the pool, in order to keep surface activity below acceptable levels and enable free access to the upper part of the reactor. An experimental apparatus was constructed to study the hot water layer stability. Thermocouples were used to measure the temperature field. A numerical analysis was conducted simultaneously. Regarding experimental results, representative temperature contour lines of the hot water layer were plotted. The temperature field was determined in the numerical analysis and temperature contour lines corresponding to those of the experimental results were plotted. The hot water layer kept stable for experimental and numerical results. Good agreement between the results for the hot water layer position and thickness has been obtained. (author). 21 refs., 40 figs., 15 tabs

  5. Report of the Federal Ministry for the Environment, Protection of Nature and Reactor Safety, on the reactor accident of Chernobyl, its repercussions, and precautions taken or to be taken - including addenda

    International Nuclear Information System (INIS)

    Petroll, M.

    1986-01-01

    Apart from the report of the Federal Ministry for the Environment, the publication contains the following chapters: 1) Monitoring of environmental radioactivity; 2) analysis of propagation processes; 3) control and measuring points of the Federal Laender to monitor environmental radioactivity; 4) determination of the local dose rate; 5) concentration of radioactivity in air and soil (graphs); 6) up-to-date knowledge of events, measures; 7) nuclear power plants in the Federal Republic of Germany (review of technical safety); 8) interim report of the Committee on Reactor Safety - Reaktorsicherheitskommission - for preliminary evaluation; 9) interim report of the Committee on Radiation Protection - Strahlenschutzkommission - for assessment and evaluation of the effects. (HP) [de

  6. Protection of biofilms against toxic shocks by the adsorption and desorption capacity of carriers in anaerobic fluidized bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Petrozzi, S. (Biological Reaction Engineering Group, Chemical Engineering Dept., ETH, Zurich (Switzerland)); Kut, O.M. (Biological Reaction Engineering Group, Chemical Engineering Dept., ETH, Zurich (Switzerland)); Dunn, I.J. (Biological Reaction Engineering Group, Chemical Engineering Dept., ETH, Zurich (Switzerland))

    1993-05-01

    The aim of this study was to select a support medium for an anaerobic biofilm fluidized bed reactor (AFBR) for waste water treatment. Six materials, shale, pumice, porous glass, quartz sand, activated carbon and anthracite were used as carriers for the biofilm. The reactors were operated in parallel for several months with vapour condensate from a sulfite cellulose process as feed. The criteria used for the evaluation were: (a) Reproducibility of the reactor performance, (b) performance of the different carriers under various loading rates, (c) stability against toxic shock loadings using 2,4,6-trichlorophenol (TCP) as toxicant, (d) recovery capacity after intoxication and starvation, (e) adsorption/desorption behavior of the carriers. A comparison between four runs showed good reproducibility of the steady state removal rates. The performance of the reactors and the stability of the degradation rates were tested for a range of loading conditions. Unbuffered, buffered and pH controlled conditions were compared. The pumice carrier was best with respect to the degradation rate achieved per carrier mass. The response of the reactors to massive TCP step loadings was tested. Loadings less than 1.5 kg TCP/m[sup 3]d resulted in initially normal gas production rates for all the systems, except the activated carbon, whose gas production was partially inhibited from the start. After increasing the load to 1.5 kg TCP/m[sup 3]d the gas production rates of all the other reactors fell abruptly to zero. Restarting after 2 months, all reactors showed methanogenic activity without requiring new inoculum. (orig.)

  7. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Celestino M, E.

    2016-01-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  8. The concept of power correction techniques and its use in the reactor regulation and protection systems in Indian PHWRs

    International Nuclear Information System (INIS)

    Vaswani, P.D.; Kelkar, M.G.; Ghoshal, B.; Ashok Kumar, B.

    2010-01-01

    Reactor Power Measurement is an essential part of the Reactor Power Control Loop in PHWRs. None of the available power measuring sensor offers characteristics which allow their direct use in the Reactor Power Control Loop. Thermal power, which is considered as relatively accurate, suffers from measurement delays and is used only as reference. Neutronic power sensors like Ion Chambers and Self Powered Neutron Detectors (SPNDs) which sense instantaneous power suffer from inaccuracies. A technique is required which makes use of both types-reference power and instantaneous power to extract real power information from the signals. This paper describes techniques to calibrate (correct) neutronic power that with the thermal reference power signals. The paper also brings out limitation of the calibration technique. (author)

  9. Nuclear fast neutron reactor cooled by a liquid metal and of which internal structures are equipped with a thermal protection device

    International Nuclear Information System (INIS)

    Lemercier, G.; Lions, N.

    1986-01-01

    The internal structures of a nuclear fast neutron reactor are covered at least partially, on the most hot side, by a thermal protection device. This device comprises modular plates arranged end to end with a certain play between themselves and taking approximately the shape of the internal structures. Each plate is fixed in its center on the internal structures by a stud. A small plate fixed at one of the corners of each plate and covering partially the adjacent plates ensures the safety fixing of these ones [fr

  10. Technical-evaluation report on the monitoring of electric power to the reactor-protection system for the Oyster Creek Nuclear Generating Station. (Docket No. 50-219)

    International Nuclear Information System (INIS)

    Selan, J.C.

    1983-01-01

    During the operating license review for Hatch 2, the Nuclear Regulatory Commission (NRC) staff raised a concern about the capability of the Class 1E reactor protection system (RPS) to operate after suffering sustained, abnormal voltage or frequency conditions from a non-Class 1E power supply. Abnormal voltage or frequency conditions could be produced as a result of one of the following causes: combinations of undetected, random single failures of the power supply components, or multiple failures of the power supply components caused by external phenomena such as a seismic event. The purpose of this report is to evaluate the licensees submittal with respect to the NRC criteria and present the reviewer's conclusion on the adequacy of the design modifications to protect the RPS from abnormal voltage and frequency conditions

  11. Protective

    Directory of Open Access Journals (Sweden)

    Wessam M. Abdel-Wahab

    2013-10-01

    Full Text Available Many active ingredients extracted from herbal and medicinal plants are extensively studied for their beneficial effects. Antioxidant activity and free radical scavenging properties of thymoquinone (TQ have been reported. The present study evaluated the possible protective effects of TQ against the toxicity and oxidative stress of sodium fluoride (NaF in the liver of rats. Rats were divided into four groups, the first group served as the control group and was administered distilled water whereas the NaF group received NaF orally at a dose of 10 mg/kg for 4 weeks, TQ group was administered TQ orally at a dose of 10 mg/kg for 5 weeks, and the NaF-TQ group was first given TQ for 1 week and was secondly administered 10 mg/kg/day NaF in association with 10 mg/kg TQ for 4 weeks. Rats intoxicated with NaF showed a significant increase in lipid peroxidation whereas the level of reduced glutathione (GSH and the activity of superoxide dismutase (SOD, catalase (CAT, glutathione S-transferase (GST and glutathione peroxidase (GPx were reduced in hepatic tissues. The proper functioning of the liver was also disrupted as indicated by alterations in the measured liver function indices and biochemical parameters. TQ supplementation counteracted the NaF-induced hepatotoxicity probably due to its strong antioxidant activity. In conclusion, the results obtained clearly indicated the role of oxidative stress in the induction of NaF toxicity and suggested hepatoprotective effects of TQ against the toxicity of fluoride compounds.

  12. Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-07

    At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

  13. Draft report of a consultants meeting on core control and protection strategy of WWER-1000 reactors. Extrabudgetary programme on the safety of WWER-1000 NPPs

    International Nuclear Information System (INIS)

    1994-01-01

    At the consultants' meeting on the 'Safety of WWER-1000 Model 320 Nuclear Power Plants' organized by the IAEA within the framework of its Extrabudgetary Programme on the Safety of WWER-1000 NPPs, which was held in Vienna, 1-5 June 1992, the problem of core control and protection strategy was identified as an issue of safety concern. Considering the safety importance of this issue, a consultants' meeting on 'Core Control and Protection Strategy for WWER-1000 Reactors' was convened in Vienna in April 1994 attended by 20 international experts in the area of core control and protection in order to review control and protection system design, to compare them with western practices and to recommend corrective measures. The first WWER-1000 NPP was put into operation in 1980 and there are currently 19 units operating. The accumulated operational experience is more than 130 reactor-years. In addition, there are 8 units under various stages of construction. The previous general observations in the area of core control and protection strategy was focused on core design objectives, core design and fuel management, fuel assembly and core component designs, including burnable absorber and control rod designs, core power distribution control strategy, core control and protection system designs and in-core and ex-core instrumentation systems. While core design objectives of WWER-1000 plants are similar to western practices in general, there are important differences on the design limits and regulatory practices followed for the compliance with the design limits. As a result of previous general observations and specific concerns on core control and protection system design, three working groups were formed to further investigate the specific issues and to compile information on safety issues based on design differences between these plants and similar western plants, to identify areas which need further analysis and make recommendations for short-term and long-term corrective

  14. Simulation of protected and unprotected loss of flow transients in a WWER-1000 reactor based on the Drift-Flux model

    Energy Technology Data Exchange (ETDEWEB)

    Baghban, Ghonche [Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of). Nuclear Science and Technology Research Inst.; Shayesteh, Mohsen [Imam Hussein Univ., Tehran (Iran, Islamic Republic of). Dept. of Physics; Bahonar, Majid [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2017-03-15

    In view of the importance of studying coolant transient behavior in a nuclear reactor, this work is devoted to the thermal-hydraulic analysis of protected and unprotected loss of flow transients in a WWER-1000 reactor. A series of corresponding mathematical and physical models based on the four-equation Drift-Flux model has been applied. Based on a multi-channel approach, the core has been divided into different regions. Each region has different characteristics as represented in a single fuel pin with its associated coolant channel. Appropriate initial and boundary conditions have been considered and two situations of tripping four and two primary pumps in a protected core in addition to situation of tripping all four pumps in an unprotected core have been analyzed. For each transient, a full range of thermal-hydraulic parameters has been obtained. For verification of the proposed model, the results have been compared with those of the RELAP5/MOD3 and Bushehr nuclear power plant Final Safety Analysis Report (FSAR). A good agreement between results has been attained for the aforementioned transients.

  15. 10 CFR 73.55 - Requirements for physical protection of licensed activities in nuclear power reactors against...

    Science.gov (United States)

    2010-01-01

    ... protection of digital computer and communication systems and networks. (ii) Site-specific conditions that... criteria set forth in § 73.54 “Protection of Digital Computer and Communication systems and Networks” of... visitors shall register their name, date, time, purpose of visit, employment affiliation, citizenship, and...

  16. Reports by the Parliamentary Office for scientific and technological assessments. Tuesday, May 24, 2011. Hearing on the protection of a reactor core and critical circuit; Comptes rendus de l' Office Parlementaire d'Evaluation des Choix Scientifiques et Technologiques. Mardi 24 mai 2011. Audition sur la protection du coeur et des circuits critiques d'un reacteur

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-05-15

    In the context created by the Fukushima accident, members of the French Parliament, representatives of the French nuclear safety authority (ASN), of the French Institute for radiation protection and nuclear safety (IRSN), and of the CEA describe and discuss the technical aspects and mechanism of defence-in-depth of nuclear reactors (i.e. the different and successive levels of protection aimed at ensuring the reactor integrity to be maintained, even in case of failure of a critical circuit). Then, they discuss advances and researches in the field of protection of reactors. Several research programs are evoked which concern different elements of a nuclear plants such as the fuel, the reactor, loss of cooling system, and so on; these programs are based either on experiments or on simulations

  17. Development of a protection system for research reactor based in Field Programmable Gate Array - FPGA; Desenvolvimento de sistema de protecao para reator nuclear de pesquisa baseado em Field Programmable Gate Array - FPGA

    Energy Technology Data Exchange (ETDEWEB)

    Martins, Roque Hudson da Silva

    2016-07-01

    This study presents a implementation purpose of a protection system for research nuclear reactors by using a programed device FPGA (Field Programmable Gate Array). As well as logic protection method involved on an automatic shutdown (TRIP) of a reactor, that ensure the security on such systems. These new control and operation mechanics are developed to guarantee that the security limits of a power plant are not exceeded, these mechanics can work isolated or in groups to safe guard the security levels. For this implementation to be completed, there will be presented the main aspects and concepts referred to protection systems, mostly about research nuclear reactors, with some applications terms exposed. The system proposed at this paper was developed following the VHDL (Very High Speed Integrated Circuits) hardware describing language, and the Modelsim software from Altera Software to program the automatic turning off routines, and hypothetical simulations for such. The results show that for every software application for supporting nuclear reactors, like security devices, they have to meet the IEC 60880 criteria. This paper have great importance, seeing that nuclear reactor security systems, are a basic element for ensure the reactor security. (author)

  18. Reactor power measuring device

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. A recommendation of the National Board for Atomic Safety and Radiation Protection for the appointment of Nuclear Safety Control Officers for research reactors

    International Nuclear Information System (INIS)

    Adler, B.

    1990-01-01

    The Ordinance on the Implementation of Atomic Safety and Radiation Protection of the GDR requires that the managers of plants where nuclear facilities are operated appoint Control Officers for the fields of radiation protection, nuclear safety, physical protection, and accounting for and control of nuclear materials. The Control Officers are staff members of the operating organization but their appointment is subject to approval by the National Board and requires adequate qualification. The main task of the Control Officers as specialists is to give advice to the plant manager who retains responsibility for the safety of nuclear facilities, and to verify on his behalf that all requirements within their competence are met by the operating group. For this reason the Control Officer has to be absolutely independent of the head of the operating group. To enable the Control Officers to accomplish all necessary control activities and to guarantee independence from the head of the operating group, the plant manager has to establish adequate regulations of operation. As a pattern for such regulations the National Board has issued a Recommendation for the Appointment of Nuclear Safety Control Officers for Research Reactors, which provides a comprehensive survey of the requisite qualification features as well as the duties and rights of these Control Officers. This recommendation will be dealt with in the presentation

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

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

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

  2. Developments in radiological protection of the environment and a commentary on its implications for new build of nuclear reactors

    International Nuclear Information System (INIS)

    Brownless, George; Lazo, Ted

    2008-01-01

    Full text: Development of radiological protection of non-human biota continues to be the focus of much interest and differing views amongst the radiological protection community. To nurture discussion of these developments, the Nuclear Energy Agency's Committee on Radiological Protection and Public Health is taking the lead in organising coverage of a spectrum of views on the topic at the International Conference on Radioecology and Radioactivity in the Environment (Bergen, 2008), with the aim of assisting the international community to construct a consensual, fit-for-purpose approach. To support discussion of these views, the session will also include scientific presentations and reports from implementers. This paper will report on these developments, based principally on the session at the Conference but also other activities in which NEA participates, to provide an up-to-date summary in this area including progress in developing understanding of how radiological protection of the environment will be implemented for the three exposure situations - planned, existing and emergency - set out in the new ICRP general recommendations. Furthermore, given the NEA's mandate across the civil nuclear energy field, the paper will give a commentary on how developments in radiological protection of the environment may interplay with new build of nuclear power plants. (author)

  3. Development of working methods used inside reactor pressure vessel at Oskarshamn from the radiation protection point of view

    International Nuclear Information System (INIS)

    Solstrand, C.

    2005-01-01

    When performing maintenance and repair work in the beginning of 1970's conventional work tools and working methods mostly were used. The focus was on how to protect workers sufficiently in a proper way to keep the doses ALARA. During the last years the focus has turned more towards construction of special work tools in order to minimise personnel doses without any special arrangements for radiation shielding. Three examples will be presented to show that the optimisation of radiation protection can lead to the development of work tools and working methods, reducing the doses, time and money. (author)

  4. A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

    International Nuclear Information System (INIS)

    Birch, M.; Brocklehurst, J.E.

    1987-12-01

    The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)

  5. A feasibility study on long-life reduced-moderation water reactor with highly protected Pu breeding by doping with minor actinides

    International Nuclear Information System (INIS)

    Hamase, Erina; Damian, Frederic; Poinot-salanon, Christine; Saito, Masaki; Sagara, Hiroshi; Han, Chi Young

    2013-01-01

    Highlights: ► The present paper is focused on a reduced-moderation water reactor. ►237 Np or 241 Am was doped into the inner blanket. ►238 Pu transmuted from MA works as a fissionable nuclide. ► It also contributes to increasing the proliferation resistance of Pu. ► Long-life core and highly protected Pu breeding were simultaneously achieved. - Abstract: The present paper is focused on the feasibility of reduced-moderation water reactor (RMWR) which could simultaneously achieve the extension of core life-time, and highly protected plutonium (Pu) breeding performance with short compound system doubling time (CSDT) by neptunium-237 ( 237 Np) or americium-241 ( 241 Am) doping of the inner blanket of the RMWR. As a preliminary analysis of the RMWR, a simplified fuel pin configuration was analyzed. In case of 60% doping of 237 Np and 241 Am in the inner blanket, the maximum available effective full power days (EFPDs) were much longer and were about 15,100 and 14,200 EFPDs, respectively. For the proliferation resistance of Pu produced in the blanket, the total Pu in the inner and lower/upper blanket was below the practically unusable criterion for an explosive device proposed by Pellaud, and was technically unfeasible for high-technology hypothetical nuclear explosive devices (HNEDs) proposed by Kessler and Kimura. The protected Pu breeding performance was analyzed and in case of 60% doping of 237 Np and 241 Am, CSDT was short by approximately 40 and 50 years. Therefore, the feasibility of RMWR was shown which could simultaneously achieve the extension of core life-time and have highly protected Pu breeding performance with short CSDT by doping the inner blanket with minor actinides (MAs). The objective of this study was thus to evaluate the performance of the concept with respect to the core life-time, proliferation resistance of Pu and CSDT. The technological feasibility of such core concept will have to be evaluated by further dedicated analyses

  6. Radiological protection considerations during the treatment of glioblastoma patients by boron neutron capture therapy at the high flux reactor in Petten, The Netherlands

    International Nuclear Information System (INIS)

    Moss, R.L.; Rassow, J.; Finke, E.; Sauerwein, W.; Stecher-Rasmussen, F.

    2001-01-01

    A clinical trial of Boron Neutron Capture Therapy (BNCT) for glioblastoma patients has been in progress at the High Flux Reactor (HFR) at Petten since October 1997. The JRC (as licence holder of the HFR) must ensure that radiological protection measures are provided. The BNCT trial is a truly European trial, whereby the treatment takes place at a facility in the Netherlands under the responsibility of clinicians from Germany and patients are treated from several European countries. Consequently, radiological protection measures satisfy both German and Dutch laws. To respect both laws, a BNCT radioprotection committee was formed under the chairmanship of an independent radioprotection expert, with members representing all disciplines in the trial. A special nuance of BNCT is that the radiation is provided by a mixed neutron/gamma beam. The radiation dose to the patient is thus a complex mix due to neutrons, gammas and neutron capture in boron, nitrogen and hydrogen, which, amongst others, need to be correctly calculated in non-commercial and validated treatment planning codes. Furthermore, due to neutron activation, measurements on the patient are taken regularly after treatment. Further investigations along these lines include dose determination using TLDs and boron distribution measurements using on-line gamma ray spectroscopy. (author)

  7. Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation with the Russian Federation, Ukraine, Armenia, Georgia and Belarus

    International Nuclear Information System (INIS)

    Dassen, Lars van; Andersson, Sarmite; Bejarano, Gabriela; Delalic, Zlatan; Ekblad, Christer; German, Olga; Grapengiesser, Sten; Karlberg, Olof; Olsson, Kjell; Sandberg, Viviana; Stenberg, Tor; Turner, Roland; Zinger, Irene

    2010-06-01

    The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral cooperation with Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in a number of projects financed by the European Union. This report gives an overview of the cooperation projects in 2009 as well as the framework in which they are performed. Summaries of each project are given in an Appendix. The project managers in the Section for Cooperation and Development in the Department of International Affairs are responsible for the cooperation projects and the implementation of the bilateral programmes. But the positive outcome of the projects is also dependent on a large number of experts at SSM who work with the regulatory functions in the nuclear and radiation protection fields in a Swedish context as well as on external consultants. Together, their experience is invaluable for the implementation of the projects. But the projects also give experience of relevance for the SSM staff.

  8. Cooperation in Nuclear Waste Management, Radiation Protection, Emergency Preparedness, Reactor Safety and Nuclear Non-Proliferation with the Russian Federation, Ukraine, Armenia, Georgia and Belarus.

    Energy Technology Data Exchange (ETDEWEB)

    Dassen, Lars van; Andersson, Sarmite; Bejarano, Gabriela; Delalic, Zlatan; Ekblad, Christer; German, Olga; Grapengiesser, Sten; Karlberg, Olof; Olsson, Kjell; Sandberg, Viviana; Stenberg, Tor; Turner, Roland; Zinger, Irene

    2010-06-15

    The Swedish Radiation Safety Authority (SSM) is trusted with the task of implementing Sweden's bilateral cooperation with Russia, Ukraine, Georgia, Belarus and Armenia in the fields of reactor safety, nuclear waste management, nuclear non-proliferation as well as radiation protection and emergency preparedness. In these fields, SSM also participates in a number of projects financed by the European Union. This report gives an overview of the cooperation projects in 2009 as well as the framework in which they are performed. Summaries of each project are given in an Appendix. The project managers in the Section for Cooperation and Development in the Department of International Affairs are responsible for the cooperation projects and the implementation of the bilateral programmes. But the positive outcome of the projects is also dependent on a large number of experts at SSM who work with the regulatory functions in the nuclear and radiation protection fields in a Swedish context as well as on external consultants. Together, their experience is invaluable for the implementation of the projects. But the projects also give experience of relevance for the SSM staff.

  9. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

  10. Principles of Inherent Self-Protection Realized in the Project of Small Size Modular Reactor SVBR-100

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Petrochenko, V.V.; Komlev, O.G.; Tormyshev, I.V.; Dedul, A.V.

    2013-01-01

    Conclusions: • From the standpoint of population, the opportunity of catastrophic consequences caused by nuclear accident is much more important than very low possibility of its realization. • The level of social acceptability of future large-scale NP must be higher. • Use of the nuclear power technology (NPT) based on RFs, in which the value of stored potential energy of different kinds is minimal, will meet that goal the most efficient. • Those RFs cannot amplify the external impacts, therefore, the scale of damages will be only determined by the external impact energy, the exhaust of radioactivity being localized. • Now there are no developed NPTs with such properties. • The NPT based on modular fast reactors SVBR-100 and verified in conditions of NS operating is to the most extent ready to be demonstrated. • Federal target program “New Generation Nuclear Power Technologies for 2010 – 2015 Years and Future Trends up to 2020” stipulates the construction of experimental-industrial power-unit SVBR-100. • The project is realized within the frameworks of state-private partnership by joint venture JSC “AKME-Engineering” organized on a parity basis by State Atomic Energy Corporation “Rosatom” and Limited Liability Company “Irkutskenergo”. • The first of a kind power unit with RF SVBR-100 will be commissioned in 2017 near the SSC NIIAR site in Dimitrovgrad (Ulyanovsk region). • Widespread common use of this NPT, which potentials are very high, is expected to begin in ∼ 2020 – 2025. In case of earlier starting, the economic risk will be high; in case it is launched later, much profit will be lost

  11. A practical methodology of radiological protection for the reduction of hot particles in BWR type reactors; Una metodologia practica de proteccion radiologica para la reduccion de particulas calientes en reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Alvarez G, G [Comision Federal de Electricidad, Gerencia del Proyecto Nucleoelectrico Laguna Verde, Disciplina de Fisica Aplicada (Mexico)

    1991-07-01

    The purpose of this work, in general form, is to describe a practical method for reduction of hot particles generated as consequence of the operational activities of BWR nuclear reactors. This methodology provides a description of the localizations and/or probable activities of finding particles highly radioactive denominated hot particles. For this purpose it was developed a strategy based on the decontamination lineaments, as well as the manipulation, gathering, registration, contention, documentation, control and final disposition of the hot particles. In addition, some recommendations are reiterated and alternative, in order to gathering the hot particles in a dynamic way given to the activities of the personal occupationally exposed in highly radioactive areas. The structure of the methodology of hot particles is supported in the radiological controls based on the Code of Federal Regulation 10 CFR 20 as well as the applicable regulatory documents. It provides an idea based on administrative controls of radiological protection, in order to suggesting the responsibilities and necessary directing for the control of the hot particles required in nuclear plants of the BWR type. (author)

  12. Tasks related to increase of RA reactor exploitation and experimental potential, 01. Designing the protection chamber in the RA reactor hall for handling the radioactive experimental equipment (I-II) Part II, Vol. II; Radovi na povecanju eksploatacionih i eksperimentalnih mogucnosti reaktora RA, 01. Projektovanje zastitne komore u hali reaktora RA za rad sa aktivnim eksperimentalnim uredjajima (I-II), II Deo, Album II

    Energy Technology Data Exchange (ETDEWEB)

    Pavicevic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1963-07-15

    This second volume of the project for construction of the protection chamber in the RA reactor hall for handling the radioactive devices includes the technical description of the chamber, calculation of the shielding wall thickness, bottom lead plate, horizontal stability of the chamber, cost estimation, and the engineering drawings.

  13. Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    1) The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2) Starting from this concept, we endeavoured to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3) Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author) [French] 1) La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2) A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3) Enfin une methode de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

  14. Methods and experimental coefficients used in the computation of reactor shielding; Methodes et coefficients experimentaux pour le calcul des protections de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)Fren. [French] 1. La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2. A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3. Enfin une mde de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)

  15. Actions to Protect the Public in an Emergency due to Severe Conditions at a Light Water Reactor. Date Effective: May 2013 (Russian Edition)

    International Nuclear Information System (INIS)

    2015-01-01

    Under Article 5.a(ii) of the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency (the 'Assistance Convention'), one function of the IAEA is to collect and disseminate to States Parties and Member States information concerning methodologies, techniques and results of research relating to response to a nuclear or radiological emergency. This publication is intended to help fulfil in part these functions assigned to the IAEA in the Assistance Convention. The aim of this publication is to provide those persons who are responsible for making and for acting on decisions in the event of an emergency at a light water reactor with an understanding of the actions that are necessary to protect the public. The publication provides a basis for developing the tools and criteria at the preparedness stage that would be needed in taking protective actions and other actions in response to such an emergency. The publication applies the safety principles stated in IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles, and it will be of assistance to Member States in meeting the requirements established in IAEA Safety Standards Series No. GS-R-2, Preparedness and Response for a Nuclear or Radiological Emergency. The application of these requirements is intended to minimize the consequences for people and the environment in any nuclear or radiological emergency. This guidance should be adapted to fit the State's organizational arrangements, language, terminology, concept of operation and capabilities. The IAEA General Conference, in resolution GC(55)/RES/9: 'Emphasizes the importance for all Member States to implement emergency preparedness and response mechanisms and develop mitigation measures at a national level, consistent with the Agency's Safety Standards, for improving emergency preparedness and response, facilitating communication in an emergency and contributing to harmonization of national criteria for

  16. Actions to Protect the Public in an Emergency due to Severe Conditions at a Light Water Reactor. Date Effective: May 2013 (Spanish Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    Under Article 5.a(ii) of the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency (the 'Assistance Convention'), one function of the IAEA is to collect and disseminate to States Parties and Member States information concerning methodologies, techniques and results of research relating to response to a nuclear or radiological emergency. This publication is intended to help fulfil in part these functions assigned to the IAEA in the Assistance Convention. The aim of this publication is to provide those persons who are responsible for making and for acting on decisions in the event of an emergency at a light water reactor with an understanding of the actions that are necessary to protect the public. The publication provides a basis for developing the tools and criteria at the preparedness stage that would be needed in taking protective actions and other actions in response to such an emergency. The publication applies the safety principles stated in IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles, and it will be of assistance to Member States in meeting the requirements established in IAEA Safety Standards Series No. GS-R-2, Preparedness and Response for a Nuclear or Radiological Emergency. The application of these requirements is intended to minimize the consequences for people and the environment in any nuclear or radiological emergency. This guidance should be adapted to fit the State's organizational arrangements, language, terminology, concept of operation and capabilities. The IAEA General Conference, in resolution GC(55)/RES/9: 'Emphasizes the importance for all Member States to implement emergency preparedness and response mechanisms and develop mitigation measures at a national level, consistent with the Agency's Safety Standards, for improving emergency preparedness and response, facilitating communication in an emergency and contributing to harmonization of national criteria for protective and other

  17. Actions to Protect the Public in an Emergency due to Severe Conditions at a Light Water Reactor. Date Effective: May 2013

    International Nuclear Information System (INIS)

    2013-01-01

    Under Article 5.a(ii) of the Convention on Assistance in the Case of a Nuclear Accident or Radiological Emergency (the 'Assistance Convention'), one function of the IAEA is to collect and disseminate to States Parties and Member States information concerning methodologies, techniques and results of research relating to response to a nuclear or radiological emergency. This publication is intended to help fulfil in part these functions assigned to the IAEA in the Assistance Convention. The aim of this publication is to provide those persons who are responsible for making and for acting on decisions in the event of an emergency at a light water reactor with an understanding of the actions that are necessary to protect the public. The publication provides a basis for developing the tools and criteria at the preparedness stage that would be needed in taking protective actions and other actions in response to such an emergency. The publication applies the safety principles stated in IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles, and it will be of assistance to Member States in meeting the requirements established in IAEA Safety Standards Series No. GS-R-2, Preparedness and Response for a Nuclear or Radiological Emergency. The application of these requirements is intended to minimize the consequences for people and the environment in any nuclear or radiological emergency. This guidance should be adapted to fit the State's organizational arrangements, language, terminology, concept of operation and capabilities. The IAEA General Conference, in resolution GC(55)/RES/9: 'Emphasizes the importance for all Member States to implement emergency preparedness and response mechanisms and develop mitigation measures at a national level, consistent with the Agency's Safety Standards, for improving emergency preparedness and response, facilitating communication in an emergency and contributing to harmonization of national criteria for protective and other

  18. Nuclear reactors for the future

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Kamble, M.T.; Dulera, I.V.

    2013-01-01

    For the sustainable development of nuclear power plants with enhanced safety features, economic competitiveness, proliferation resistance and physical protection, several advanced reactor developments have been initiated world-wide. The major advanced reactor initiatives and the proposed advanced reactor concepts have been briefly reviewed along with their advantages and challenges. Various advanced reactor designs being pursued in India have also been briefly described in the paper. (author)

  19. RB research nuclear reactor, Annual report for 1984, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

    1984-01-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

  20. Nuclear Reactor RA Safety Report, Vol. 14, Safety protection measures; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 14, Sigurnosne zastitne mere

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-01

    Nuclear reactor accidents can be caused by three type of errors: failure of reactor components including (1) control and measuring instrumentation, (2) errors in operation procedure, (3) natural disasters. Safety during reactor operation are secured during its design and construction and later during operation. Both construction and administrative procedures are applied to attain safe operation. Technical safety features include fission product barriers, fuel elements cladding, primary reactor components (reactor vessel, primary cooling pipes, heat exchanger in the pump), reactor building. Safety system is the system for safe reactor shutdown and auxiliary safety system. RA reactor operating regulations and instructions are administrative acts applied to avoid possible human error caused accidents. [Serbo-Croat] Uzroci udesa na nuklearnim reaktorima mogu se svrstati u jednu od sledece tri grupe: (1) otkaz pojedinih delova opreme, ukljucujuci mernu i kontrolnu instrumentaciju, (2) greske u pogonu i eksploataciji, (3) prirodne nepogode, katastrofe. Bezbednost i sigurnost u radu nuklearnog reaktora osiguravaju se odredjenim merama koje se preduzimaju pri njegovoj izgradnji i kasnije njegovoj elsploataciji. Te mere se mogu podeliti u sledece dve kategorije: (1) mere tehnicke zastite, i (2) administrativne mere. Mere tehnicke zastite sastoje se od barijere fissionih produkata, kosuljice gorivnih elemenata, primarnog kola reaktora RA (reaktorski sud, cevovod primarnog kola, toploizmenjivac u pimpi), zgrada reaktora. Sigurnosni sistem cini sistem za sigurnosno zaustavljanje reaktora i pomocni sigurnosni sistem. Kroz odgovarajuce propise i uputstva za rad na reaktoru RA primenjene su administrativne mere neophodne za sprecavanje udes koji bi mogao nastati kao posledica ljudskog faktora.

  1. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

  2. 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

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

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

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

  4. Guidance for the application of an assessment methodology for innovative nuclear energy systems. INPRO manual - Physical protection. Vol. 6 of the final report of phase 1 of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2008-11-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in the year 2000, based on a resolution of the IAEA General Conference (GC(44)/RES/21). The main objectives of INPRO are (1) to help to ensure that nuclear energy is available to contribute in fulfilling energy needs in the 21st century in a sustainable manner, (2) to bring together both technology holders and technology users to consider jointly the international and national actions required to achieve desired innovations in nuclear reactors and fuel cycles; and (3) to create a forum to involve all relevant stakeholders that will have an impact on, draw from, and complement the activities of existing institutions, as well as ongoing initiatives at the national and international level. This document follows the guidelines of the INPRO report M ethodology for the assessment of innovative nuclear reactors and fuel cycles, Report of Phase 1B (first part) of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) , IAEA-TECDOC-1434 (2004), together with its previous report G uidance for the evaluation for innovative nuclear reactors and fuel cycles, Report of Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), IAEATECDOC-1362 (2003). This INPRO manual is comprised of an overview volume and eight additional volumes covering the areas of economics (Volume 2), infrastructure (Volume 3), waste management (Volume 4), proliferation resistance (Volume 5), physical protection (Volume 6), environment (Volume 7), safety of reactors (Volume 8), and safety of nuclear fuel cycle facilities (Volume 9). The INPRO Manual for the area of physical protection (Volume 6) provides guidance to the assessor of an INS (innovative nuclear energy system) under a physical protection regime in a country that is planning to install a nuclear power program (or maintaining or enlarging an existing one), and describes the application of the

  5. Remote level radiation monitoring system for the brazilian IEA-R1 nuclear research reactor for routine radiation protection procedures and as a support tool in case of radiological emergency

    International Nuclear Information System (INIS)

    Cardenas, Jose P.N.; Romero Filho, Christovam R.; Madi Filho, Tufic

    2008-01-01

    Nuclear facilities must monitoring radiation levels to establish procedures for radiological protection staff involving workers and the public. The Instituto de Pesquisas Energeticas e Nucleares - IPEN has 5 important plants and in case of accident in one of them, the Institute keeps operational an Emergency Response Plan (ERP). This document (ERP) is designed to coordinate all procedures to assure safe and secure conditions for workers, environment and the public. One of this plants is the IEA-R1 reactor, it is the oldest nuclear research reactor (pool type) in Latin America, reached it first criticality in September of 1957. The reactor is used 60 hours/week with continuous operation and with nominal power of 3.5 MW, with technical conditions to operate at 5 MW thermal power. This reactor has a Radiological Emergency Plan that establishes the implementation of rules for workers and people living at the exclusion area in the case of an emergency situation. This paper aims to describe the implementation of a computational system developed for remote radiation monitoring, in a continuous schedule of IEA-R1 nuclear research reactor containment building. Results of this action can be used as a support mean in a radiological emergency. All necessary modules for radiation detection, signals conditioners and processing, data acquisition board, software development and computer specifications are described. The data acquisition system operating in the reactor shows readings concerned to radiation environment such as activity, doses and concentration in real time and displays a periodical data bank (Data Base) of this features allowing through the surveillance of the operation records anytime, leading to studies and analysis of radiation levels. Results of this data acquisition are shown by means of computer graphics screens developed for windows environment using Visual Basic software. (author)

  6. Oregon State University TRIGA Reactor annual report

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-31

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

  7. Oregon State University TRIGA Reactor annual report

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  8. H Reactor

    Data.gov (United States)

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

  9. Official announcement of the directive on protection of nuclear power plant equipped with LWR-type reactors from human intrusion or other interference by third parties. Announcement of BMU (German Federal Ministry Environment), of 6 Dec. 1995 - RS I 3 13151 - 6/14

    International Nuclear Information System (INIS)

    1996-01-01

    An operating permit for a nuclear power plant is to be granted only if the applicant and facility operator presents evidence guaranteeing the legally required physical protection and other security measures for protection from human instrusion and other type of interference. As a basis for review and licensing, the competent authorities in 1987 have issued a directive specifying the requirements to be met for physical protection of nuclear power plant equipped with PWR-type reactors, and in 1994 followed a second, analogous directive relating to nuclear power plant with BWR-type reactors. The directive now announced for physical protection of nuclear power plant equipped with LWR-type reactors combines and replaces the two former ones, and from the date of the announcement is the only applicable directive. The text of the directive is not reproduced for reasons of secrecy protection. (orig./CB) [de

  10. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  11. Research reactor DHRUVA

    International Nuclear Information System (INIS)

    Veeraraghaven, N.

    1990-01-01

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

  12. KS-150 reactor control

    International Nuclear Information System (INIS)

    Wagner, K.

    1974-01-01

    A thorough description is presented of the control and protection system of the Bohunice A-1 reactor. The system including auxiliary facilities was developed, manufactured and installed at the reactor by the SKODA Works, Plzen. The system parameters are listed and a brief account is also given of the development efforts and of the physical and power start-up of the A-1 nuclear power plant. (L.O.)

  13. Power reactor design trends

    International Nuclear Information System (INIS)

    Hogan, W.J.

    1985-01-01

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

  14. RB Research nuclear reactor, Annual report for 2005

    International Nuclear Information System (INIS)

    Milosevic, M.; Dasic, N.; Ljubenov, V.; Pesic, M.; Nikolic, D; Jevremovic, M.; Minic, D.

    2006-01-01

    Report on RB reactor operation during 2005 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2005

  15. Radiation protection at the RA Reactor in 1995, Part -2, Annex 2, Decontamination and actions, collection of liquid effluents and solid radioactive waste; Deo 2 - Prilog 2 - Dekontaminacija i intervencije, skupljanje tecnih efluenata i cvrstih radioaktivnih otpadnih materijala

    Energy Technology Data Exchange (ETDEWEB)

    Mandic, M; Vukovic, Z; Lazic, S; Plecas, I; Voko, A [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1995-12-01

    Certain amount of solid waste results from RA reactor operation, the mean quantity of which depends on the duration of reactor operation and related activities. During repair, when reactor is not operated as well under accidental conditions, the quantity of waste is higher, dependent on the type of repair and comprehensiveness of decontamination of the working surface, contaminated tools and components. The waste is sorted and packed on the spot where they appeared according to the existing regulations and principles of radiation protection with aim to minimize unnecessary exposure of the radiation protection personnel who deals with control, transport, radioactive waste treatment and decontamination. During exceptional operations (decontamination, repair, bigger volume of contaminated material, etc.) professional staff of the Radiation protection department gives recommendations and helps in planning the actions related to repair, sorting and packaging of radioactive waste in special containers, identification of the contaminants, etc. [Serbo-Croat] Tokom rada reaktora RA dolazi do stvaranja odredjenih cvrstih otpadnih materijala cija prosecna kolicina zavisi od vremena rada reaktora i aktivnosti koje se tamo obavljaju. Tokom remonta, kada reaktor ne radi kao i pri akcidentalnim situacijama nastaju vece kolicine otpadnih materijala koje zavise od obima i vrste remontnih operacija i obima dekontaminacije kontaminirane radne povrsine i kontaminiranog alata, predmeta, opreme, itd. Nastali otpadni materijali se razvrstavaju i pakuju na mestu nastanka prema odgovarajucim propisima u skladu sa principima zastite od zracenja i aspekta bezbednosti u cilju minimiziranja nepotrebnog ozracivanja ljudstva za preuzimanje, kontrolu, transport, naknadnu obradu RAO i dekontaminaciju. Pri nerutinskim operacijama (dekontaminacija, remont, kontaminiarni otpadni materijal velike zapremine i sl.), strucna sluzba Institita ZASTITA pruza strucne konsultacije i pomaze pri planiranju

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

    International Nuclear Information System (INIS)

    Richards, W.J.

    1980-01-01

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

  17. A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods; Um estudo de arquiteturas de hardware para aplicacao em sistemas digitais de protecao de reatores nucleares - metodos de analise de confiabilidade e seguranca

    Energy Technology Data Exchange (ETDEWEB)

    Benko, Pedro Luiz

    1997-07-01

    A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

  18. Comparison of the worth of control and protection system rods of different design on the basis of the measurements in BN-600 reactor

    International Nuclear Information System (INIS)

    Vasilyev, B.A.; Roslyakov, V.F.; Farakshin, M.R.

    1988-01-01

    The results of the worth measurements of the basic and experimental absorbing rods of BN-600 reactor are presented. The procedure used for the rods worth comparison on the basis of calculated and experimental data interpretation is described here. Basic and experimental rods relative worth is also presented. (author). 5 refs, 3 figs, 2 tabs

  19. Strengthening DiD in Emergency Preparedness and Response by Pre-Establishing Tools and Criteria for the Effective Protection of the Public During a Severe Emergency at a Light Water Reactor or its Spent Fuel Pool

    Energy Technology Data Exchange (ETDEWEB)

    Mckenna, T.; Welter, P. Vilar; Callen, J.; Buglova, E., E-mail: T.Mckenna@iaea.org [International Atomic Energy Agency (IAEA), Department of Nuclear Safety and Security, Wagramer Strasse 5, P.O. Box 100, 1400 Vienna (Austria)

    2014-10-15

    Defence in depth can be divided into two parts: first, to prevent accidents and, second, if prevention fails, to limit their consequences and prevent any evolution to more serious conditions. This paper will cover the second part, by providing tools and criteria to be used during a severe emergency to limit the consequences to the public from a severe accident. Severe radiation-induced consequences among the public off-site are only possible if there is significant damage to fuel in the reactor core or spent fuel pools. Consequently, the tools and criteria have been specifically developed for individuals responsible for making and for acting on decisions to protect the public in the event of an emergency involving actual or projected severe damage to the fuel in the reactor core or spent fuel pool of a light water reactor (LWR). These tools and criteria, developed by the IAEA’s Incident and Emergency Centre (IEC), will facilitate the implementation of the ‘Emergency Response’ defence in depth concept. (author)

  20. Heavy water reactors physics; Physique des reacteurs a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Girard, Y; Lourme, P; Naudet, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    An important research programme on heavy water reactor physics has been carried out in France for quite a few years. The decision to build the EL 4 prototype and so to choose the heavy water gas cooled type has renewed the interest in this programme and at the same time given to it a more specific orientation A summary of the results gained in this field is presented in this paper. In the first part are described the experimental investigations, most of them were carried out in the criticality facility AQUILON II. The experiments are grouped in four parts - Systematic studies of lattices Buckling measurements. - Specific studies of gas-cooled lattices. - Fine structure, spectral indices measurements etc... - Measurements on lattices or samples containing Uranium of various enrichment or Plutonium. The second part is devoted to a summary of the theoretical studies. The whole results have allowed an improvement of the calculation methods, have led to a better understanding of the neutron balance in lattices, and have permitted the establishment of a set of formula to predict not only the clean fuel conditions but also the evolution of the nuclear properties with irradiation. Some specific studies on power reactor are quoted. (authors) [French] Un important programme d'etudes sur la physique des reacteurs a eau lourde est mene en France depuis assez longtemps. La decision de construire le prototype EL 4 et de s'engager ainsi dans la filiere des reacteurs a eau lourde refroidis par gaz a redonne un nouvel interet a ce programme et l'a en meme temps oriente dans une direction plus particuliere. La presente communication, rassemble les resultats des etudes faites dans ce domaine depuis la derniere conference de Geneve. Dans la premiere partie on decrit les etudes experimentales dont la plupart ont ete effectuees dans la pile d'experiences critiques Aquilon II. Les experiences sont groupees en quatre ensembles: etude systematique de reseaux (mesures de laplaciens) etudes

  1. Control of WWER-440 nuclear reactor

    International Nuclear Information System (INIS)

    Wagner, K.; Drab, F.; Grof, V.

    1978-01-01

    The V-1 reactor control systems are described. The data acquisition and processing system fulfils four main functions, ie., reactor start-up and power increase to 10% of the rated power, automatic power control within 3% and 110% of the rated power, reactivity compensation, and reactor protection. The automatic control system ensures constant steam pressure maintained with an accuracy of +-0.05 MPa. Reactivity compensation and spatial power distribution is mainly safeguarded by boric acid control. The V-1 reactor protection system has four levels of accident protection depending on the gravity of the failure. The philosophy of automation of the V-1 reactor control and protection system is based on autonomous automatic controlers and on the direct control of the individual sets and technological equipment. In conclusion, development trends are briefly outlined of control and protection systems of light water reactor power plants. (Z.M.)

  2. The safety of fast reactors

    International Nuclear Information System (INIS)

    Justin, F.

    1976-01-01

    A response is made to the main questions that a man in the street may arise concerning fast breeder reactors, in particular: the advantages of this line, dangerous materials contained in fast breeder reactors, containment shells protecting the environment from radiations, main studies now in progress [fr

  3. The implementation and evaluation of physical protection system of the IEA-R1 reactor; Implementacao e avaliacao do sistema de protecao fisica do reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Vaz, Antonio Carlos Alves

    2016-11-01

    The September 11, 2001 terrorist attacks in New York, the accident at the Fukushima nuclear power plant on March 2011 and the recent attacks in Paris on November 2015 are examples of events that justify the efforts of the International Agency of Energy Atomic - IAEA to improve security at nuclear facility. The Brazilian government has been collaborating with this project and investing resources to improve the Physical Protection System - PPS of the nuclear research reactor system, technically is associated with the elements of detection, delay and response. The PPS is an integrated system of people, equipment and procedures used to protect nuclear facilities and radioactive sources against threat, theft or sabotage. The PPS works to avoid, to mitigate or to minimize the consequences caused by these actions. This study evaluates the PPS of the reactor, identifying the vulnerabilities and suggesting ways to improve the system effectiveness. The analyses were based on the methodology developed by Sandia National Laboratories´ security experts in Albuquerque - USA, allowing the system evaluation through hypothetical and probabilistic analyzes; identifying threats, determining the targets and analyzing the possible adversaries paths. From the methodology adopted was obtained the value around 40% for PE indicator, which shows the need to improve the system to minimizing the vulnerabilities. (author)

  4. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

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

  5. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  6. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

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

  7. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

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

  8. Physical security at research reactors

    International Nuclear Information System (INIS)

    Clark, R.A.

    1977-01-01

    Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems

  9. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ; Diseno, construccion e implementacion de dos circuitos redundantes de la logica de actuacion del sistema de proteccion de la nueva consola de control del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Celestino M, E.

    2016-07-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  10. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  11. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

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

  12. Radiation protection at the RA nuclear reactor in 1987, Part II.a. Control of radioactivity in the environment of the RA nuclear reactor (precipitation, fallout, water, soil, plants, fruit)

    International Nuclear Information System (INIS)

    Martic, M.; Ajdacic, N.; Jovanovic, J.

    1987-01-01

    Control of radioactivity in the biosphere in the vicinity of the RA reactor is part of the radioactivity control done regularly for the whole territory of the Vinca Institute. According to the measurements during 1987 it was found that the total contamination of the precipitation was highest in January compared to the period before Chernobylsk accident. Mean monthly value of the total beta cavity was highest in April 5.41 times higher than the relevant value in 1986. This is a preliminary report, the measurement data will be presented after after analysis in the annual report [sr

  13. EPR (European Pressurized Reactor)

    International Nuclear Information System (INIS)

    2015-01-01

    This document presents the EPR (European Pressurized Reactor), a modernised version of PWRs which uses nuclear fission. It indicates to which category it belongs (third generation). It briefly describes its operation: recalls on nuclear fission, electricity production in a nuclear reactor. It presents and comments its characteristics: power, thermal efficiency, redundant systems for safety control, double protective enclosure, expected lifetime, use of MOX fuel, modular design. It discusses economic stakes (expected higher nuclear electricity competitiveness, but high construction costs), and safety challenges (design characteristics, critics by nuclear safety authorities about the safety data processing system). It presents the main involved actors (Areva, EDF) and competitors in the field of advanced reactors (Rosatom with its VVER 1200, General Electric with its ABWR and its ESBWR, Mitsubishi with its APWR, Westinghouse with its AP100) while outlining the importance of certifications and delays to obtain them. After having evoked key data on EPR fuel consumption, it indicates reactors under construction, evokes potential markets and perspectives

  14. Elements on reactor control

    International Nuclear Information System (INIS)

    Bruna, G.B.

    1998-01-01

    In order to achieve the two-fold goal of maximizing the energy obtained from reactor fuel and ensuring the large flexibility of plant operation in respect to safety regulations and keeping the reactor integrity the control of PWRs is generally based on real time monitoring and analysing of independent neutronic parameters: thermal power release, axial power distribution in the core and temperatures of the primary loop. Two control chains more or less coupled according to the control chosen mode are in charge of the control of these parameters. With the brief history of control in French power reactors the advanced X control mode adopted by Framatome for N4 plants is described in detail. A summary of N4 reactor control and protection system is included

  15. Advanced spheromak fusion reactor

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1996-01-01

    The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid 'blanket' that breeds tritium, converts neutron kinetic energy to heat, and protects the reactor vessel from severe neutron damage. However, the physics is more complex, so that considerable research is required to learn how to achieve the reactor potential. Critical physics problems and possible ways of solving them are described. The opportunities and issues associated with a possible liquid wall are considered to direct future research

  16. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1988

    International Nuclear Information System (INIS)

    1989-11-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  17. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1987

    International Nuclear Information System (INIS)

    1988-06-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  18. Surveys of research projects concerning nuclear facility safety financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1991

    International Nuclear Information System (INIS)

    1992-09-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  19. Anticorrosional protection in nuclear power station objects

    International Nuclear Information System (INIS)

    Czarnocki, A.; Kwiatkowski, A.

    1976-01-01

    The distribution and qualities of chemical protection and demands concerning preparation of the bottom for protecting coats in nuclear power station objects are discussed. The solutions of protections applied abroad and in the objects of ''MARIA'' reactor are presented. (author)

  20. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

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

  1. Thermal Reactor Safety

    International Nuclear Information System (INIS)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods

  2. Thermal Reactor Safety

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-01

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  3. RB Research nuclear reactor, Annual report for 2004

    International Nuclear Information System (INIS)

    Dasic, N.; Pesic, M.; Nikolic, D; Jevremovic, M.; Eskirovic, B.

    2005-02-01

    Report on RB reactor operation during 2004 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. It contains data about reactor operation during previous 8 years. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation, Annex 1. contains data about heavy water degradation, and Annex 2 is the certificate about the crane bridge in the reactor hall

  4. RB research nuclear reactor, Annual report for 1983, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

    1983-01-01

    The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

  5. Depth protection system

    International Nuclear Information System (INIS)

    Arita, Setsuo; Izumi, Shigeru; Suzuki, Satoru; Noguchi, Atomi.

    1988-01-01

    Purpose: To previously set a nuclear reactor toward safety side by the reactor scram if an emergency core cooling system is failed to operate. Constitution If abnormality occurs in an emergency core cooling system or an aqueous boric acid injection system, a reactor protection system is operated and, if the reactor protection system shows an abnormal state, a control rod withdrawal inhibition system is operated as a fundamental way. For instance, when the driving power source voltage for the emergency core cooling system is detected and, if it is lower than a predetermined value, the reactor protection system is operated. Alternatively, if the voltage goes lower than the predetermined value, the control rod withdrawal is inhibited. In addition, stopping for the feedwater system is inhibited. Further, integrity of the driving means for the emergency core cooling system is positively checked and the protection function is operated depending on the result of check. Since the nuclear reactor can be set toward the safety side even if the voltage for the driving power source of the aqueous boric acid injection system is lower than a predetermined value, the reactor safety can further be improved. (Horiuchi, T.)

  6. 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.)

  7. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

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

  8. Research nuclear reactor RA - Annual Report 1989

    International Nuclear Information System (INIS)

    Sotic, O.

    1989-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  9. RA Research nuclear reactor - Annual report 1987

    International Nuclear Information System (INIS)

    1987-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  10. RA Research reactor, Annual report 1988

    International Nuclear Information System (INIS)

    Sotic, O.

    1988-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  11. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

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

  12. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  13. RB research nuclear reactor, Annual report for 1989, I - III

    International Nuclear Information System (INIS)

    Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1989-12-01

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

  14. The manual of coldness engineering; Formulaire du froid

    Energy Technology Data Exchange (ETDEWEB)

    Rapin, P.; Jacquard, P.

    2001-07-01

    This book is a compilation of theoretical and practical data which allow the design, dimensioning, installation and maintenance of refrigerating systems for the industry and buildings. This 11. edition comprises several updates in particular in the domain of refrigerating fluids (environmental problems), technology of systems (automatisms, electrical devices), and fluidic and electrical schemes. Content: introduction, coldness production, technology of refrigerating machineries, automatisms, isothermal constructions and refrigerating statuses, applications of coldness, apparatuses, implementation, appendixes. (J.S.)

  15. Guides et formulaires | CRDI - Centre de recherches pour le ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    Demande de subvention de recherche du CRDI · Budget de proposition · Lignes directrices du CRDI pour la préparation du rapport d'étape technique · Lignes directrices du CRDI pour la préparation du rapport technique final · Lignes directrices du CRDI pour les dépenses de projet admissibles · Lignes directrices pour la ...

  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. Necessity of research reactors

    International Nuclear Information System (INIS)

    Ito, Tetsuo

    2016-01-01

    Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)

  18. Fourth Generation Reactor Concepts

    International Nuclear Information System (INIS)

    Furtek, A.

    2008-01-01

    were selected to generation IV by the GIF to further studies: Gas-Cooled Fast Reactor (GFR), Lead-Cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), Sodium-Cooled Fast Reactor (SFR), Supercritical Water-Cooled Reactor (SCWR), Very High Temperature Reactor (VHTR). These six systems would each need a dedicated effort in research and development. Some consideration for the fuel and recycling technology are common and can be shared. These common areas encompass: fuel cycles, fuels and materials choice, energy products, risk and safety, economics and proliferation and physical protection concerns.(author)

  19. Radiation protection at the RA reactor in 1987, Part I: Control of the working environment - dosimetry and radiation protection at the RA reactor, Annex 1; Prilog 1, Zastita od zracenja kod reaktora RA u 1987. godini - Deo 1: Kontrola radne sredine - poslovi dozimetrije i tehnicke zastite od zracenja kod reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M; Bjelanovic, J; Minincic, Z; Komatina, R; Raicevic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Laboratory for radiation and environmental proetecion, Beograd (Serbia and Montenegro)

    1987-12-15

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted was less than 6.0 mSv during past 10 months. Individual exposures for 9/10 of the personnel were less than 1/10 of the annual permissible exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. During 1987 there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel. [Serbo-Croat] U ovom izvestaju prikazani su analizirani reprezentativni rezultati sakupljeni u okviru kontrole radne sredine i tehnicke zastite od zracenja reaktora RA. U prvom delu izvestaja izlozeni su podaci o osnovnim vidovima izlaganja zracenju i statisticki pregled ukupnog broja radiacionih merenja. Dati su takodje rezultati merenja sadrzaja radioktivnih gasova i aerosola u vazduhu, kao i stepena kontaminacije povrsina, odece i otkrivenih delova tela osoblja. U drugom delu izvestaja izlozeni su rezultati analize ozracivanja radnog osoblja. Utvrdjeno je da je maksimalna individualna doza spoljasnjeg izlaganja u proteklih 10 meseci bila 6,0 mSv, a da su pojadinacna izlaganja vise od 9/10 radnog osoblja bila manja od 1/10 godisnje granicne vrednosti. Dati su takodje uporedni podaci o ozracivanju osoblja u prethodnoj, kao i u pet proteklih godina, iz kojih

  20. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

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

  1. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

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

  2. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

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

  3. Reactor handbook. 2. rev. ed.

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  4. FFTF and CRBRP reactor vessels

    International Nuclear Information System (INIS)

    Morgan, R.E.

    1977-01-01

    The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

  5. Protecting nuclear power plants. Chapter 2. On the importance of the security and safety of the reactor pressure vessel to external threats

    International Nuclear Information System (INIS)

    Ballesteros, A.; Gonzalez, J.; Debarberis, L.

    2006-01-01

    Nuclear power plants have blong been recognized as potential targets of terrorist attacks, and critics have long questioned the adequacy of the existing measures to defend against such attacks. The 11-S 2001, 11-M 2004 and 7-J 2005 attacks in USA, Spain and UK illustrated the deadly intention and abilities of modern terrorist groups. These attacks also brought to surface long standing concerns about the vulnerability of nuclear installations to possible terrorist attacks. Commercial nuclear reactors contain large inventory of radioactive fission products which, if dispersed, could pose a direct radiation hazard on the population. The reactor pressure vessel (RPV), which contains the nuclear fuel, is the most critical component of the plant. This paper shows that small amount of explosive material can produce irreversible damage in the RPV and the release of radioactive material. Therefor, access of working personal to the vicinity of the RPV during the refuelling outage should be stricktly limited. It should be considered a high priority security issue

  6. Protective head of sensors

    International Nuclear Information System (INIS)

    Liska, K.; Anton, P.

    1987-01-01

    The discovery concerns the protective heads of diagnostic assemblies of nuclear power plants for conductors of the sensors from the fuel and control parts of the said assemblies. A detailed description is presented of the design of the protective head which, as compared with the previous design, allows quick and simple assembly with reduced risk of damaging the sensors. The protective head may be used for diagnostic assemblies both in power and in research reactors and it will be used for WWER reactor assemblies. (A.K.). 3 figs

  7. Design concepts for the reactor protection and control process instrumentation digital upgrade project at the Donald C. Cook Nuclear Plant units 1 and 2

    International Nuclear Information System (INIS)

    Carruth, R.C.; Sotos, W.G.

    1996-01-01

    As the nation's nuclear power plants age, the need to consider upgrading of their electronic protection and control systems becomes more urgent. Hardware obsolescence and mechanical wear out resulting from frequent calibration and surveillance play a major role in defining their useful life. At Cook Nuclear Plant, a decision was made to replace a major portion of the plant's protection and control systems with newer technology. This paper describes the engineering processes involved in this successful upgrade and explains the basis for many decisions made while performing the digital upgrade

  8. Control of the working environment dosimetry and technical radiation protection at the RA reactor, Part I; Deo I: Kontrola radne sredine - poslovi dozimetrije i tehnicke zastite od zracenja kod reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, N; Bjelanovic, J; Minicic, Z; Komatina, R; Raicevic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1984-12-15

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted to 16.9 mSV during past 10 months. Individual exposures for 9/10 of the personnel were less than 1/10 of the permissible annual exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. The last part analyzes accidents occurred at the reactor during 1984. It was found that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel.

  9. Backfitting of the FRG reactors

    International Nuclear Information System (INIS)

    Krull, W.

    1990-01-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U 3 Si 2 fuel. Both cooling towers were repaired. Replacement of instrumentation is planned

  10. Development of key technologies in DPSSL system for fast-ignition, laser fusion reactor - FIREX, HALNA, and protection of final optics

    International Nuclear Information System (INIS)

    Norimatsu, T.; Azechi, H.; Fujimoto, Y.; Jitsuno, T.; Kanabe, T.; Kodama, R.; Kondo, K.; Miyanaga, N.; Nagatomo, H.; Nakatsuka, M.; Shiraga, H.; Tanaka, K.A.; Tsubakimoto, K.; Yamanaka, M.; Yasuhara, R.; Izawa, Y.; Kawashima, T.; Kurita, T.; Matsumoto, O.; Tsuchiya, Y.; Sekine, T.; Kan, H.

    2005-01-01

    A critical path to a laser fusion power plant is construction of a reliable, efficient, high repetitive energy driver including the relation with the reactor environment. At ILE, Osaka University, FIREX project has been proposed and the phase I to show heating of compressed fuel to 5 keV has started with construction of the FIREX laser. This project will demonstrate physics of fast ignition and elemental studies are carried out to obtain persuasive data to find the path to the goal. A diode-laser-pumped, solid-state-laser (DPSSL) HALNA-10 succeeded in operation of 7.5J output power at 10 Hz rep-rate. Contamination of final optics by metal vapor was studied using a 1/10 model of the beam duct. The result indicated that contamination can be controlled with high speed shutters and a low pressure buffer gas. (author)

  11. RB Research nuclear reactor, Annual report for 2007

    International Nuclear Information System (INIS)

    Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.; Sipka, Dj.

    2008-01-01

    Report on RB reactor operation during 2007 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2007. Majority of measurement were related to spent fuel from the RA reactor, safety of transportation containers and verification of relevant computer codes

  12. 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

  13. Radiation protection and environmental protection

    International Nuclear Information System (INIS)

    Xie Zi; Dong Liucan; Zhang Yongxing

    1994-01-01

    A collection of short papers is presented which review aspects of research in radiation and environmental protection carried out by the Chinese Institute of Atomic Energy in 1991. The topics covered are: the analysis of Po 210 in the gaseous effluent of coal-fired boilers; the determination of natural radionuclide levels in various industrial waste slags and management countermeasures; assessment of the collective radiation dose from natural sources for the Chinese population travelling by water; the preliminary environmental impact report for the multipurpose heavy water research reactor constructed by China for the Islamic Republic of Algeria. (UK)

  14. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

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

  15. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

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

  16. 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

  17. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Raisic, N.

    1962-09-01

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

  19. Heterogeneous reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

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

  20. The nuclear reactor systems

    International Nuclear Information System (INIS)

    Bacher, P.

    2008-01-01

    This paper describes the various nuclear reactor systems, starting with the Generation II, then the present development of the Generation III and the stakes and challenges of the future Generation IV. Some have found appropriate to oppose reactor systems or generations one to another, especially by minimizing the enhancements of generation III compared to generation II or by expecting the earth from generation IV (meaning that generation III is already obsolete). In the first part of the document (chapter 2), some keys are given to the reader to develop its proper opinion. Chapter 3 describes more precisely the various reactor systems and generations. Chapter 4 discusses the large industrial manoeuvres around the generation III, and the last chapter gives some economical references, taking into account, for the various means of power generation, the impediments linked to climate protection

  1. Reactor control device

    International Nuclear Information System (INIS)

    Araki, Takao; Inoue, Toyokazu.

    1981-01-01

    Purpose: To protect the reactor floor by alleviating the shock imparted to the reactor floor by a dropped control rod when a wire rope accidentally breaks. Constitution: A control rod is hung by wire rope from a control rod drive, and shock absorbers are mounted at the upper and lower portions of the control rod. The outer diameter of the upper shock absorber is made larger than the inner diameter of a control rod inserting hole formed in the reactor core. If the control rod drops, the upper absorber is stopped at the upper tapered portion of the inserting hole. Thus, the dropping energy of the control rod can be sufficiently absorbed by the upper and lower shock absorbers. (Kamimura, M.)

  2. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Science.gov (United States)

    2013-12-09

    [email protected] . Both of the Office of New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U...: timothy.kolb@nrc.gov . SUPPLEMENTARY INFORMATION: I. Accessing Information and Submitting Comments A...

  3. University Reactor Matching Grants Program

    International Nuclear Information System (INIS)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-01-01

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

  4. Safety inspections to TRIGA reactors

    International Nuclear Information System (INIS)

    Byszewski, W.

    1988-01-01

    The operational safety advisory programme was created to provide useful assistance and advice from an international perspective to research reactor operators and regulators on how to enhance operational safety and radiation protection on their reactors. Safety missions cover not only the operational safety of reactors themselves, but also the safety of associated experimental loops, isotope laboratories and other experimental facilities. Safety missions are also performed on request in other Member States which are interested in receiving impartial advice and assistance in order to enhance the safety of research reactors. The results of the inspections have shown that in some countries there are problems with radiation protection practices and nuclear safety. Very often the Safety Analysis Report is not updated, regulatory supervision needs clarification and improvement, maintenance procedures should be more formalised and records and reports are not maintained properly. In many cases population density around the facility has increased affecting the validity of the original safety analysis

  5. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-01

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

  6. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  7. 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)

  8. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

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

  9. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

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

  10. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

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

  11. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

  12. Radiation detectors for reactors

    International Nuclear Information System (INIS)

    Balagi, V.

    2005-01-01

    Detection and measurement of radiation plays a vital role in nuclear reactors from the point of view of control and safety, personnel protection and process control applications. Various types of radiation are measured over a wide range of intensity. Consequently a variety of detectors find use in nuclear reactors. Some of these devices have been developed in Electronics Division. They include gas-filled detectors such as 10 B-lined proportional counters and chambers, fission detectors and BF 3 counters are used for the measurement of neutron flux both for reactor control and safety, process control as well as health physics instrumentation. In-core neutron flux instrumentation employs the use detectors such as miniature fission detectors and self-powered detectors. In this development effort, several indigenous materials, technologies and innovations have been employed to suit the specific requirement of nuclear reactor applications. This has particular significance in view of the fact that several new types of reactors such as P-4, PWR and AHWR critical facilities, FBTR, PFBR as well as the refurbishment of old units like CIRUS are being developed. The development work has sought to overcome some difficulties associated with the non-availability of isotopically enriched neutron-sensing materials, achieving all-welded construction etc. The present paper describes some of these innovations and performance results. (author)

  13. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

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

  14. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

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

  15. Reactor container

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  16. 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

  17. Safety equipment in a reactor

    International Nuclear Information System (INIS)

    Shiratori, Hirozo; Ishiyama, Satoshi; Ugawa, Yukio.

    1976-01-01

    Object: To safely retain, even if fuel should be molten and flown through the bottom of a container in a reactor, the molten fuel to remove heat generation of the fuel to prevent occurrence of a critical trouble. Structure: A reactor container housing a core and coolant has thereunder a separation dome in a central portion thereof and a partitioning plate coaxially and circularly disposed in the periphery of the separation dome, with a tray formed of magnesium oxide being disposed. Further, a cooling path system is provided so as to surround the tray. The cooling path system and the reactor container are surrounded and protected by a reactor wall provided with heat insulating refractory bricks, a coolant pouring system extends through the reactor wall, and the coolant is supplied to the tray. (Furukawa, Y.)

  18. Oscillating liquid flow ICF Reactor

    International Nuclear Information System (INIS)

    Petzoldt, R.W.

    1990-01-01

    Oscillating liquid flow in a falling molten salt inertial confinement fusion reactor is predicted to rapidly clear driver beam paths of residual liquid droplets. Oscillating flow will also provide adequate neutron and x-ray protection for the reactor structure with a short (2-m) fall distance permitting an 8 Hz repetition rate. A reactor chamber configuration is presented with specific features to clear the entire heavy-ion beam path of splashed molten salt. The structural components, including the structure between beam ports, are shielded. 3 refs., 12 figs

  19. Operation and maintenance of the RB reactor, Annual report for 1977

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1977-01-01

    The annual report for 1977 includes the following: utilization of the RB reactor; new regulations and instructions for reactor operation; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Five annexes are concerned with: testing the properties of preamplifiers for linear and logarithmic experimental channels; properties of the neutron converter; maintenance of the reactor equipment; purchase of new equipment; and the program for training reactor operators

  20. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

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

  1. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

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

  2. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

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

  3. Operation and maintenance of the RB reactor, Annual report for 1980

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.; Markovic, H.; Zivkovic, B.; Gogdanovic, M.; Petronijevic, M.

    1980-12-01

    This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 9 Annexes as follows: Utilization of the RB reactor from 1976 - 1980; program of reactor utilization from 1981-1985; contents of the RB reactor safety report; maintenance of the reactor components and equipment in 1980; verification of reactor reliability after the earthquake (May 18 1980); refurbishment of equipment in 1980, and purchasing new equipment from 1981-1985; review of radiation doses in the reactor building and exposure doses for the reactor staff; personnel data and financial data

  4. RB research nuclear reactor, Annual report for 1982; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1982-12-15

    This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 7 Annexes as follows: Maintenance of reactor equipment in 1982; contents of the RB reactor safety report; review of radiation doses in the reactor building and exposure doses for the reactor staff; utilization of the RB reactor in 1982; and financial data.

  5. 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

  6. The reactor Cabri

    International Nuclear Information System (INIS)

    Ailloud, J.; Millot, J.P.

    1964-01-01

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

  7. RB Research nuclear reactor, Annual report for 2006

    International Nuclear Information System (INIS)

    Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.

    2007-01-01

    Report on RB reactor operation during 2006 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains detailed data concerned with measurements performed at the RB reactor and a number of significant results obtained

  8. Research into radiation protection. 1994 Programme report. Report on radiation departmental research programme on radiation protection, sponsored by the Federal Ministry for the Environment, Nature Conservation and Reactor Safety, and placed under the administrative and subject competence of the Federal Radiation Protection Office

    International Nuclear Information System (INIS)

    Goedde, R.; Schmitt-Hannig, A.; Thieme, M.

    1994-10-01

    On behalf of the Ministery for Environment, Nature Conservation and Nuclear Safety (BMU), the Federal Office for Radiation Protection is placing research and study contracts in the field of radiation protection. The results of these projects are used for developing radiation protection rules and to fulfill the special radiation protection tasks of the BMU, required by law. Planning, expert and administrative management, placing, assistance as well as expert evaluation of the results from these research projects lies within the responsibility of the Federal Office for Radiation Protection. This report provides information on preliminary and final results of radiation protection projects within the BMU Department Research Programme of the year 1994. (orig.) [de

  9. Research into radiation protection. 1995 Programme report. Report on radiation departmental research programme on radiation protection, sponsored by the Federal Ministry for the Environment, Nature Conservation and Reactor Safety, and placed under the administrative and subject competence of the Federal Radiation Protection Office

    International Nuclear Information System (INIS)

    Thieme, M.; Goedde, R.; Schmitt-Hannig, A.

    1996-01-01

    On behalf of the Ministry for Environment, Nature Conservation and Nuclear Safety (BMU), the Federal Office for Radiation Protection is placing research and study contracts in the field of radiation protection. The results of these projects are used for developing radiation protection rules and to fulfill the special radiation protection tasks of the BMU, required by law. Planning, expert and administrative management, placing, assistance as well as expert evaluation of the results from these research projects lies within the responsibility of the Federal Office for Radiation Protection. This report provides information on preliminary and final results of radiation protection projects within the BMU Department Research Programme of the year 1995. (orig.) [de

  10. 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

  11. 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.)

  12. Reactor container

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  13. REACTOR SHIELD

    Science.gov (United States)

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

    1959-02-17

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

  14. 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.

  15. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  17. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  18. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

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

  19. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

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

  20. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  1. Design strategy for control of inherently safe reactors

    International Nuclear Information System (INIS)

    Chisholm, G.H.

    1984-01-01

    Reactor power plant safety is assured through a combination of engineered barriers to radiation release (e.g., reactor containment) in combination with active reactor safety systems to shut the reactor down and remove decay heat. While not specifically identified as safety systems, the control systems responsible for continuous operation of plant subsystems are the first line of defense for mitigating radiation releases and for plant protection. Inherently safe reactors take advantage of passive system features for decay-heat removal and reactor shutdown functions normally ascribed to active reactor safety systems. The advent of these reactors may permit restructuring of the present control system design strategy. This restructuring is based on the fact that authority for protection against unlikely accidents is, as much as practical, placed upon the passive features of the system instead of the traditional placement upon the PPS. Consequently, reactor control may be simplified, allowing the reliability of control systems to be improved and more easily defended

  2. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    Mattern, J.

    1976-01-01

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

  3. Nuclear reactor fuel sub-assemblies

    International Nuclear Information System (INIS)

    Dodd, J.A.

    1981-01-01

    An improved fuel sub-assembly for a liquid metal cooled fast breeder reactor, is described, in which fatigue damage due to buffeting by cross-current flows is reduced and protection is provided against damage by contact with other reactor structures during loading and unloading of the sub-assembly. (U.K.)

  4. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  5. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

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

  6. Radiation protection/shield design

    International Nuclear Information System (INIS)

    Disney, R.K.

    1977-01-01

    Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.) [de

  7. Reactor safeguards against insider sabotage

    International Nuclear Information System (INIS)

    Bennett, H.A.

    1982-03-01

    A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested

  8. Backfitting of the FRG reactors

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)

    1990-05-01

    The FRG-research reactors The GKSS-research centre is operating two research reactors of the pool type fueled with MTR-type type fuel elements. The research reactors FRG-1 and FRG-2 having power levels of 5 MW and 15 MW are in operation for 31 year and 27 years respectively. They are comparably old like other research reactors. The reactors are operating at present at approximately 180 days (FRG-1) and between 210 and 250 days (FRG-2) per year. Both reactors are located in the same reactor hall in a connecting pool system. Backfitting measures are needed for our and other research reactors to ensure a high level of safety and availability. The main backfitting activities during last ten years were concerned with: comparison of the existing design with today demands (criteria, guidelines, standards etc.); and probability approach for events from outside like aeroplane crashes and earthquakes; the main accidents were rediscussed like startup from low and full power, loss of coolant flow, loss of heat sink, loss of coolant and fuel plate melting; a new reactor protection system had to be installed, following today's demands; a new crane has been installed in the reactor hall. A cold neutron source has been installed to increase the flux of cold neutrons by a factor of 14. The FRG-l is being converted from 93% enriched U with Alx fuel to 20% enriched U with U{sub 3}Si{sub 2} fuel. Both cooling towers were repaired. Replacement of instrumentation is planned.

  9. Occupational health physics at a fusion reactor

    International Nuclear Information System (INIS)

    Shank, K.E.; Easterly, C.E.; Shoup, R.L.

    1975-01-01

    Future generation of electrical power using controlled thermonuclear reactors will involve both traditional and new concerns for health protection. A review of the problems associated with exposures to tritium and magnetic fields is presented with emphasis on the occupational worker. The radiological aspects of tritium, inventories and loss rates of tritium for fusion reactors, and protection of the occupational worker are discussed. Magnetic fields in which workers may be exposed routinely and possible biological effects are also discussed

  10. Radiation protection at the RA Reactor in 1986, Part -2, Annex 1, Radioactivity control of working environment, dosimetry; Deo 2 - Prilog 1 - Kontrola radne sredine - poslovi dozimetrije i tehnicke zastite od zracenja kod reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Ninkovic, M; Bjelanovic, J; Minincic, Z; Komatina, R; Raicevic, J [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1986-12-01

    This report contains data and analysis of the of measured sample results collected during radiation protection control in the working environment of the RA reactor. First part contains basic exposure values and statistical review of the the total number of radiation measurements. It includes contents of radioactive gasses and effluents in the air, as well as the level of surface contamination of clothes and uncovered parts of the personnel bodies. Second part deals with the analysis of personnel doses. It was found that the maximum individual dose from external irradiation amounted to 20.5 mSV during past 10 months. Individual exposures for 7/10 of the personnel were less than 1/10 of the annual permissible exposure. Data are compared to radiation doses for last year and previous five years. Third part of this annex contains basic data about the quantity of collected radioactive waste, total quantity of contaminated and decontaminated surfaces. The last part analyzes accidents occurred at the reactor during 1986. It was found that there have been no accidents that could cause significant contamination of working surfaces and components nor radiation exposure of the personnel. [Serbo-Croat] U ovom izvestaju prikazani su i analizirani reprezentativni rezultati sakupljeni u okviru kontrole radne sredine i tehnicke zastite od zracenja reaktora RA. U prvom delu izvestaja izlozeni su podaci o osnovnim vidovima izlaganja zracenju i statisticki pregled ukupnog broja radiacionih merenja. Dati su takodje rezultati merenja sadrzaja radioktivnih gasova i aerosola u vazduhu, kao i stepena kontaminacije povrsina, odece i otkrivenih delova tela osoblja. U drugom delu izvestaja izlozeni su rezultati analize ozracivanja radnog osoblja. Utvrdjeno je da je maksimalna individualna doza spoljasnjeg izlaganja u proteklih 10 meseci bila 20,5 mSv, a da su pojadinacna izlaganja vise od 7/10 radnog osoblja bila manja od 1/10 godisnje granicne vrednosti. Dati su takodje uporedni podaci o

  11. Decommissioning a nuclear reactor

    International Nuclear Information System (INIS)

    Montoya, G.M.

    1991-01-01

    The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs

  12. The fuel of nuclear reactors

    International Nuclear Information System (INIS)

    1995-03-01

    This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

  13. Decommissioning of Salaspils nuclear reactor

    International Nuclear Information System (INIS)

    Abramenkovs, A.; Malnachs, J.; Popelis, A.

    2002-01-01

    In May 1995, the Latvian Government decided to shut down the Research Reactor Salaspils (SRR) and to dispense with nuclear energy in future. The reactor has been out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH from 1998-1999. he Latvian Government decided on 26 October 1999 to start the direct dismantling to 'green field' in 2001. The results of decommissioning and dismantling performed in 1999-2001 are presented and discussed. The main efforts were devoted to collecting and conditioning 'historical' radioactive waste from different storages outside and inside the reactor hall. All radioactive material more than 20 tons were conditioned in concrete containers for disposal in the radioactive waste depository 'Radons' in the Baldone site. Personal protective and radiation measurement equipment was upgraded significantly. All non-radioactive equipment and material outside the reactor buildings were free-released and dismantled for reuse or conventional disposal. Weakly contaminated material from the reactor hall was collected and removed for free-release measurements. The technology of dismantling of the reactor's systems, i.e. second cooling circuit, zero power reactors and equipment, is discussed in the paper. (author)

  14. 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.

  15. 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)

  16. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  17. Structure of thermonuclear reactor wall

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro.

    1991-01-01

    In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

  18. Lessons from feedback of safety operating experience for reactor physics

    International Nuclear Information System (INIS)

    Suchomel, J.; Rapavy, S.

    1999-01-01

    Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

  19. 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

  20. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.