WorldWideScience

Sample records for reactor system analysis

  1. Safety analysis of reactor's cooling system

    International Nuclear Information System (INIS)

    1999-01-01

    Results of the analysis of reactor's RBMK-1500 coolant system during normal operation mode, hydrodynamic testing and in the case of earthquake are presented. Analysis was performed using RELAP5 code. Calculations showed the most vulnerable place in the reactor's coolant system. It was found that in the case of earthquake the horizontal support system of drum separator could be damaged

  2. 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.

  3. Systems analysis of the CANDU 3 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

    1993-07-01

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

  4. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  5. Computation system for nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals

  6. Parametric systems analysis for ICF hybrid reactors

    International Nuclear Information System (INIS)

    Berwald, D.H.; Maniscalco, J.A.; Chapin, D.L.

    1981-01-01

    Parametric design and systems analysis for inertial confinement fusion-fission hybrids are presented. These results were generated as part of the Electric Power Research Institute (EPRI) sponsored Feasibility Assessment of Fusion-Fission Hybrids, using an Inertial Confinement Fusion (ICF) hybrid power plant design code developed in conjunction with the feasibility assessment. The SYMECON systems analysis code, developed by Westinghouse, was used to generate economic results for symbiotic electricity generation systems consisting of the hybrid and its client Light Water Reactors (LWRs). These results explore the entire fusion parameter space for uranium fast fission blanket hybrids, thorium fast fission blanket hybrids, and thorium suppressed fission blanket types are discussed, and system sensitivities to design uncertainties are explored

  7. Integrated systems analysis of the PIUS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  8. Integrated systems analysis of the PIUS reactor

    International Nuclear Information System (INIS)

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects ampersand Criticality Analysis (FMECA) and Hazards ampersand Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions

  9. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.

    1983-04-01

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  10. Systemization of Design and Analysis Technology for Advanced Reactor

    International Nuclear Information System (INIS)

    Kim, Keung Koo; Lee, J.; Zee, S. K.

    2009-01-01

    The present study is performed to establish the base for the license application of the original technology by systemization and enhancement of the technology that is indispensable for the design and analysis of the advanced reactors including integral reactors. Technical reports and topical reports are prepared for this purpose on some important design/analysis methodology; design and analysis computer programs, structural integrity evaluation of main components and structures, digital I and C systems and man-machine interface design. PPS design concept is complemented reflecting typical safety analysis results. And test plans and requirements are developed for the verification of the advanced reactor technology. Moreover, studies are performed to draw up plans to apply to current or advanced power reactors the original technologies or base technologies such as patents, computer programs, test results, design concepts of the systems and components of the advanced reactors. Finally, pending issues are studied of the advanced reactors to improve the economics and technology realization

  11. Computer System Analysis for Decommissioning Management of Nuclear Reactor

    International Nuclear Information System (INIS)

    Nurokhim; Sumarbagiono

    2008-01-01

    Nuclear reactor decommissioning is a complex activity that should be planed and implemented carefully. A system based on computer need to be developed to support nuclear reactor decommissioning. Some computer systems have been studied for management of nuclear power reactor. Software system COSMARD and DEXUS that have been developed in Japan and IDMT in Italy used as models for analysis and discussion. Its can be concluded that a computer system for nuclear reactor decommissioning management is quite complex that involved some computer code for radioactive inventory database calculation, calculation module on the stages of decommissioning phase, and spatial data system development for virtual reality. (author)

  12. Reactor noise analysis applications in NPP I and C systems

    Energy Technology Data Exchange (ETDEWEB)

    Gloeckler, O. [International Atomic Energy Agency, Wagramer Strosse 5, A-1400 Vienna, Austria Ontario Power Generation, 230 Westney Road South, Ajax, Ont. L1S 7R3 (Canada)

    2006-07-01

    Reactor noise analysis techniques are used in many NPPs on a routine basis as 'inspection tools' to get information on the dynamics of reactor processes and their instrumentation in a passive, non-intrusive way. The paper discusses some of the tasks and requirements an NPP has to take to implement and to use the full advantages of reactor noise analysis techniques. Typical signal noise analysis applications developed for the monitoring of the reactor shutdown system and control system instrumentation of the Candu units of Ontario Power Generation and Bruce Power are also presented. (authors)

  13. 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)

  14. Prism reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-08-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  15. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

  16. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  17. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  18. A systems analysis of the ARIES tokamak reactors

    International Nuclear Information System (INIS)

    Bathke, C.G.

    1992-01-01

    The multi-institutional ARIES study has completed a series of cost-of-electricity optimized conceptual designs of commercial tokamak fusion reactors that vary the assumed advances in technology and physics. A comparison of these designs indicates the cost benefit of various design options. A parametric systems analysis suggests a possible means to obtain a marginally competitive fusion reactor

  19. JAERI thermal reactor standard code system for reactor design and analysis SRAC

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro

    1985-01-01

    SRAC, JAERI thermal reactor standard code system for reactor design and analysis, developed in Japan Atomic Energy Research Institute, is for all types of thermal neutron nuclear design and analysis. The code system has undergone extensive verifications to confirm its functions, and has been used in core modification of the research reactor, detailed design of the multi-purpose high temperature gas reactor and analysis of the experiment with a critical assembly. In nuclear calculation with the code system, multi-group lattice calculation is first made with the libraries. Then, with the resultant homogeneous equivalent group constants, reactor core calculation is made. Described are the following: purpose and development of the code system, functions of the SRAC system, bench mark tests and usage state and future development. (Mori, K.)

  20. Review of the treat upgrade reactor scram system reliability analysis

    International Nuclear Information System (INIS)

    Montague, D.F.; Fussell, J.B.; Krois, P.A.; Morelock, T.C.; Knee, H.E.; Manning, J.J.; Haas, P.M.; West, K.W.

    1984-10-01

    In order to resolve some key LMFBR safety issues, ANL personnel are modifying the TREAT reactor to handle much larger experiments. As a result of these modifications, the upgraded Treat reactor will not always operate in a self-limited mode. During certain experiments in the upgraded TREAT reactor, it is possible that the fuel could be damaged by overheating if, once the computer systems fail, the reactor scram system (RSS) fails on demand. To help ensure that the upgraded TREAT reactor is shut down when required, ANL personnel have designed a triply redundant RSS for the facility. The RSS is designed to meet three reliability goals: (1) a loss of capability failure probability of 10 -9 /demand (independent failures only); (2) an inadvertent shutdown probability of 10 -3 /experiment; and (3) protection agaist any known potential common cause failures. According to ANL's reliability analysis of the RSS, this system substantially meets these goals

  1. Establishment of computer code system for nuclear reactor design - analysis

    International Nuclear Information System (INIS)

    Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.

    1996-01-01

    Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab

  2. Development of the next generation reactor analysis code system, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki

    2011-03-01

    A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)

  3. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  4. Fault tree analysis for reactor systems

    International Nuclear Information System (INIS)

    Crosetti, P.A.

    1971-01-01

    Reliability analysis is playing an increasingly important role in quantitative assessment of system performance for assuring nuclear safety, improving plant performance and plant life, and reducing plant operating costs. The complexity of today's nuclear plants warrant the use of techniques which will provide a comprehensive evaluation of systems in their total context. In particular, fault tree analysis with probability evaluation can play a key role in assuring nuclear safety, in improving plant performance and plant life, and in reducing plant operating costs. The technique provides an all inclusive, versatile mathematical tool for analyzing complex systems. Its application can include a complete plant as well as any of the systems and subsystems. Fault tree analysis provides an objective basis for analyzing system design, performing trade-off studies, analyzing common mode failures, demonstrating compliance with AEC requirements, and justifying system changes or additions. The logic of the approach makes it readily understandable and, therefore, it serves as an effective visibility tool for both engineering and management. (U.S.)

  5. Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

    International Nuclear Information System (INIS)

    Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

    1992-01-01

    A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

  6. Data acquisition and processing system for reactor noise analysis

    International Nuclear Information System (INIS)

    Costa Oliveira, J.; Morais Da Veiga, C.; Forjaz Trigueiros, D.; Pombo Duarte, J.

    1975-01-01

    A data acquisition and processing system for reactor noise analysis by time correlation methods is described, consisting in one to four data feeding channels (transducer, associated electronics and V/f converter), a sampling unit, a landline transmission system and a PDP 15 computer. This system is being applied to study the kinetic parameters of the 'Reactor Portugues de Investigacao', a swimming-pool 1MW reactor. The main features that make such a data acquisition and processing system a useful tool to perform noise analysis are: the improved characteristics of analog-to-digital converters employed to quantize the signals; the use of an on-line computer which allows a great accumulation and a rapid treatment of data together with an easy check of the correctness of the experiments; and the adoption of the time cross-correlation technique using two-detectors which by-pass the limitation of low efficiency detectors. (author)

  7. Multivariate statistical pattern recognition system for reactor noise analysis

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

    1976-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system

  8. Multivariate statistical pattern recognition system for reactor noise analysis

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

    1975-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system. 19 references

  9. 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.)

  10. Accelerator-driven transmutation reactor analysis code system (ATRAS)

    Energy Technology Data Exchange (ETDEWEB)

    Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)

  11. Sophistication of burnup analysis system for fast reactor

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Hyoudou, Hideaki; Tatsumi, Masahiro

    2010-02-01

    Improvement on prediction accuracy for neutronics property of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified constants library as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores, however, improvement of not only static properties but also burnup properties is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup properties using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous research, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for production systems. In the present study, we implemented functions for cell calculations and burnup calculations. With this, whole steps in analysis can be carried out with only this system. In addition, we modified the specification of user input to improve the convenience of this system. Since implementations being done so

  12. Analysis of reactor trips originating in balance of plant systems

    International Nuclear Information System (INIS)

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W.

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs

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

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

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

  14. Comparative analysis of nuclear reactor control system designs

    International Nuclear Information System (INIS)

    Russcher, G.E.

    1975-01-01

    Control systems are vital to the safe operation of nuclear reactors. Their seismic design requirements are some of the most important criteria governing reactor system design evaluation. Consequently, the seismic analysis for nuclear reactors is directed to include not only the mechanical and structural seismic capabilities of a reactor, but the control system functional requirements as well. In the study described an alternate conceptual design of a safety rod system was compared with a prototypic system design to assess their relative functional reliabilities under design seismic conditions. The comparative methods utilized standard success tree and decision tree techniques to determine the relative figures of merit. The study showed: (1) The methodology utilized can provide both qualitative and quantitative bases for design decisions regarding seismic functional capabilities of two systems under comparison, (2) the process emphasizes the visibility of particular design features that are subject to common mode failure while under seismic loading, and (3) minimal improvement was shown to be available in overall system seismic performance of an independent conceptual design, however, it also showed the system would be subject to a new set of operational uncertainties which would have to be resolved by extensive development programs

  15. Sophistication of burnup analysis system for fast reactor (2)

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hirai, Yasushi; Tatsumi, Masahiro

    2010-10-01

    Improvement on prediction accuracy for neutronics characteristics of fast reactor cores is one of the most important study domains in terms of both achievement of high economical plant efficiency based on reasonably advanced designs and increased reliability and safety margins. In former study, considerable improvement on prediction accuracy in neutronics design has been achieved in the development of the unified cross-section set as a fruit of a series of critical experiments such as JUPITER in application of the reactor constant adjustments. For design of fast reactor cores improvement of not only static characteristics but also burnup characteristics is very important. For such purpose, it is necessary to improve the prediction accuracy on burnup characteristics using actual burnup data of 'JOYO' and 'MONJU', experimental and prototype fast reactors. Recently, study on effective burnup method for minor actinides becomes important theme. However, there is a problem that analysis work tends to become inefficient for lack of functionality suitable for analysis of composition change due to burnup since the conventional analysis system is targeted to critical assembly systems. Therefore development of burnup analysis system for fast reactors with modularity and flexibility is being done that would contribute to actual core design work and improvement of prediction accuracy. In the previous study, we have developed a prototype system which has functions of performing core and burnup calculations using given constant files (PDS files) and information based on simple and easy user input data. It has also functions of fuel shuffling which is indispensable for power reactor analysis systems. In the present study, by extending the prototype system, features for handling of control rods and energy collapse of group constants have been designed and implemented. Computational results from the present analysis system are stored into restart files which can be accessible by

  16. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  17. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    International Nuclear Information System (INIS)

    Shropshire, D.E.

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program's understanding of the cost drivers that will determine nuclear power's cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-irradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  18. 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.

  19. SRAC: JAERI thermal reactor standard code system for reactor design and analysis

    International Nuclear Information System (INIS)

    Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.

    1983-01-01

    The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)

  20. Event tree analysis for the system of hybrid reactor

    International Nuclear Information System (INIS)

    Yang Yongwei; Qiu Lijian

    1993-01-01

    The application of probabilistic risk assessment for fusion-fission hybrid reactor is introduced. A hybrid reactor system has been analysed using event trees. According to the character of the conceptual design of Hefei Fusion-fission Experimental Hybrid Breeding Reactor, the probabilities of the event tree series induced by 4 typical initiating events were calculated. The results showed that the conceptual design is safe and reasonable. through this paper, the safety character of hybrid reactor system has been understood more deeply. Some suggestions valuable to safety design for hybrid reactor have been proposed

  1. Singular perturbation analysis of relaxation oscillations in reactor systems

    International Nuclear Information System (INIS)

    Ward, M.E.; Lee, J.C.

    1987-01-01

    A singular perturbation method for the analysis of large power oscillations in nuclear reactors is applied to obtain phase-plane solutions of the Ergen-Weinberg model. The system equations, recast in an appropriate form, directly give a first approximation to the closed trajectory in which the system behaviour is idealized as relaxation oscillations. Further approximations in the phase plane are determined using separate perturbation series on individual parts of the oscillation, with variations in the assignment of dependent and independent variables to consistently obtain convergent series. The accuracy of each order of the phase-plane solution increases with the magnitude of the power pulse in the actual physical situation. For realistic reactor conditions, both the trajectory and period of oscillation are well predicted using the first two terms of each perturbation series

  2. Analysis of dynamic stability and safety of the reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    This document defines the approximations done for establishing a mathematical model of a reactor. Since the model should be used for safety analysis, it was important to choose a mathematical model less stable than the reactor itself. The analysis was performed on the analog computer RAS. Results obtained and conclusions concerned with three possible reactor accidents are presented [sr

  3. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  4. Failure mode and effects analysis on typical reactor trip system

    International Nuclear Information System (INIS)

    Eisawy, E.A.

    2010-01-01

    An updated failure mode and effects analysis, FMEA , has been performed on a typical reactor trip system. This upgrade helps to avoid system damage and ,as a result, extends the system service life. It also provides for simplified maintenance and surveillance testing. The operating conditions under which the system is to carry out its function and the operational profile expected for the system have been determined. The results of the FMEA have been given in terms of operating states of the subsystem.The results are given in form of table which is set up such that for a given failure one can read across it and determine which items remain operating in the system. From this data one can identify the number of components operating in the system for monitors pressure exceeds the setpoint pressure.

  5. Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

    International Nuclear Information System (INIS)

    Morrone, A.; Brussalis, W.G.

    1975-01-01

    The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de

  6. 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

  7. Advanced High Temperature Reactor Systems and Economic Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Peretz, Fred J [ORNL; Qualls, A L [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience

  8. Reactor System Design

    International Nuclear Information System (INIS)

    Chi, S. K.; Kim, G. K.; Yeo, J. W.

    2006-08-01

    SMART NPP(Nuclear Power Plant) has been developed for duel purpose, electricity generation and energy supply for seawater desalination. The objective of this project IS to design the reactor system of SMART pilot plant(SMART-P) which will be built and operated for the integrated technology verification of SMART. SMART-P is an integral reactor in which primary components of reactor coolant system are enclosed in single pressure vessel without connecting pipes. The major components installed within a vessel includes a core, twelve steam generator cassettes, a low-temperature self pressurizer, twelve control rod drives, and two main coolant pumps. SMART-P reactor system design was categorized to the reactor coe design, fluid system design, reactor mechanical design, major component design and MMIS design. Reactor safety -analysis and performance analysis were performed for developed SMART=P reactor system. Also, the preparation of safety analysis report, and the technical support for licensing acquisition are performed

  9. 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

  10. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

    2007-06-01

    The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

  11. 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

  12. 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)

  13. Parametric systems analysis of the Modular Stellarator Reactor (MSR)

    International Nuclear Information System (INIS)

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

    1982-05-01

    The close coupling in the stellarator/torsatron/heliotron (S/T/H) between coil design (peak field, current density, forces), magnetics topology (transform, shear, well depth), and plasma performance (equilibrium, stability, transport, beta) complicates the reactor assessment more so than for most magnetic confinement systems. In order to provide an additional degree of resolution of this problem for the Modular Stellarator Reactor (MSR), a parametric systems model has been developed and applied. This model reduces key issues associted ith plasma performance, first-wall/blanket/shield (FW/B/S), and coil design to a simple relationship between beta, system geometry, and a number of indicators of overall plant performance. The results of this analysis can then be used to guide more detailed, multidimensional plasma, magnetics, and coil design efforts towards technically and economically viable operating regimes. In general, it is shown that beta values > 0.08 may be needed if the MSR approach is to be substantially competitive with other approaches to magnetic fusion in terms of system power density, mass utilization, and cost for total power output around 4.0 GWt; lower powers will require even higher betas

  14. Current development in data acquision and processing system for reactor noise analysis in PUSPATI

    International Nuclear Information System (INIS)

    Mohamad Amin Sharifuldin Salleh.

    1986-11-01

    A data acquisition and processing system for reactor noise analysis is described. It consists of four-channel isolation amplifier, a seven-channel DC amplifier, a four-channel analog to digital converter, analog filters, a microcomputer system and a plotter. This system is being applied to investigate the reactor dynamics of the PUSPATI TRIGA MK II reactor. (author)

  15. Development of the versatile reactor analysis code system, MARBLE2

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi

    2015-07-01

    The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)

  16. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

    Shwageraus, E.; Fridman, E.

    2008-01-01

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

  19. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  20. 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.

  1. N-reactor charge-discharge system analysis

    International Nuclear Information System (INIS)

    Tokarz, R.D.; Marr, G.D.; Nesbitt, J.F.

    1982-09-01

    This report documents an analysis of the existing systems in the N-Reactor fuel flow path. It recommends equipment improvements and changes in that path to allow the charge-discharge rates to be increased to 500 tubes per outage without increasing reactor outage time. The estimated program cost of $14 million is projected over an estimated 3-year period. It does not include costs detailed as part of the existing restoration program or any costs that are considered as normal maintenance. The recommendations contained in this report provide a direction and goal for every critical aspect of the fuel flow path. The way in which these recommendations are implemented may greatly affect the schedule and costs. Previous studies by UNC have shown that enhanced fuel element handling has the potential of increasing productivity by 33 days at a cost benefit estimated at $18 million per year. Enhanced fuel handling provides the greatest potential for productivity improvement of any of the areas considered in these studies

  2. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  3. System model for analysis of the mirror fusion-fission reactor

    International Nuclear Information System (INIS)

    Bender, D.J.; Carlson, G.A.

    1977-01-01

    This report describes a system model for the mirror fusion-fission reactor. In this model we include a reactor description as well as analyses of capital cost and blanket fuel management. In addition, we provide an economic analysis evaluating the cost of producing the two hybrid products, fissile fuel and electricity. We also furnish the results of a limited parametric analysis of the modeled reactor, illustrating the technological and economic implications of varying some important reactor design parameters

  4. Nuclear reactor descriptions for space power systems analysis

    Science.gov (United States)

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

    1972-01-01

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

  5. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    International Nuclear Information System (INIS)

    Heising, Carolyn D.

    1998-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach to plant maintenance and control, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R-charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specifications limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (author)

  6. Statistical analysis of the Ft. Calhoun reactor coolant pump system

    International Nuclear Information System (INIS)

    Patel, Bimal; Heising, C.D.

    1997-01-01

    In engineering science, statistical quality control techniques have traditionally been applied to control manufacturing processes. An application to commercial nuclear power plant maintenance and control is presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) of the Ft. Calhoun nuclear power plant. This research uses capability analysis, Shewhart X-bar, R charts, canonical correlation methods, and design of experiments to analyze the process for the state of statistical control. The results obtained show that six out of ten parameters are under control specification limits and four parameters are not in the state of statistical control. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with ample time to respond to possible emergency situations and thus improve plant safety and reliability. (Author)

  7. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    2000-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  8. HORECA. Hoger onderwijs reactor elementary core analysis system. User's manual

    International Nuclear Information System (INIS)

    Battum, E. van; Serov, I.V.

    1993-07-01

    HORECA is developed at IRI Delft for quick analysis of power distribution, burnup and safety for the HOR. It can be used for the manual search of a better loading of the reactor. HORECA is based on the Penn State Fuel Management Package and uses the MCRAC code included in this package as a calculation engine. (orig./HP)

  9. 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

  10. THYDE-NEU: Nuclear reactor system analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro

    2002-03-01

    THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)

  11. Sensitivity Analysis of Reactor Regulating System for SMART

    International Nuclear Information System (INIS)

    Jeon, Yu Lim; Kang, Han Ok; Lee, Seong Wook; Park, Cheon Tae

    2009-01-01

    The integral reactor technology is one of the Small and Medium sized Reactor (SMR) which has recently come into a spotlight due to its suitability for various fields. SMART (System integrated Modular Advanced ReacTor), a small sized integral type PWR with a rated thermal power of 330MWt is one of the advanced SMR. SMART developed by the Korea Atomic Energy Research Institute (KAERI), has a capacity to provide 40,000 m3 per day of potable water and 90 MW of electricity (Chang et al., 2000). Figure 1 shows the SMART which adopts a sensible mixture of new innovative design features and proven technologies aimed at achieving highly enhanced safety and improved economics. Design features contributing to a safety enhancement are basically inherent safety improving features and passive safety features. Fundamental thermal-hydraulic experiments were carried out during the design concepts development to assure the fundamental behavior of major concepts of the SMART systems. A TASS/SMR is a suitable code for accident and performance analyses of SMART. In this paper, we proposed a new power control logic for stable operating outputs of Reactor Regulating System (RRS) of SMART. We analyzed the sensitivity of operating parameter for various operating conditions

  12. CAREM-25 Reactor Second Shutdown System Consolidation Analysis

    International Nuclear Information System (INIS)

    Gimenez, Marcelo; Zanocco, Pablo; Schlamp, Miguel

    2000-01-01

    CAREM Reactor Second Shutdown System (SSS) injects boron into the primary circuit in case of First Shutdown System failure in order to stop the nuclear reaction and to maintain the core in a safe condition during cold shutdown.It also has another safety function which is to inject water in the primary system at any pressure in case of LOCA.Different system requirements are analyzed during a SSS spurious trip and LOCA's transients.Two different alternatives are presented for the stand by condition pressurized system, they are solid mode and hot water layer. Both cases fulfill the design requirements from the safety point of view

  13. Computational analysis of battery optimized reactor integral system

    International Nuclear Information System (INIS)

    Hwang, J. S.; Son, H. M.; Jeong, W. S.; Kim, T. W.; Suh, K. Y.

    2007-01-01

    Battery Optimized Reactor Integral System (BORIS) is being developed as a multi-purpose fast spectrum reactor cooled by lead (Pb). BORIS is an integral optimized reactor with an ultra-long life core. BORIS aims to satisfy various energy demands maintaining inherent safety with the primary coolant Pb, and improving economics. BORIS is being designed to generate 23 MW t h with 10 MW e for at least twenty consecutive years without refueling and to meet the Generation IV Nuclear Energy System goals of sustainability, safety, reliability, and economics. BORIS is conceptualized to be used as the main power and heat source for remote areas and barren lands, and also considered to be deployed for desalinisation purpose. BORIS, based on modular components to be viable for rapid construction and easy maintenance, adopts an integrated heat exchanger system operated by natural circulation of Pb without pumps to realize a small sized reactor. The BORIS primary system is designed through an optimization study. Thermal hydraulic characteristics during a reactor steady state with heat source and sink by core and heat exchanger, respectively, have been carried out by utilizing a computational fluid dynamics code and hand calculations based on first principles. This paper analyzes a transient condition of the BORIS primary system. The Pb coolant was selected for its lower chemical activity with air or water than sodium (Na) and good thermal characteristics. The reactor transient conditions such as core blockage, heat exchanger failure, and loss of heat sink, were selected for this study. Blockage in the core or its inlet structure causes localized flow starvation in one or several fuel assemblies. The coolant loop blockages cause a more or less uniform flow reduction across the core, which may trigger coolant temperature transient. General conservation equations were applied to model the primary system transients. Numerical approaches were adopted to discretized the governing

  14. Performance analysis of Brayton cycle system for space power reactor

    International Nuclear Information System (INIS)

    Li Zhi; Yang Xiaoyong; Zhao Gang; Wang Jie; Zhang Zuoyi

    2017-01-01

    The closed Brayton cycle system now is the potential choice as the power conversion system for High Temperature Gas-cooled Reactors because of its high energy conversion efficiency and compact configuration. The helium is the best working fluid for the system for its chemical stability and small neutron absorption cross section. However, the Helium has small mole mass and big specific volume, which would lead to larger pipes and heat exchanger. What's more, the big compressor enthalpy rise of helium would also lead to an unacceptably large number of compressor's stage. For space use, it's more important to satisfy the limit of the system's volume and mass, instead of the requirement of the system's thermal capacity. So Noble-Gas binary mixture of helium and xenon is presented as the working fluid for space Brayton cycle. This paper makes a mathematical model for space Brayton cycle system by Fortran language, then analyzes the binary mixture of helium and xenon's properties and effects on power conversion units of the space power reactor, which would be helpful to understand and design the space power reactor. The results show that xenon would lead to a worse system's thermodynamic property, the cycle's efficiency and specific power decrease as xenon's mole fraction increasing. On the other hand, proper amount of xenon would decrease the enthalpy changes in turbomachines, which would be good for turbomachines' design. Another optimization method – the specific power optimization is also proposed to make a comparison. (author)

  15. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    International Nuclear Information System (INIS)

    Kobori, Hikaru; Kasada, Ryuta; Hiwatari, Ryoji; Konishi, Satoshi

    2016-01-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO_2 emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO_2 emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO_2 emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  16. Improvement of system code importing evaluation of Life Cycle Analysis of tokamak fusion power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kobori, Hikaru [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Kasada, Ryuta, E-mail: r-kasada@iae.kyoto-u.ac.jp [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan); Hiwatari, Ryoji [Central Research Institute of Electric Power Industry, Tokyo (Japan); Konishi, Satoshi [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2016-11-01

    Highlights: • We incorporated the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code. • We calculated CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. • We found that the objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. • The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. • The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant. - Abstract: This study incorporate the Life Cycle Analysis (LCA) of tokamak type DEMO reactor and following commercial reactors as an extension of a system code to calculate CO{sub 2} emissions from reactor construction, operation and decommissioning that is considered as a major environmental cost. Competitiveness of tokamak fusion power reactors is expected to be evaluated by the cost and environmental impact represented by the CO{sub 2} emissions, compared with present and future power generating systems such as fossil, nuclear and renewables. Result indicated that (1) The objective of conceptual design of the tokamak fusion power reactor is moved by changing evaluation index. (2) The tokamak fusion reactor can reduce CO{sub 2} emissions in the life cycle effectively by reduction of the amount involved in the replacement of internal components. (3) The tokamak fusion reactor achieves under 0.174$/kWh electricity cost, the tokamak fusion reactor is contestable with 1500 degrees-class LNG-fired combined cycle power plant.

  17. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    International Nuclear Information System (INIS)

    Moiseyev, A.V.

    2008-01-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k eff , control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  18. System of Modelling and Calculation Analysis of Neutron- Physical Experiments at Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Moiseyev, A.V. [SSC RF - IPPE, 1 Bondarenko Square, Obninsk, Kaluga Region 249033 (Russian Federation)

    2008-07-01

    There is an actual task on storage, processing and analysis of the unique experimental data received on power fast reactors for their subsequent use in projects of fast reactors of new (4.) generation. For modeling and carrying out analysis of experiments the integrated computing system MODEXSYS has been developed. In this system the mechanism for consecutive calculation of a fast reactor states with the detailed description of its components is created. The system includes the database describing fast reactor states, results of neutron-physical characteristics measurements at fast reactor, calculation and benchmark models of experiments and calculation results. In system convenient search means and the special graphics shell are provided. It has Interfaces for processing of calculation results and their analysis. MODEXSYS system has been applied for analysis of three types of experiments at fast reactor: k{sub eff}, control rod worth and energy release distribution. The most important results of this analysis are described. Application of MODEXSYS system will raise accuracy and reliability of forecasting of fast reactors neutron-physical characteristics; for BN-600 reactor recommended level of accuracy is resulted. (authors)

  19. Numerical analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

    International Nuclear Information System (INIS)

    Schweizer, Fernando Lage Araújo

    2014-01-01

    The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)

  20. Development of PC-based FFT system for reactor dynamic analysis

    International Nuclear Information System (INIS)

    Ansari, S.; Baig, A.R.

    1993-03-01

    A personal computer based fast fourier transform (FFT) analyzer has been developed for frequency spectrum analysis of signals from nuclear reactor. The system can perform window smoothing, computation of auto- and cross-power spectral density, coherence and auto and cross-correlation functions. The feature of 16 analogue signals acquisition with high precision and high sampling frequency makes the analyzer suitable for malfunction diagnosis of nuclear reactors using reactor noise analysis. The development work for the fourier analyzer was undertaken as a part of IAEA research contract no. 5925/RB. The applications of the FFT analyzer are described in reactor transfer function measurements and nuclear instrumentation channels frequency response testing. (author)

  1. Design and analysis of pressurized water reactor systems

    International Nuclear Information System (INIS)

    Juhn, P.E.; Kim, Y.H.

    1979-01-01

    To help develop nuclear engineering technologies in local industry sectors, technical and economical data on pressurized water reactor systems and components have been collected, systematically analyzed and computerized to a certain degree. Codes and standards necessary for engineering design of PWR systems have been surveyed and clarified in terms of NSSS, turbine-generator system and BOP, then again rearranged with respect to quality classes and seismic classes. Some design manuals, criteria and guidelines regarding design, construction, test and operation of PWR plants have also been surveyed and collected. Benchmark cost calculation for the construction of a 900 MWe PWR plant, according to the standard format, was carried out, and computer model on construction costs was improved and updated by considering the local supply of labor and materials. And for the indigeneous development of PWR equipment and materials, such data as delivery schedule and manufacturers of 52 systems and 36,000 components have also been reviewed herein. (author)

  2. System Requirements Analysis for a Computer-based Procedure in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaek Wan; Jang, Gwi Sook; Seo, Sang Moon; Shin, Sung Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    This can address many of the routine problems related to human error in the use of conventional, hard-copy operating procedures. An operating supporting system is also required in a research reactor. A well-made CBP can address the staffing issues of a research reactor and reduce the human errors by minimizing the operator's routine tasks. A CBP for a research reactor has not been proposed yet. Also, CBPs developed for nuclear power plants have powerful and various technical functions to cover complicated plant operation situations. However, many of the functions may not be required for a research reactor. Thus, it is not reasonable to apply the CBP to a research reactor directly. Also, customizing of the CBP is not cost-effective. Therefore, a compact CBP should be developed for a research reactor. This paper introduces high level requirements derived by the system requirements analysis activity as the first stage of system implementation. Operation support tools are under consideration for application to research reactors. In particular, as a full digitalization of the main control room, application of a computer-based procedure system has been required as a part of man-machine interface system because it makes an impact on the operating staffing and human errors of a research reactor. To establish computer-based system requirements for a research reactor, this paper addressed international standards and previous practices on nuclear plants.

  3. Pressurized water reactor system model for control system design and analysis

    International Nuclear Information System (INIS)

    Cooper, K.F.; Cain, J.T.

    1975-01-01

    Satisfactory operation of present generation Pressurized Water Reactor (PWR) Nuclear Power systems requires that several independent and interactive control systems be designed. Since it is not practical to use an actual PWR system as a design tool, a mathematical model of the system must be developed as a design and analysis tool. The model presented has been developed to be used as an aid in applying optimal control theory to design and implement new control systems for PWR plants. To be applicable, the model developed must represent the PWR system in its normal operating range. For safety analysis the operating conditions of the system are usually abnormal and, therefore, the system modeling requirements are different from those for control system design and analysis

  4. A severe accident analysis for the system-integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Jae, Moosung

    2015-01-01

    The System-Integrated Modular Advanced Reactor (SMART) that has been recently designed in KOREA and has acquired standard design certification from the nuclear power regulatory body (NSSC) is an integral type reactor with 330MW thermal power. It is a small sized reactor in which the core, steam generator, pressurizer, and reactor coolant pump that are in existing pressurized light water reactors are designed to be within a pressure vessel without any separate pipe connection. In addition, this reactor has much different design characteristics from existing pressurized light water reactors such as the adoption of a passive residual heat removal system and a cavity flooding system. Therefore, the safety of the SMART against severe accidents should be checked through severe accident analysis reflecting the design characteristics of the SMART. For severe accident analysis, an analysis model has been developed reflecting the design information presented in the standard design safety analysis report. The severe accident analysis model has been developed using the MELCOR code that is widely used to evaluate pressurized LWR severe accidents. The steady state accident analysis model for the SMART has been simulated. According to the analysis results, the developed model reflecting the design of the SMART is found to be appropriate. Severe accident analysis has been performed for the representative accident scenarios that lead to core damage to check the appropriateness of the severe accident management plan for the SMART. The SMART has been shown to be safe enough to prevent severe accidents by utilizing severe accident management systems such as a containment spray system, a passive hydrogen recombiner, and a cavity flooding system. In addition, the SMART is judged to have been technically improved remarkably compared to existing PWRs. The SMART has been designed to have a larger reactor coolant inventory compared to its core's thermal power, a large surface area in

  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. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

    2005-03-01

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  7. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1983-01-01

    The invention provides a safety system for a nuclear reactor which uses a parallel combination of computer type look-up tables each of which receives data on a particular parameter (from transducers located in the reactor system) and each of which produces the functional counterpart of that particular parameter. The various functional counterparts are then added together to form a control signal for shutting down the reactor. The functional counterparts are developed by analysis of experimental thermal and hydraulic data, which are used to form expressions that define safe conditions

  8. 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

  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. Reliability analysis of reactor systems by applying probability method; Analiza pouzdanosti reaktorskih sistema primenom metoda verovatnoce

    Energy Technology Data Exchange (ETDEWEB)

    Milivojevic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1974-12-15

    Probability method was chosen for analysing the reactor system reliability is considered realistic since it is based on verified experimental data. In fact this is a statistical method. The probability method developed takes into account the probability distribution of permitted levels of relevant parameters and their particular influence on the reliability of the system as a whole. The proposed method is rather general, and was used for problem of thermal safety analysis of reactor system. This analysis enables to analyze basic properties of the system under different operation conditions, expressed in form of probability they show the reliability of the system on the whole as well as reliability of each component.

  11. Advanced reactor passive system reliability demonstration analysis for an external event

    International Nuclear Information System (INIS)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

    2017-01-01

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event

  12. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    Directory of Open Access Journals (Sweden)

    Matthew Bucknor

    2017-03-01

    Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

  13. Advanced reactor passive system reliability demonstration analysis for an external event

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

    2017-03-15

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

  14. Development of a central PC-based system for reactor signal monitoring and analysis

    International Nuclear Information System (INIS)

    Karim, A.; Ansari, S.A.; Rauf Baig, A.

    1998-01-01

    A personal computer based system was developed for on-line monitoring, signal processing and display of important reactor parameters of the Pakistan Research Reactor-1. The system was designed for assistance to both reactor operator and users. It performs three main functions. The first is the centralized radiation monitoring in and around the reactor building. The computer acquires signals from radiation monitoring channels and continuously displays them on distributed monitors. Trend monitoring and alarm generation is also done. In case of any abnormal condition the radiation level data is automatically stored in computer memory for detailed off-line analysis. In the second part the computer does the performance testing of nuclear instrumentation channels by signal statistical analysis, and generates alarm in case the channel standard deviation error exceeds the permissible error. Mean values of important nuclear signals are also displayed on distributed monitors as a part of reactor safety parameters display system. The third function is on-line computation of reactor physics parameters of the core which are important from operational and safety points-of-view. The signals from radiation protection system and nuclear instrumentation channels in the reactor were interfaced with the computer for this purpose. The development work was done under an IAEA research contract as a part of coordinated research programme. (author)

  15. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  16. Analysis of man-machine interaction for control and display system in main control room of light water reactor

    International Nuclear Information System (INIS)

    Santosa, Kussigit; Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Darlis; Sudiono, Bambang

    1998-01-01

    One of potential hazard in Nuclear Power Plant is the failure of its operation. The accident or operation failure in the reactor must be concerned event its probability is low. The important thing should be concerned is 'Analysis of Man-Machine Interaction (MMI) for Control and Display System in Main Control Room (MCR) of Nuclear Power Reactor', especially LWR type. Control and Display System in MCR of Reactor is the main part of MMI link process in Reactor MCR work system. Signal from display system showed performance process in reactor, while this signal will be received by operator. This signal will be described through central nerve for making decision what kind must be done. Then the operator manage the next process of reactor operation through control system. So by knowing Analysis of Man-Machine Interaction for Control and Display System in Main Control Room of Power Reactor, we can understand human error probability of the operator in reactor operation

  17. Reference ZrH reactor power system for NASA space station post-operational reentry analysis

    International Nuclear Information System (INIS)

    Elliott, R.D.

    1970-01-01

    The flight dynamic and heating of a spent ZrH reactor power system returning from orbit at the end of its useful life are analyzed. The results of this analysis indicate that the reactor with a large portion of the lithium shield still surrounding it will impact the earth at a velocity of from 660 to 820 ft/sec, depending upon whether it tumbles or becomes stabilized during the latter part of its trajectory. (U.S.)

  18. FAST: An advanced code system for fast reactor transient analysis

    International Nuclear Information System (INIS)

    Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh

    2005-01-01

    One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems

  19. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis. (orig./GL)

  20. Analysis for RSG-GAS operational characteristics of reactor cooling system

    International Nuclear Information System (INIS)

    Nurhappy, T.

    1998-01-01

    Analysis of operational characteristics of reactor cooling systems (JE01 and PA) is aimed at determining the effects of operation and maintenance patterns to the operational characteristic of the system. Analysis is carried out by virtue of the operating and maintenance data from 1987 to 1997, comprising the operating hours (duration) and data on operating failures of the systems. Results of study show that, either separately or jointly, the operating and maintenance patterns will qualitatively affect the operational characteristic of the systems

  1. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  2. Development of a central PC-based system for reactor signal monitoring and analysis

    International Nuclear Information System (INIS)

    Karim, A.; Ansari, S.A.; Baig, A.R.

    1996-05-01

    A personal computer based system was developed for on-line monitoring, signal processing and display of important parameters of the Pakistan Reactor-1. The system was designed for assistance to both reactor operator and users. It performs three main functions. The first is the centralized radiation monitoring in and around the reactor building. The computer acquires signals from radiation monitoring channels and continuously displays them on distributed monitors. Trend monitoring and alarm generation is also done. In case of any abnormal condition the radiation level data is automatically stored in computer memory for detailed off-line analysis. In the second part the computer does the performance testing of nuclear instrumentation channels by signal statistical analysis and generates alarm in case the channel standard deviation error exceeds the permissible error. Mean values of important nuclear signals are also displayed on distributed monitors as a part of reactor safety parameters display system. The third function is on-line computation of reactor physics parameters of the core which are important from operational and safety point-of-view. The signals from radiation protection system and nuclear instrumentation channels in the reactor were interfaced with the computer for this purpose. The development work was done under an IAEA research contract as a part of coordinated research programme. (author) 12 figs

  3. Development of advanced nuclear core analysis system applicable to various reactor types

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2002-03-01

    This fiscal year, aiming at development of an advanced detailed analysis system applicable to nuclear core performance analysis of various fast reactors currently considered, the concept of cross section library set was examined and the specification of library set was determined. That is to say, referring the world most advanced reactor physics analysis system ERANOS (European Reactor Analysis Optimized System) and the result of preceding research 'preparation of next generation cross section library', 900 energy groups structure, concrete cross section data to be included and the format of cross section library were defined. And we performed elaborate work revising the group cross section production system which was prepared in the preceding research. After that the revision work was completed, to confirm the capability of revised cross section production system, we produced a prototype 450 groups cross section library. And we carried out a series of bench mark tests including analysis of small fast reactors utilizing this prototype cross section library and confirmed that the prototype cross section library has sufficient accuracy for predicting core performance. Furthermore, we estimated the computer resource information such as memory size, hard disk capacity and calculation time, etc. necessary for producing 900 groups detailed cross section library. In addition, we identified problems to be solved for developing a cell calculation code installed in our detailed analysis system. (author)

  4. Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Honda, T.; Maki, K.; Okazaki, T.

    1994-01-01

    Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

  5. Optimizing the design and operation of reactor emergency systems using reliability analysis techniques

    International Nuclear Information System (INIS)

    Snaith, E.R.

    1975-01-01

    Following a reactor trip various reactor emergency systems, e.g. essential power supplies, emergency core cooling and boiler feed water arrangements are required to operate with a high degree of reliability. These systems must therefore be critically assessed to confirm their capability of operation and determine their reliability of performance. The use of probability analysis techniques enables the potential operating reliability of the systems to be calculated and this can then be compared with the overall reliability requirements. However, a system reliability analysis does much more than calculate an overall reliability value for the system. It establishes the reliability of all parts of the system and thus identifies the most sensitive areas of unreliability. This indicates the areas where any required improvements should be made and enables the overall systems' designs and modes of operation to be optimized, to meet the system and hence the overall reactor safety criteria. This paper gives specific examples of sensitive areas of unreliability that were identified as a result of a reliability analysis that was carried out on a reactor emergency core cooling system. Details are given of modifications to design and operation that were implemented with a resulting improvement in reliability of various reactor sub-systems. The report concludes that an initial calculation of system reliability should represent only the beginning of continuing process of system assessment. Data on equipment and system performance, particularly in those areas shown to be sensitive in their effect on the overall nuclear power plant reliability, should be collected and processed to give reliability data. These data should then be applied in further probabilistic analyses and the results correlated with the original analysis. This will demonstrate whether the required and the originally predicted system reliability is likely to be achieved, in the light of the actual history to date of

  6. 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

  7. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

  8. Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

    International Nuclear Information System (INIS)

    Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

    2016-01-01

    Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

  9. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    International Nuclear Information System (INIS)

    Park, Jong-beom; Park, No-Cheol; Lee, Sang-Jeong; Park, Young-Pil; Choi, Youngin

    2017-01-01

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  10. Seismic analysis of the APR1400 nuclear reactor system using a verified beam element model

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong-beom [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Park, No-Cheol, E-mail: pnch@yonsei.ac.kr [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Lee, Sang-Jeong; Park, Young-Pil [Department of Mechanical Engineering, Yonsei University, 50 Yonsei-ro, Seodaemun-gu, Seoul 03722 (Korea, Republic of); Choi, Youngin [Korea Institute of Nuclear Safety, 62 Gwahak-ro, Yuseong-gu, Daejeon 34142 (Korea, Republic of)

    2017-03-15

    Highlights: • A simplified beam element model is constructed based on the real dynamic characteristics of the APR1400. • Time history analysis is performed to calculate the seismic responses of the structures. • Large deformations can be observed at the in-phase mode of reactor vessel and core support barrel. - Abstract: Structural integrity is the first priority in the design of nuclear reactor internal structures. In particular, nuclear reactor internals should be designed to endure external forces, such as those due to earthquakes. Many researchers have performed finite element analyses to meet these design requirements. Generally, a seismic analysis model should reflect the dynamic characteristics of the target system. However, seismic analysis based on the finite element method requires long computation times as well as huge storage space. In this research, a beam element model was developed and confirmed based on the real dynamic characteristics of an advanced pressurized water nuclear reactor 1400 (APR1400) system. That verification process enhances the accuracy of the finite element analysis using the beam elements, remarkably. Also, the beam element model reduces seismic analysis costs. Therefore, the beam element model was used to perform the seismic analysis. Then, the safety of the APR1400 was assessed based on a seismic analysis of the time history responses of its structures. Thus, efficient, accurate seismic analysis was demonstrated using the proposed beam element model.

  11. 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)

  12. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  13. System analysis of nuclear safety of VVER reactor with MOX fuel

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, A.D.; Zharkov, V.P.; Suslov, I.R. [Russia, Moscow Malaya Krasnoselskaya St. (Russian Federation); Boyarinov, V.F.; Kevrolev, V.V.; Tchibinyaev, A.V.; Tsibulskiy, V.F. [RRC KI, Russia, Moscow (Russian Federation); Kochurov, B.P. [ITEP, Russia, Moscow (Russian Federation); Giovanni, B. [NFPSC, FRAMATOME (France)

    2005-07-01

    The report presents a short summary of the results achieved in the ISTC (International Science and Technology Center) project 'System analysis of nuclear safety of VVER reactor with MOX fuel' (April 2005). The studies within the project are of a systematic character and include the solutions of 15 tasks. The report gives an overview of the major blocks of these tasks: neutron transport equation solution; calculations of isotopic vectors, analysis of the impact of uncertainties on predicted reactor functionals. The calculation methods, the verification results and the corresponding codes are briefly described. (authors)

  14. A seismic analysis of the driving system for the pulsed reactor

    International Nuclear Information System (INIS)

    Hu Yongtao; Fu Shixiang; Zeng Jianhua; Hong Jingfeng

    1991-01-01

    The driving system of the pulsed reactor contains control rods, pulsing o rod and sample rack. They are slender, and their drive function is required more strictly. First, a complete model which contains all driving system and reactor bridge is used. Then the substructure models are adopted. The results of calculation are compared with the experimental results. It shows that the analysis results are reliable and the substructure method is simple, available and utility. The seismic safety is evaluated by the results from response spectra method

  15. Simplified analysis of PRISM RVACS [Reactor Vessel Auxiliary Cooling System] performance without liner spill-over

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1990-01-01

    Simplified analysis of the performance of the PRISM RVACS decay heat removal system under off-normal conditions, i.e., without the liner spill-over, is described. Without the spilling of hot-pool sodium over the liner and the resultant down-flow along the inside of the reactor vessel wall, the RVACS system performance becomes dominated by the radial heat condition and radiation. Simple estimates of the resulting heat conduction and radiation processes support GE's contention that the RVACS performance is not severely impacted by the absence of spillover, and can improve significantly if sodium has leaked into the region between the reactor and containment vessels. 7 refs

  16. Scram and nonlinear reactor system seismic analysis for the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Morrone, A.

    1975-01-01

    A description is given of the analysis and results for the Fast Flux Test Facility (FFTF) reactor system which was analyzed for both scram times and seismic responses such as bending moments and impact forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node. The results give time history plots of various seismic responses and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. (U.S.)

  17. A strategy analysis of the fast breeder reactor introduction and nuclear fuel cycle systems deployment

    International Nuclear Information System (INIS)

    Wajima, Tsunetaka; Kawashima, Katsuyuki; Yamashita, Takashi

    1996-01-01

    A study is made on a strategy analysis of the long term nuclear fuel cycle systems deployment in accordance with the nuclear power growth projection and fast breeder reactor (FBR) introduction. In the analysis, the reprocessed plutonium (Pu) is charged into the reactor in such a way that the reprocessed Pu is not stored outside the reactor, i.e., there is no excess Pu outside the reactor. The analysis characterized the fuel cycle systems, and showed the usefulness of the present method to determine future directions for the FBR introduction and nuclear fuel cycle systems deployment. Concerning an intermediate-term strategy, the time of introduction and required capacities of a second commercial LWR reprocessing plant, Pu-thermal, and the first FBR reprocessing plant deployment are evaluated. A long term strategy analysis shows that the two or three large plants are run in parallel for each fuel cycle facility and that FBR related facilities deal with a markedly large amount of Pu. It is concluded that the early stage introduction of FBRs of significant capacities seems necessary to materialize a consistent total FBR/fuel cycle system where Pu balance becomes feasible through its flexible operation of, for instance, adjusting breeding ratio, in order to keep the transparency of the Pu utilization. (author)

  18. CENTAR code for extended nonlinear transient analysis of extraterrestrial reactor systems

    International Nuclear Information System (INIS)

    Nassersharif, B.; Peer, J.S.; DeHart, M.D.

    1987-01-01

    Current interest in the application of nuclear reactor-driven power systems to space missions has generated a need for a systems simulation code to model and analyze space reactor systems; such a code has been initiated at Texas A and M, and the first version is nearing completion; release was anticipated in the fall of 1987. This code, named CENTAR (Code for Extended Nonlinear Transient Analysis of Extraterrestrial Reactor Systems), is designed specifically for space systems and is highly vectorizable. CENTAR is composed of several specialized modules. A fluids module is used to model fluid behavior throughout the system. A wall heat transfer module models the heat transfer characteristics of all walls, insulation, and structure around the system. A fuel element thermal analysis module is used to predict the temperature behavior and heat transfer characteristics of the reactor fuel rods. A kinetics module uses a six-group point kinetics formulation to model reactivity feedback and control and the ANS 5.1 decay-heat curve to model shutdown decay-heat production. A pump module models the behavior of thermoelectric-electromagnetic pumps, and a heat exchanger module models not only thermal effects in thermoelectric heat exchangers, but also predicts electrical power production for a given configuration. Finally, an accumulator module models coolant expansion/contraction accumulators

  19. Systems analysis of a 100-MWe modular liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Morris, E.E.; Rhow, S.K.; Switick, D.M.

    1985-01-01

    The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

  20. 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

  1. 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

  2. Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

  3. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Seo, K. W

    2006-01-15

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

  4. Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.

    2006-01-01

    A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

  5. Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

    International Nuclear Information System (INIS)

    2014-02-01

    The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

  6. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

    JAPC purchased RETRAN, a program for transient thermal hydraulic analysis of complex fluid flow system, from the U.S. Electric Power Research Institute in 1992. Since then, JAPC has been utilizing RETRAN to evaluate safety margins of actual plant operation, in coping with troubles (investigating trouble causes and establishing countermeasures), and supporting reactor operation (reviewing operational procedures etc.). In this paper, a result of plant analysis performed on a CVCS reactor primary coolant leakage incident which occurred at JAPC's Tsuruga-2 plant (4-loop PWR, 3423 MWt, 1160 MW) on July 12 of 1999 and, based on the result, we made a plan to modify our operational procedure for reactor primary coolant leakage events in order to make earlier plant shutdown and this reduced primary coolant leakage. (author)

  7. Dynamic analysis of Korean nuclear fuel cycle with fast reactor systems

    International Nuclear Information System (INIS)

    Jeong, Chang Joon

    2004-12-01

    The Korean nuclear fuel cycle scenario was analyzed by the dynamic analysis method, including Pressurized Water Reactor (PWR), Canadian Deuterium Uranium (CANDU) and fast reactor systems. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 1%. After setting up the once-through fuel cycle model, the Korea Advanced Liquid Metal Reactor (KALIMER) scenario was modeled to investigate the fuel cycle parameters. For the analysis of the fast reactor fuel cycle, both KAILMER-150 and KALIMER-600 reactors were considered. In this analysis, the spent fuel inventory as well as the amount of plutonium, Minor Actinides (MA) and Fission Products (FP) of the recycling fuel cycle was estimated and compared to that of the once-through fuel cycle. Results of the once-through fuel cycle calculation showed that the demand grows up to 64 GWe and total amount of spent fuel would be ∼102 kt in 2100. If the KALIMER scenario is implemented, the total spent fuel inventory can be reduced by ∼80%. However it was found that the KALIMER scenario does not contribute to reduce the amount of MA and FP, which is important when designing a repository. For the further destruction of MA, an actinide burner can be considered in the future nuclear fuel cycle

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

    International Nuclear Information System (INIS)

    Bonfietti, Gerson

    2003-01-01

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

  9. Reactor operational transient analysis

    International Nuclear Information System (INIS)

    Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.

    1983-01-01

    To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)

  10. SAFIRE: A systems analysis code for ICF [inertial confinement fusion] reactor economics

    International Nuclear Information System (INIS)

    McCarville, T.J.; Meier, W.R.; Carson, C.F.; Glasgow, B.B.

    1987-01-01

    The SAFIRE (Systems Analysis for ICF Reactor Economics) code incorporates analytical models for scaling the cost and performance of several inertial confinement fusion reactor concepts for electric power. The code allows us to vary design parameters (e.g., driver energy, chamber pulse rate, net electric power) and evaluate the resulting change in capital cost of power plant and the busbar cost of electricity. The SAFIRE code can be used to identify the most attractive operating space and to identify those design parameters with the greatest leverage for improving the economics of inertial confinement fusion electric power plants

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

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N J.H.

    1966-12-15

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

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

    International Nuclear Information System (INIS)

    Schuh, N.J.H.

    1966-12-01

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

  13. PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.

    1996-09-01

    The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)

  14. A ''SUPERCODE'' for systems analysis of tokamak experiments and reactors

    International Nuclear Information System (INIS)

    Haney, S.W.; Barr, W.L.; Crotinger, J.A.; Perkins, L.J.; Solomon, C.J.; Chaniotakis, E.A.; Freidberg, J.P.; Wei, J.; Mandrekas, J.

    1992-01-01

    A new code, named SUPERCODE, has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of their physics and engineering models. The latter contain state-of-the-art plasma physics modeling but are limited in engineering content and are time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2-D plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formulation there is only a modest increase in CPU time over 0-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, we have expended considerable effort to make the code user- and programmer friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community

  15. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1991-01-01

    In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors

  16. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs

  17. Nuclear reactor shutdown system

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  18. Reliability analysis of diverse safety logic systems of fast breeder reactor

    International Nuclear Information System (INIS)

    Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

    2006-01-01

    Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)

  19. Development of an advanced code system for fast-reactor transient analysis

    International Nuclear Information System (INIS)

    Konstantin Mikityuk; Sandro Pelloni; Paul Coddington

    2005-01-01

    FAST (Fast-spectrum Advanced Systems for power production and resource management) is a recently approved PSI activity in the area of fast spectrum core and safety analysis with emphasis on generic developments and Generation IV systems. In frames of the FAST project we will study both statics and transients core physics, reactor system behaviour and safety; related international experiments. The main current goal of the project is to develop unique analytical and code capability for core and safety analysis of critical (and sub-critical) fast spectrum systems with an initial emphasis on a gas cooled fast reactors. A structure of the code system is shown on Fig. 1. The main components of the FAST code system are 1) ERANOS code for preparation of basic x-sections and their partial derivatives; 2) PARCS transient nodal-method multi-group neutron diffusion code for simulation of spatial (3D) neutron kinetics in hexagonal and square geometries; 3) TRAC/AAA code for system thermal hydraulics; 4) FRED transient model for fuel thermal-mechanical behaviour; 5) PVM system as an interface between separate parts of the code system. The paper presents a structure of the code system (Fig. 1), organization of interfaces and data exchanges between main parts of the code system, examples of verification and application of separate codes and the system as a whole. (authors)

  20. TRIGA reactor main systems

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  1. Perspectives on the development of next generation reactor systems safety analysis codes

    International Nuclear Information System (INIS)

    Zhang, H.

    2015-01-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  2. Perspectives on the development of next generation reactor systems safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States)

    2015-07-01

    'Full text:' Existing reactor system analysis codes, such as RELAP5-3D and TRAC, have gained worldwide success in supporting reactor safety analyses, as well as design and licensing of new reactors. These codes are important assets to the nuclear engineering research community, as well as to the nuclear industry. However, most of these codes were originally developed during the 1970s', and it becomes necessary to develop next-generation reactor system analysis codes for several reasons. Firstly, as new reactor designs emerge, there are new challenges emerging in numerical simulations of reactor systems such as long lasting transients and multi-physics phenomena. These new requirements are beyond the range of applicability of the existing system analysis codes. Advanced modeling and numerical methods must be taken into consideration to improve the existing capabilities. Secondly, by developing next-generation reactor system analysis codes, the knowledge (know how) in two phase flow modeling and the highly complex constitutive models will be transferred to the young generation of nuclear engineers. And thirdly, all computer codes have limited shelf life. It becomes less and less cost-effective to maintain a legacy code, due to the fast change of computer hardware and software environment. There are several critical perspectives in terms of developing next-generation reactor system analysis codes: 1) The success of the next-generation codes must be built upon the success of the existing codes. The knowledge of the existing codes, not just simply the manuals and codes, but knowing why and how, must be transferred to the next-generation codes. The next-generation codes should encompass the capability of the existing codes. The shortcomings of existing codes should be identified, understood, and properly categorized, for example into model deficiencies or numerical method deficiencies. 2) State-of-the-art models and numerical methods must be considered to

  3. A CAMAC based real-time noise analysis system for nuclear reactors

    International Nuclear Information System (INIS)

    Ciftcioglu, O.

    1987-01-01

    A CAMAC based real-time noise analysis system was designed for the TRIGA MARK II nuclear reactor at the Institute for Nuclear Energy, Istanbul. The input analog signals obtained from the radiation detectors are introduced to the system through CAMAC interface. The signals coverted into digital form are processed by a PDP-11 computer. The fast data processing based on auto/cross power spectral density computations is carried out by means of assembly written FFT algorithms in real-time and the spectra obtained are displayed on a CAMAC driven display system as an additional monitoring device. The system has the advantage of being software programmable and controlled by a CAMAC system so that it is operated under porgram control for reactor surveillance, anomaly detection and diagnosis. The system can also be used for the identification of nonstationary operational characteristics of the reactor in long term by comparing the noise power spectra with the corresponding reference noise patterns prepared in advance. (orig.)

  4. Seismic analysis of the reactor coolant system of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Borsoi, L.; Sollogoub, P.

    1986-01-01

    For safety considerations, seismic analyses are performed of the Reactor Coolant System (R.C.S.) of PWR Plants. After a brief description of the R.C.S. and R.C.S. operation, the paper presents the two types of analysis used to determine the effect of earthquake on the R.C.S.: modal spectral analysis and nonlinear time history analysis. The paper finally shows how seismic loadings are combined with other types of loadings and illustrates how the consideration of seismic loads affects R.C.S. design [fr

  5. 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

  6. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    International Nuclear Information System (INIS)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given

  7. Dynamic analysis of the condensate feedwater system in boiling water reactor plants

    International Nuclear Information System (INIS)

    Tanji, J.; Omori, T.

    1982-01-01

    The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

  8. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  9. Nuclear reactor safety systems

    International Nuclear Information System (INIS)

    Ball, R.M.; Roberts, R.C.

    1980-01-01

    A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

  10. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    International Nuclear Information System (INIS)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

  11. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    International Nuclear Information System (INIS)

    Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

    2010-10-01

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  12. Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

  13. 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.

  14. Analysis on Configuration of I and C Systems for an Advanced HANARO Reactors

    International Nuclear Information System (INIS)

    Park, Gee Yong; Jung, H. S.; Ryu, J. S.; Park, C.

    2006-01-01

    In an advanced HANARO reactor (AHR), the instrumentation and control (I and C) systems are designed based on the digital system rather than the analog system installed in an existing HANARO instrumentation and control systems. While the safety and functionality of analog-based instrumentation and control system are experienced over a long period of operating time and also well-validated, the obsolescence and the lack of flexibility of this system have to move from the analog technology to the digital technology in the instrumentation and control systems to be used in nuclear power plants as well as nuclear research reactors. For establishing the adequate structure of instrumentation and control systems for an AHR, various instrumentation and control architectures are analyzed for their merits and demerits for use in I and C systems of an AHR and the most promising instrumentation and control architecture for an AHR are drawn from this analysis. The conceptual configuration of a digital-based safety shutdown system is proposed in this report

  15. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  16. Reactor surveillance by noise analysis

    International Nuclear Information System (INIS)

    Ciftcioglu, Ozer

    1988-01-01

    A real-time noise analysis system is designed for the TRIGA reactor at Istanbul Technical University. By means of the noise techniques, reactor surveillance is performed together with failure diagnosis. The fast data processing is carried out by FFT in real-time so that malfunction or non-stationary operation of the reactor in long term can be identified by comparing the noise power spectra with the corresponding reference patterns while the decision making procedure is accomplished by the method of hypothesis testing. The system being computer based safety instrumentation involves CAMAC in conjunction with the RT-11 (PDP-11) single user dedicated environment. (author)

  17. Analysis of dynamic stability and safety of reactor system by reactor simulator; Analiza dinamicke stabilnosti i sigurnosti reaktorskog sistema pomocu reaktorskog simulatora

    Energy Technology Data Exchange (ETDEWEB)

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

    1963-11-15

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

  18. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    International Nuclear Information System (INIS)

    Shi, Chengbin; Cheng, Maosong; Liu, Guimin

    2016-01-01

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  19. Development and application of a system analysis code for liquid fueled molten salt reactors based on RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

    2016-08-15

    Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center

  20. Requirement analysis and architecture of data communication system for integral reactor

    International Nuclear Information System (INIS)

    Jeong, K. I.; Kwon, H. J.; Park, J. H.; Park, H. Y.; Koo, I. S.

    2005-05-01

    When digitalizing the Instrumentation and Control(I and C) systems in Nuclear Power Plants(NPP), a communication network is required for exchanging the digitalized data between I and C equipments in a NPP. A requirements analysis and an analysis of design elements and techniques are required for the design of a communication network. Through the requirements analysis of the code and regulation documents such as NUREG/CR-6082, section 7.9 of NUREG 0800 , IEEE Standard 7-4.3.2 and IEEE Standard 603, the extracted requirements can be used as a design basis and design concept for a detailed design of a communication network in the I and C system of an integral reactor. Design elements and techniques such as a physical topology, protocol transmission media and interconnection device should be considered for designing a communication network. Each design element and technique should be analyzed and evaluated as a portion of the integrated communication network design. In this report, the basic design requirements related to the design of communication network are investigated by using the code and regulation documents and an analysis of the design elements and techniques is performed. Based on these investigation and analysis, the overall architecture including the safety communication network and the non-safety communication network is proposed for an integral reactor

  1. Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSAS is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  2. Reactor safety assessment system

    International Nuclear Information System (INIS)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category

  3. Development of automatic reactor vessel inspection systems: development of data acquisition and analysis system for the nuclear vessel weld

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. H.; Lim, H. T.; Um, B. G. [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2001-03-01

    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine the reactor vessel weldsIn order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed in this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition and analysis software was developed. 11 refs., 6 figs., 9 tabs. (Author)

  4. Analysis and upgrade of instrumentation and control systems for the modernization of research reactors

    International Nuclear Information System (INIS)

    1988-01-01

    This document provides assistance in the review and planning process for the upgrade of instrumentation and control systems (I and C systems) and related safety features of the reactor protection system for research reactors. In the interest of safety a need was realized to evaluate the performance of outdated I and C systems. An advisory group was assembled to develop guidelines and to provide recommendations for the upgrade of I and C systems. The recommendations on I and C systems upgrade contained in this document were developed by the advisory group using as guidelines the established safety criteria and operating standards for research reactors. 24 refs

  5. RETRAN code analysis of Tsuruga-2 plant chemical volume control system (CVCS) reactor coolant leakage incident

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    2002-01-01

    In the Chemical Volume Control System (CVCS) reactor primary coolant leakage incident, which occurred in Tsuruga-2 (4-loop PWR, 3,423 MWt, 1,160 MWe) on July 12, 1999, it took about 14 hours before the leakage isolation. The delayed leakage isolation and a large amount of leakage have become a social concern. Effective procedure modification was studied. Three betterments were proposed based on a qualitative analysis to reduce the pressure and temperature of the primary loop as fast as possible by the current plant facilities while maintaining enough subcooling of the primary loop. I analyzed the incident with RETRAN code in order to quantitatively evaluate the leakage reduction when these betterments are adopted. This paper is very new because it created a typical analysis method for PWR plant behavior during plant shutdown procedure which conventional RETRAN transient analyses rarely dealt with. Also the event time is very long. To carry out this analysis successfully, I devised new models such as an Residual Heat Removal System (RHR) model etc. and simplified parts of the conventional model. Based on the analysis results, I confirmed that leakage can be reduced by about 30% by adopting these betterments. Then the Japan Atomic Power Company (JAPC) modified the operational procedure for reactor primary coolant leakage events adopting these betterments. (author)

  6. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    International Nuclear Information System (INIS)

    Rahardjo, H.P.

    2011-01-01

    Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipment. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogeneously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occurred in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code) was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs) of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA) in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, expansion load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (re-routing) is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node 30, also a

  7. Analysis of transition to fuel cycle system with continuous recycling in fast and thermal reactors - 5060

    International Nuclear Information System (INIS)

    Passereini, S.; Feng, B.; Fei, T.; Kim, T.K.; Taiwo, T.A.; Brown, N.R.; Cuadra, A.

    2015-01-01

    A recent Evaluation and Screening study of nuclear fuel cycle options identified a few groups of options as most promising. One of these most promising Evaluation Groups (EGs) is characterized by the continuous recycling of uranium (U) and transuranics (TRU) with natural uranium feed in both fast and thermal critical reactors. This evaluation group, designated as EG30, is represented by an example fuel cycle option that employs a two-technology, two-stage fuel cycle system. The first stage involves the continuous recycling of co-extracted U/TRU in Sodium-cooled Fast Reactors (SFRs) with metallic fuel and breeding ratio greater than 1. The second stage involves the use of the surplus TRU in Mixed Oxide (MOX) fuel in Pressurized Water Reactors that are MOX-capable (MOX-PWRs). This paper presents and discusses preliminary fuel cycle analysis results from the fuel cycle codes VISION and DYMOND for the transition to this fuel cycle option from the current once-through cycle in the United States (U.S.) that consists of Light Water Reactors (LWRs) that only use conventional UO 2 fuel. The analyses in this paper are applicable for a constant 100 GWe capacity, roughly the size of the U.S. nuclear fleet. Two main strategies for the transition to EG30 were analyzed: 1) deploying both SFRs and MOX-PWRs in parallel or 2) deploying them in series with the SFR fleet first. With an estimated retirement schedule for the existing LWRs, an assumed reactor lifetime of 60 years, and no growth, the nuclear system fully transitions to the new fuel cycle within 100 years for both strategies without SFR fuel shortages. Compared to the once-through cycle, transition to the SFR/MOX-PWR fleet with continuous recycle was shown to offer significant reductions in uranium consumption and waste disposal requirements. In addition, these initial calculations revealed a few notable modeling and strategy questions regarding how recycled resources are allocated, reactors that can switch between

  8. FFTF reactor assembly system technology

    International Nuclear Information System (INIS)

    Mangelsdorf, T.A.

    1975-01-01

    An overview is presented of the FFTF reactor and plant together with descriptions of core components, core internals, core system, primary and secondary control rod system, reactor instrumentation, reactor vessel and closure head, and supporting test programs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-06-21

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

  10. Computational analysis of supercritical CO2 Brayton cycle power conversion system for fusion reactor

    International Nuclear Information System (INIS)

    Halimi, Burhanuddin; Suh, Kune Y.

    2012-01-01

    Highlights: ► Computational analysis of S-CO 2 Brayton cycle power conversion system. ► Validation of numerical model with literature data. ► Recompression S-CO 2 Brayton cycle thermal efficiency of 42.44%. ► Reheating concept to enhance the cycle thermal efficiency. ► Higher efficiency achieved by the proposed concept. - Abstract: The Optimized Supercritical Cycle Analysis (OSCA) code is being developed to analyze the design of a supercritical carbon dioxide (S-CO 2 ) driven Brayton cycle for a fusion reactor as part of the Modular Optimal Balance Integral System (MOBIS). This system is based on a recompression Brayton cycle. S-CO 2 is adopted as the working fluid for MOBIS because of its easy availability, high density and low chemical reactivity. The reheating concept is introduced to enhance the cycle thermal efficiency. The helium-cooled lithium lead model AB of DEMO fusion reactor is used as reference in this paper.

  11. Analysis of Reactor Pressurized Thermal Shock Conditions Considering Upgrading of Systems Important to Safety

    International Nuclear Information System (INIS)

    Mazurok, A.S; Vyshemirskyij, M.P.

    2015-01-01

    The paper analyzes conditions of pressurized thermal shock on the reactor pressure vessel taking into account upgrading of the emergency core cooling system and primary overpressure protection system. For representative accident scenarios, calculation and comparative analysis was carried out. These scenarios include a small leak from the hot leg and PRZ SV stuck opening with re closure after 3600 sec and 3 SG heat transfer tube rupture. The efficiency of mass flow control by valves on the pump head (emergency core cooling systems) and cold overpressure protection (primary overpressure protection system) was analyzed. The thermal hydraulic model for RELAP5/Mod3.2 code with detailed downcomer (DC) model and changes in accordance with upgrades was used for calculations. Detailed (realistic) modeling of piping and equipment was performed. The upgrades prevent excessive primary cooling and, consequently, help to preserve the RPV integrity and to avoid the formation of a through crack, which can lead to a severe accident

  12. Application of RELAP/SCDAPSIM with integrated uncertainty options to research reactor systems thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Allison, C.M.; Hohorst, J.K.; Perez, M.; Reventos, F.

    2010-01-01

    The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. One such member developed option is an integrated uncertainty analysis package being developed jointly by the Technical University of Catalonia (UPC) and Innovative Systems Software (ISS). This paper briefly summarizes the features of RELAP/SCDAPSIM/MOD4.0 and the integrated uncertainty analysis package, and then presents an example of how the integrated uncertainty package can be setup and used for a simple pipe flow problem. (author)

  13. A sensitivity analysis and assessment on the reactivity, economics and resorce implications of reactor systems and cycles with respect to uncertainity in nuclear data and other reactor parameters

    International Nuclear Information System (INIS)

    Quan, B.L.

    1980-01-01

    A general sensitivity analysis system for analyzing the effects of uncertainity in nuclear data and reactor parameters on fuel cycle economics, resources and physics has been developed. The sensitivity analysis has been performed on various reactor systems and cycles such as the thorium cycles, plutonium cycles, CANDU reactor fuel cycles and alternate once-through LWR cycles such as the 18 month cycle. Sensitivity coefficients were generated for a variety of materials pertinent to the LWR fuel cycle using a series of fast running codes developed for this purpose and running on a local PDP-15 computer. Their relative order of importance were assessed and the reasons explaining this difference were examined. This work is a result of EPRI project in determining the data needs for the LWR industry and should be valuable in identifying areas in which data improvements are worthwhile

  14. Development of Input/Output System for the Reactor Transient Analysis System (RETAS)

    International Nuclear Information System (INIS)

    Suh, Jae Seung; Kang, Doo Hyuk; Cho, Yeon Sik; Ahn, Seung Hoon; Cho, Yong Jin

    2009-01-01

    A Korea Institute of Nuclear Safety Reactor Transient Analysis System (KINS-RETAS) aims at providing a realistic prediction of core and RCS response to the potential or actual event scenarios in Korean nuclear power plants (NPPs). A thermal hydraulic system code MARS is a pivot code of the RETAS, and used to predict thermal hydraulic (TH) behaviors in the core and associated systems. MARS alone can be applied to many types of transients, but is sometimes coupled with the other codes developed for different objectives. Many tools have been developed to aid users in preparing input and displaying the transient information and output data. Output file and Graphical User Interfaces (GUI) that help prepare input decks, as seen in SNAP (Gitnick, 1998), VISA (K.D. Kim, 2007) and display aids include the eFAST (KINS, 2007). The tools listed above are graphical interfaces. The input deck builders allow the user to create a functional diagram of the plant, pictorially on the screen. The functional diagram, when annotated with control volume and junction numbers, is a nodalization diagram. Data required for an input deck is entered for volumes and junctions through a mouse-driven menu and pop-up dialog; after the information is complete, an input deck is generated. Display GUIs show data from MARS calculations, either during or after the transient. The RETAS requires the user to first generate a set of 'input', two dimensional pictures of the plant on which some of the data is displayed either numerically or with a color map. The RETAS can generate XY-plots of the data. Time histories of plant conditions can be seen via the plots or through the RETAS's replay mode. The user input was combined with design input from MARS developers and experts from both the GUI and ergonomics fields. A partial list of capabilities follows. - 3D display for neutronics. - Easier method (less user time and effort) to generate 'input' for the 3D displays. - Detailed view of data at volume or

  15. Development of Input/Output System for the Reactor Transient Analysis System (RETAS)

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jae Seung; Kang, Doo Hyuk; Cho, Yeon Sik [ENESYS, Daejeon (Korea, Republic of); Ahn, Seung Hoon; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    A Korea Institute of Nuclear Safety Reactor Transient Analysis System (KINS-RETAS) aims at providing a realistic prediction of core and RCS response to the potential or actual event scenarios in Korean nuclear power plants (NPPs). A thermal hydraulic system code MARS is a pivot code of the RETAS, and used to predict thermal hydraulic (TH) behaviors in the core and associated systems. MARS alone can be applied to many types of transients, but is sometimes coupled with the other codes developed for different objectives. Many tools have been developed to aid users in preparing input and displaying the transient information and output data. Output file and Graphical User Interfaces (GUI) that help prepare input decks, as seen in SNAP (Gitnick, 1998), VISA (K.D. Kim, 2007) and display aids include the eFAST (KINS, 2007). The tools listed above are graphical interfaces. The input deck builders allow the user to create a functional diagram of the plant, pictorially on the screen. The functional diagram, when annotated with control volume and junction numbers, is a nodalization diagram. Data required for an input deck is entered for volumes and junctions through a mouse-driven menu and pop-up dialog; after the information is complete, an input deck is generated. Display GUIs show data from MARS calculations, either during or after the transient. The RETAS requires the user to first generate a set of 'input', two dimensional pictures of the plant on which some of the data is displayed either numerically or with a color map. The RETAS can generate XY-plots of the data. Time histories of plant conditions can be seen via the plots or through the RETAS's replay mode. The user input was combined with design input from MARS developers and experts from both the GUI and ergonomics fields. A partial list of capabilities follows. - 3D display for neutronics. - Easier method (less user time and effort) to generate 'input' for the 3D displays. - Detailed view

  16. Nuclear reactor monitoring system

    International Nuclear Information System (INIS)

    Drummond, C.N.; Bybee, R.T.; Mason, F.L.; Worsham, H.J.

    1976-01-01

    The invention pertains to an improved monitoring system for the neutron flux in a nuclear reactor. It is proposed to combine neutron flux detectors, a thermoelement, and a background radiation detector in one measuring unit. The spatial arrangement of these elements is fixed with great exactness; they are enclosed by an elastic cover and are brought into position in the reactor with the aid of a bent tube. The arrangement has a low failure rate and is easy to maintain. (HP) [de

  17. Reactor feedwater system

    International Nuclear Information System (INIS)

    Kagaya, Hiroyuki; Tominaga, Kenji.

    1993-01-01

    In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

  18. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  19. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  20. Incorporation of personal computers in a research reactor instrumentation system for data monitoring and analysis

    International Nuclear Information System (INIS)

    Leopando, L.S.

    1998-01-01

    The research contract was implemented by obtaining off-the shelf personal computer hardware and data acquisition cards, designing the interconnection with the instrumentation system, writing and debugging the software, and the assembling and testing the set-up. The hardware was designed to allow all variables monitored by the instrumentation system to be accessible to the computers, without requiring any major modification of the instrumentation system and without compromising reactor safety in any way. The computer hardware addition was also designed to have no effect on any existing function of the instrumentation system. The software was designed to implement only graphical display and automated logging of reactor variables. Additional functionality could be easily added in the future with software revision because all the reactor variables are already available in the computer. It would even be possible to ''close the loop'' and control the reactor through software. It was found that most of the effort in an undertaking of this sort will be in software development, but the job can be done even by non-computer specialized reactor people working with programming languages they are already familiar with. It was also found that the continuing rapid advance of personal computer technology makes it essential that such a project be undertaken with inevitability of future hardware upgrading in mind. The hardware techniques and the software developed may find applicability in other research reactors, especially those with a generic analog research reactor TRIGA console. (author)

  1. Monte Carlo Analysis of the Accelerator-Driven System at Kyoto University Research Reactor Institute

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2016-04-01

    Full Text Available An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan, a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft–Walton type accelerator, which generates the external neutron source by deuterium–tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

  2. Monte Carlo analysis of the accelerator-driven system at Kyoto University Research Reactor Institute

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Kyeong; Lee, Deok Jung [Nuclear Engineering Division, Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Lee, Hyun Chul [VHTR Technology Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Pyeon, Cheol Ho [Nuclear Engineering Science Division, Kyoto University Research Reactor Institute, Osaka (Japan); Shin, Ho Cheol [Core and Fuel Analysis Group, Korea Hydro and Nuclear Power Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    An accelerator-driven system consists of a subcritical reactor and a controllable external neutron source. The reactor in an accelerator-driven system can sustain fission reactions in a subcritical state using an external neutron source, which is an intrinsic safety feature of the system. The system can provide efficient transmutations of nuclear wastes such as minor actinides and long-lived fission products and generate electricity. Recently at Kyoto University Research Reactor Institute (KURRI; Kyoto, Japan), a series of reactor physics experiments was conducted with the Kyoto University Critical Assembly and a Cockcroft-Walton type accelerator, which generates the external neutron source by deuterium-tritium reactions. In this paper, neutronic analyses of a series of experiments have been re-estimated by using the latest Monte Carlo code and nuclear data libraries. This feasibility study is presented through the comparison of Monte Carlo simulation results with measurements.

  3. 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

  4. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Alvim, Antonio C.M.

    2017-01-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  5. Scale analysis of decay heat removal system between HTR-10 and HTR-PM reactors under accidental conditions

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D.; Alvim, Antonio C.M. [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Lapa, Celso M.F., E-mail: thiagodbtr@gmail.com, E-mail: lapa@ien.gov.br, E-mail: alvim@nuclear.ufrj.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    The 10 MW high-temperature gas-cooled test module (HTR-10) is a graphite-moderated and helium-cooled pebble bed reactor prototype that was designed to demonstrate the technical and safety feasibility of this type of reactor project under normal and accidental conditions. In addition, one of the systems responsible for ensuring the safe operation of this type of reactor is the passive decay heat removal system (DHRS), which operates using passive heat removal processes. A demonstration of the heat removal capacity of the DHRS under accidental conditions was analyzed based on a benchmark problem for design-based accidents on an HTR-10, i.e., the pressurized loss of forced cooling (PLOFC) described in technical reports produced by the International Atomic Energy Agency. In fact, the HTR-10 is also a proof-of-concept reactor for the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), which generates approximately 25 times more heat than the HTR-10, with a thermal power of 250 MW, thereby requiring a DHRS with a higher system capacity. Thus, because an HTR-10 is a prototype reactor for an HTR-PM, a scaling analysis of the heat transfer process from the reactor to the DHRS was carried out between the HTR-10 and HTR-PM systems to verify the distortions of scale and the differences between the main dimensionless numbers from the two projects. (author)

  6. Development of advanced nuclear core analysis system applicable to various reactor types (II)

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2003-03-01

    A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

  7. Developing a PC-based expert system for fault analysis of reactor instruments

    International Nuclear Information System (INIS)

    Diwakar, M.P.; Rathod, N.C.; Bairi, B.R.; Darbhe, M.D.; Joglekar, S.S.

    1989-01-01

    This paper describes the development of an expert system for fault analysis of electronic instruments in the CIRUS nuclear reactor. The system was developed in Prolog on an IBM PC-XT compatible computer. A 'model-based' approach (Button et al, 1986) was adopted combining 'frames' and 'rules' to provide flexible control over the inferencing mechanisms. Frames represent the domain-objects as well as the inter-object relationships. They include 'demons' or 'active values' for triggering actions. Rules, along with frames, are used for fault analysis. The rules can be activated either in a data-driven or a goal-driven manner. The use of frames makes rule management easier. It is felt that developing in-house shell proved advantageous, compared to using commercially available shells. Choosing the model-based approach was efficient compared to a production system architecture. Therefore, the use of hybrid representations for diagnostic applications is advocated. Based on the experience, some general recommendations for developing such systems are presented. The expert system helps novice operators to understand the process of diagnosis and achieve a significant required level of competence. The system may not achieve the required level of proficiency by itself, but it can be used to train operators to become experts. (author). 12 refs

  8. System study application to the safety analysis of the exhaust and the tritium systems of a fusion reactor

    International Nuclear Information System (INIS)

    Djerassi, H.; Rouillard, J.; Leger, D.; Zappellini, G.; Gambi, G.

    1988-01-01

    An applicative example of the general methodology system study to the safety analysis of the exhaust and tritium systems, in a tokamak device, is shown. The framework of the study is split into the following tasks: initial information collection, functional analysis, failure scenarios identification and description, reliability data assessment, accident sequence quantification, consequence seriousness evaluation, risk assessment. Results concerning risk contribution from direct failures show that, in the exhaust system and in the tritium system, the risk contribution to public is rather smaller than the safety design targets. Nevertheless, if the reactor building is not taken into account, the risk contribution from the exhaust system can be significant. Risk contribution to the workers seems to be not to heavy

  9. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    International Nuclear Information System (INIS)

    Glasstone, S.; Sesonske, A.

    1994-01-01

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

  10. 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

  11. Hydraulic analysis of emergency core cooling system of reactor RP-10

    International Nuclear Information System (INIS)

    Gallardo Padilla, Alberto; Moreyra, Geraldo Lazaro; Nieto Malpartida, Manuel

    2002-01-01

    For design of the Emergency Core Cooling System (ECCS) of reactor RP-10 from Peru is very important the hydraulic analysis of this system. In this paper, based on a basic design of the ECCS are showed the conservation equations, the parabolic movement, being deduced from them the equations to evaluate regarding the time the variables to consider in the design: level of the emergency water in the reserve tank, flow, reaches of sprinkle, etc. In this analysis is considered a quasi-stationary flow for simplify the calculation. The developed model was implemented in a computer program denominated ECCSRP10, in language Fortran 77, whose results are shown in form graph. From analysis of results we can conclude that for the system of pipe of the ECCS the appropriate diameter is of 2 , and that the maximum flow possible to give is of 5 m 3 /h for to assure a minimum time of refrigeration of 150000 seconds. Experimental tests were made in a prototype of the pipe system being demonstrated that the obtained results of the simplified calculation agree with the values registered with a global approach of 10%. (author)

  12. Nuclear Reactor Engineering Analysis Laboratory

    International Nuclear Information System (INIS)

    Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

    1998-01-01

    The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels

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

    International Nuclear Information System (INIS)

    Liu Zhaohui; Yang Xiaohua; Liao Longtao; Wu Zhiqiang

    2015-01-01

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

  14. Expert systems for the analysis of transients on nuclear reactors: crisis analysis, sextant, a general purpose physical analyser

    International Nuclear Information System (INIS)

    Barbet, N.; Dumas, M.; Mihelich, G.; Souchet, Y.; Thomas, J.B.

    1987-04-01

    Two developments of expert systems intended to work on line to the analysis of nuclear reactor transients are reported. During an hypothetical crisis occurring in a nuclear facility, a staff of the Institute for Protection and Nuclear Safety (IPSN) has to assess the risk to local population. The expert system is intended to work as an assistant to the staff. At the present time, it deals with the availability of the safety systems of the plant (e.g. ECCS), depending on the functional state of the support systems. A next step is to take into account the physical transient of the reactor (mass and energy balance, pressure, flows). In order to reach this goal as in the development of other similar expert systems, a physical analyser is required. This is the aim of SEXTANT, which combines several knowledge bases concerning measurements, models and qualitative behaviour of the plant with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. Both expert systems require powerful shells for their development. SPIRAL is such a toolkit for the development of expert systems devoted to the computer aided management of complex processes

  15. Nuclear reactor refueling system

    International Nuclear Information System (INIS)

    Wade, E.E.

    1978-01-01

    A system for transferring fuel assemblies between a nuclear reactor core and a fuel storage area while the fuel assembies remain completely submerged in a continuous body of coolant is described. The system comprises an in-vessel fuel transfer machine located inside the reactor vessel and an ex-vessel fuel transfer machine located in a fuel storage tank. The in-vessel fuel transfer machine comprises two independently rotatable frames with a pivotable fuel transfer apparatus disposed on the lower rotatable frame. The ex-vessel fuel transfer machine comprises one frame with a pivotable fuel transfer apparatus disposed thereon. The pivotable apparatuses are capable of being aligned with each other to transfer a fuel assembly between the reactor vessel and fuel storage tank while the fuel assembly remains completely submerged in a continuous body of coolant. 9 claims, 7 figures

  16. Application of data analysis techniques to nuclear reactor systems code to accuracy assessment

    International Nuclear Information System (INIS)

    Kunz, R.F.; Kasmala, G.F.; Murray, C.J.; Mahaffy, J.H.

    2000-01-01

    An automated code assessment program (ACAP) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems (NRS) code results and experimental measurements. This software was developed under subcontract to the United States Nuclear Regulatory Commission for use in its NRS code consolidation efforts. In this paper, background on the topic of NRS accuracy and uncertainty assessment is provided which motivates the development of and defines basic software requirements for ACAP. A survey of data analysis techniques was performed, focusing on the applicability of methods in the construction of NRS code-data comparison measures. The results of this review process, which further defined the scope, user interface and process for using ACAP are also summarized. A description of the software package and several sample applications to NRS data sets are provided. Its functionality and ability to provide objective accuracy assessment figures are demonstrated. (author)

  17. Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

    International Nuclear Information System (INIS)

    Kim, Jong Hyun; Jin, Dong Sik; Chang, Soon Heung

    2013-01-01

    Highlights: • Developed new safety analysis methodology of moderator system failures for CANDU-6. • The new methodology used the TH-physics coupling concept. • Thermalhydraulic code is CATHENA, physics code is RFSP-IST. • Moderator system failure ends to the subcriticality through self-shutdown. -- Abstract: The new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module – CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively

  18. Design and development of fast pneumatic transfer system (PTS) for instrumental neutron activation analysis at Jordan research reactor

    International Nuclear Information System (INIS)

    Chung, Yongsam; Kim, Sunha; Moon, Jonghwa; Choi, Jinbok; Lee, Jongmin; Ryu, Jungsu

    2013-01-01

    A pneumatic transfer system (PTS) is one of the important equipment used for an neutron irradiation of a target material for an instrumental neutron activation analysis (INAA) in a research reactor. In particular, a rapid pneumatic transportation of irradiation capsule is essential for an accurate measurement of a short half-life nuclide. Three types of PTS for NAA facility at the Jordan Research and Training Reactor (JRTR) were newly developed for a functional improvement involving a manual and an automatic system which is equipped with programmable logic controller, software, and 13 devices to facilitate optimal operation of the system. In this paper, the designs and construction of these PTS, the operation and control of the system are described. In addition, a functional and operational test of the system were carried out as one of the basic requirement and characteristic parameters, and the results were reported to provide a user information as well as for the management and safety of the reactor

  19. Implementation of New Reactivity Measurement System and New Reactor Noise Analysis Equipment in a VVER-440 Nuclear Power Plant

    Science.gov (United States)

    Vegh, János; Kiss, Sándor; Lipcsei, Sándor; Horvath, Csaba; Pos, István; Kiss, Gábor

    2010-10-01

    The paper deals with two recently developed, high-precision nuclear measurement systems installed at the VVER-440 units of the Hungarian Paks NPP. Both developments were motivated by the reactor power increase to 108%, and by the planned plant service time extension. The first part describes the RMR start-up reactivity measurement system with advanced services. High-precision picoampere meters were installed at each reactor unit and measured ionization chamber current signals are handled by a portable computer providing data acquisition and online reactivity calculation service. Detailed offline evaluation and analysis of reactor start-up measurements can be performed on the portable unit, too. The second part of the paper describes a new reactor noise diagnostics system using state-of-the-art data acquisition hardware and signal processing methods. Details of the new reactor noise measurement evaluation software are also outlined. Noise diagnostics at Paks NPP is a standard tool for core anomaly detection and for long-term noise trend monitoring. Regular application of these systems is illustrated by real plant data, e.g., results of standard reactivity measurements during a reactor startup session are given. Noise applications are also illustrated by real plant measurements; results of core anomaly detection are presented.

  20. Thermodynamic exergy analysis for small modular reactor in nuclear hybrid energy system - 15110

    International Nuclear Information System (INIS)

    Boldon, L.; Liu, L.; Sabharwall, P.; Rabiti, C.; Bragg-Sitton, S.M.

    2015-01-01

    To assess the inherent value of energy in a thermal system, it is necessary to understand both the quantity and quality of energy available or the exergy. We study the case where nuclear energy through a small modular reactor (SMR) is supplementing the available wind energy through storage to meet the needs of the electrical grid. Nuclear power is also being used for the production of hydrogen via high temperature steam electrolysis. For a SMR exergy analysis, both the physical and economic environments must be considered. The physical environment incorporates the energy, raw materials, and reference environment, where the reference environment refers to natural resources available without limit and without cost. This paper aims to explore the use of exergy analysis methods to estimate and optimize SMR resources and costs for individual subsystems, based on thermodynamic principles-resource utilization and efficiency. The paper will present background information on exergy theory; identify the core subsystems in an SMR plant coupled with storage systems in support of renewable energy and hydrogen production; perform a thermodynamic exergy analysis; determine the cost allocation among these subsystems; and calculate unit 'exergetic' costs, unit 'exergo-economic' costs, and first and second law efficiencies. Exergetic and 'exergo-economic' costs ultimately determine how individual subsystems contribute to overall profitability and how efficiencies and consumption may be optimized to improve profitability, making SMRs more competitive with other generation technologies

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  2. Real-time analysis and display of reactor system mass inventory

    International Nuclear Information System (INIS)

    Dao, L.T.; Meachum, T.R.

    1982-01-01

    A mass inventory system (MIS) to evaluate, in real-time, the coolant distribution within the primary coolant system of the Loss-of-Fluid Test (LOFT) reactor has been developed. The computer-based system calculates and displays the coolant levels by two methods: using level measurements and performing a mass balance. The MIS is designed to provide up-to-date, intelligible information on the coolant distribution during any LOFT experiment. During LOFT experiments in which the primary coolant pumps are on, the method also provides void fraction information and the anticipated liquid level in the reactor vessel should the pumps be turned off

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

    International Nuclear Information System (INIS)

    Evrard, J.M.; Souchet, Y.

    1988-01-01

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

  4. A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; Lingerfelt, Eric J.; Wojtowicz, Anna

    2015-01-01

    Highlights: • Data analysis for high-performance simulations of reactors will be a problem that we address with a new management system. • We describe new input-output libraries for nuclear reactor simulations. • We describe a new user interface for visualizing and analyzing simulation results. • We show the utility of these systems with a 17 × 17 fuel assembly example simulation. • The availability of the code and avenues for collaboration are presented. - Abstract: Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system’s requirements and design

  5. Development of electronic system for reactivity measurement and reactor noise analysis

    International Nuclear Information System (INIS)

    Strohl, Claude Emile; Soares, Adalberto Jose

    1996-01-01

    In nuclear power reactors, the neutron detector signal is dependent of the number of fissions and the reactor power level. The detector signal can be divided into two components; a DC component, proportional to the average value and an AC component, which is the fluctuating part superimposed to the DC component. The analysis of the fluctuating part of the signal is called noise analysis and allow us to investigate phenomena occurring within the reactor vessel, such as vibration of fuels elements and coolant density, temperature, pressure and flow changes. On the other hand, the measure of the static DC part allow us to measure the local power density. This paper describes the development of a personal computer based signal conditioning card that, together with a personal computer commercial data acquisition card, can be used for noise analysis measurements and reactivity measurements of signals coming from ionization chambers or SPD's. (author)

  6. Nuclear reactor trip system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

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

  7. Neutronics shielding analysis of the last mirror-beam duct system for a laser fusion power reactor

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.; Klein, A.C.

    1981-01-01

    A Monte Carlo three-dimensional neutronics analysis for the last mirror-beam duct system for the SOLASE conceptual laser-driven fusion power reactor design is presented. Detailed geometric configurations including the reactor cavity, the two last mirrors, and the three-section two-right-angle bends duct are modeled. Measurements are given of the dimensions and compositions of the reactor components, and of neutron scalar fluxes, spatial dependencies and neutron volumetric heating rates for the cases of aluminum or Boral as laser beam duct liners, and ordinary concrete or lead mortar as shield material. A three-dimensional modeling of laser-driven reactor penetrations is employed. The particle leakage is found to be excessively high for the configuration of the conceptual design considered and the advantages and disadvantages of various solutions, such as the use of Boral as a duct liner and the use of lead mortar instead of ordinary concrete as a shield material, are considered

  8. Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

    2017-06-26

    Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

  9. Application of safety checklist to the analysis of the IEA-R1 reactor water retreatment system

    International Nuclear Information System (INIS)

    Sauer, Maria Eugenia Lago Jacques; Sara Neto, Antonio Jorge; Lima, Toni Carlos Caboclo de; Ribeiro, Maria Alice Morato

    2005-01-01

    In 1999, the management of the IEA-R1 Research Reactor (pool type - 5 MWth), located at IPEN/CNEN-SP, started the evaluation of the Reactor Pool Water Retreatment System to identify operational aspects, which could compromise the operators safety. The purpose was to identify and propose enhancements to the system which would be installed to substitute for the existing one. This process was conducted through a qualitative study of the system in operation. This study was carried out by a team composed of specialists in reactor operation, systems maintenance and radiological protection, and one safety analyst. The study consisted, basically, in local inspections to verify the physical and operational conditions of each equipment / component as well as aspects related to maintenance activities of the system. The process control and the operator procedures associated with the retreatment of the reactor pool water were also reviewed. The methodology adopted to develop the study was based in process hazard analysis technique named Safety Checklist. This paper presents a summary of this study and the main results obtained. Some operational and safety problems identified, the prevention and/or correction means to avoid them, and the recommendations and suggestions that have been implemented to the new design of the IEA-R1 Reactor Water Retreatment System, whose installation was concluded in 2003, are also presented. (author)

  10. Power Reactor Information System (PRIS)

    International Nuclear Information System (INIS)

    Spiegelberg, R.

    1992-01-01

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

  11. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  12. Reactor system safety assurance

    International Nuclear Information System (INIS)

    Mattson, R.J.

    1984-01-01

    The philosophy of reactor safety is that design should follow established and conservative engineering practices, there should be safety margins in all modes of plant operation, special systems should be provided for accidents, and safety systems should have redundant components. This philosophy provides ''defense in depth.'' Additionally, the safety of nuclear power plants relies on ''safety systems'' to assure acceptable response to design basis events. Operating experience has shown the need to study plant response to more frequent upset conditions and to account for the influence of operators and non-safety systems on overall performance. Defense in depth is being supplemented by risk and reliability assessment

  13. BOLD/VENTURE-4, Reactor Analysis System with Sensitivity and Burnup

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of program or function: The system of codes can be used to solve nuclear reactor core static neutronics and reactor history exposure problems. BOLD/VENTURE-4: First order perturbation and time-dependent sensitivity theories can be applied. Control rod positioning may be modeled explicitly and refueling treated with repositioning and recycle. Special capability is coded to model the continuously fueled core and to solve the importance and dominant harmonics problems. The modules of the code system are: VENTNEUT: VENTURE neutronics module; DRIVER and CONTRL: Control module; BURNER: Exposure calculation for reactor core analysis; FILEDTOR: File editor; INPROSER: Input processor; EXPOSURE: BURNER code module; REACRATE: Reaction rate calculation; CNTRODPO: Control rod positioning; FUELMANG: Fuel management positioning and accounting; PERTUBAT: Perturbation reactivity importance analyses; sensitivity analysis; DEPTHMOD: Static and time-dependent perturbation sensitivity analysis. The special processors are: DVENTR: Handles the input to the VENTURE module; DCMACR: Converts CITATION macroscopic cross sections to microscopic cross sections; DCRSPR: Produces input for the CROSPROS module; DUTLIN: Adds or replaces problem input data without exiting the program; DENMAN: Repositions fuel; DMISLY: Miscellaneous tasks. Standard interface files between modules are binary sequential files that follow a standardized format. VENTURE-PC: The microcomputer version is a subset of the mainframe version. The modules and special processors which are not part of VENTURE-PC are: REACRATE, CNTRODPO, PERTUBAT, FUELMANG, DEPTHMOD, DMISLY. 2 - method of solution: BOLD-VENTURE-4: The neutronics problems are solved by applying the multigroup diffusion theory representation of neutron transport applying an over-relaxation inner iteration, outer iteration scheme. Special modeling is used or source correction is done during iteration to solve importance and harmonics problems. No

  14. Modelling for great breaks accident analysis in the primary system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Serrano, M.A.B.; Vanni, E.A.

    1981-01-01

    An analysis is made for a break in the cold leg, of the guillotine type with discharge coefficient C sub(D)=1.0, for the Angra 1 reactor. The computer codes, geometrical models and options used are described. A comparison between the method used and the requirements in the Appendix K of 10 CRF 50 is done. (Author) [pt

  15. Reliability analysis of 2400 MWth gas-cooled fast reactor natural circulation decay heat removal system

    International Nuclear Information System (INIS)

    Marques, M.; Bassi, C.; Bentivoglio, F.

    2012-01-01

    In support to a PSA (Probability Safety Assessment) performed at the design level on the 2400 MWth Gas-cooled Fast Reactor, the functional reliability of the decay heat removal system (DHR) working in natural circulation has been estimated in two transient situations corresponding to an 'aggravated' Loss of Flow Accident (LOFA) and a Loss of Coolant Accident (LOCA). The reliability analysis was based on the RMPS methodology. Reliability and global sensitivity analyses use uncertainty propagation by Monte Carlo techniques. The DHR system consists of 1) 3 dedicated DHR loops: the choice of 3 loops (3*100% redundancy) is made in assuming that one could be lost due to the accident initiating event (break for example) and that another one must be supposed unavailable (single failure criterion); 2) a metallic guard containment enclosing the primary system (referred as close containment), not pressurized in normal operation, having a free volume such as the fast primary helium expansion gives an equilibrium pressure of 1.0 MPa, in the first part of the transient (few hours). Each dedicated DHR loop designed to work in forced circulation with blowers or in natural circulation, is composed of 1) a primary loop (cross-duct connected to the core vessel), with a driving height of 10 meters between core and DHX mid-plan; 2) a secondary circuit filled with pressurized water at 1.0 MPa (driving height of 5 meters for natural circulation DHR); 3) a ternary pool, initially at 50 C. degrees, whose volume is determined to handle one day heat extraction (after this time delay, additional measures are foreseen to fill up the pool). The results obtained on the reliability of the DHR system and on the most important input parameters are very different from one scenario to the other showing the necessity for the PSA to perform specific reliability analysis of the passive system for each considered scenario. The analysis shows that the DHR system working in natural circulation is

  16. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco

    2016-01-01

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  17. Thermal-hydraulic analysis of an innovative decay heat removal system for lead-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Giannetti, Fabio; Vitale Di Maio, Damiano; Naviglio, Antonio; Caruso, Gianfranco, E-mail: gianfranco.caruso@uniroma1.it

    2016-08-15

    Highlights: • LOOP thermal-hydraulic transient analysis for lead-cooled fast reactors. • Passive decay heat removal system concept to avoid lead freezing. • Solution developed for the diversification of the decay heat removal functions. • RELAP5 vs. RELAP5-3D comparison for lead applications. - Abstract: Improvement of safety requirements in GEN IV reactors needs more reliable safety systems, among which the decay heat removal system (DHR) is one of the most important. Complying with the diversification criteria and based on pure passive and very reliable components, an additional DHR for the ALFRED reactor (Advanced Lead Fast Reactor European Demonstrator) has been proposed and its thermal-hydraulic performances are analyzed. It consists in a coupling of two innovative subsystems: the radiative-based direct heat exchanger (DHX), and the pool heat exchanger (PHX). Preliminary thermal-hydraulic analyses, by using RELAP5 and RELAP5-3D© computer programs, have been carried out showing that the whole system can safely operate, in natural circulation, for a long term. Sensitivity analyses for: the emissivity of the DHX surfaces, the PHX water heat transfer coefficient (HTC) and the lead HTC have been carried out. In addition, the effects of the density variation uncertainty on the results has been analyzed and compared. It allowed to assess the feasibility of the system and to evaluate the acceptable range of the studied parameters. A comparison of the results obtained with RELAP5 and RELAP5-3D© has been carried out and the analysis of the differences of the two codes for lead is presented. The features of the innovative DHR allow to match the decay heat removal performance with the trend of the reactor decay heat power after shutdown, minimizing at the same time the risk of lead freezing. This system, proposed for the diversification of the DHR in the LFRs, could be applicable in the other pool-type liquid metal fast reactors.

  18. Analysis and primary design of the I and C system architecture for HTGR-10 reactor

    International Nuclear Information System (INIS)

    Yang Zijue; Zhao Guoji

    1993-01-01

    The consideration of making good use of the-state-of-the-arts technology in designing advanced Instrumentation and Control System architecture is discussed. A fully distributed and fully micro-computerized, local network based Instrumentation and Control System is designed for the HTR-10 reactor. The advantages of the system architecture include high reliability and availability, flexibility, economics, etc. It also fits for other production processes

  19. Analysis of water hammer in control rod drive systems of boiling water reactor nuclear power plants

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Lau, S.

    1983-01-01

    The method of characteristics is applied to analyze water hammer in BWR (Boiling Water Reactor) Control Rod Drive (CRD) Systems following fast opening of scram valves. The modelling of the CRD mechanism is presented. Numerical predictions are compared to experimental data. (author)

  20. BWR reactor management system

    International Nuclear Information System (INIS)

    Makino, Kakuji; Kawamura, Atsuo; Yoshioka, Ritsuo; Neda, Toshikatsu.

    1979-01-01

    It is necessary to grasp the delicate state of operation in reactor cores in view of the control of burn-up and power output at the time of the operation management of BWRs. Enormous labor has been required for the collection, processing and evaluation of the data. It is desirable to obtain the safer, more efficient and faster method of operation control by predicting the states in cores including the change of xenon and reflecting them to operation plans as well as by tracing with high accuracy the past burn-up history for a long period. At present, the on-line evaluation of the states in cores is carried out with the process computers attached to respective units, but the amount of data required for core operation management of high degree far exceeds their capacity. From such viewpoints, the research and development on the reactor management system were carried out. The data processing concerning core operation management is performed with newly installed computers utilizing the data from existing process computers, and the operation of reactor cores, the qualitative improvement of management works, labor saving, and fast, efficient operation control are feasible with it. This system was installed in an actual plant in October, 1977. The composition of the system, the prediction of the change in local output distribution accompanying control rod operation, the prediction of the change in the states in cores due to the flow rate of coolant, and the function of collecting plant data are explained. (Kako, I.)

  1. Nuclear reactor sealing system

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1983-01-01

    A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

  2. Systems Analysis of a Fast Steam-Cooled Reactor of 1000 MW(E)

    Energy Technology Data Exchange (ETDEWEB)

    Smidt, D.; Frisch, W.; Hofmann, F.; Moers, H.; Schramm, K.; Spilker, H. [Institut fuer Reaktorentwicklung, Kernforschungszentrum, Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany); Kiefhaber, E. [Institut fuer Neutronenphysik und Reaktortechnik Kernforschungszentrum Karlsruhe, Karlsruhe, Federal Republic of Germany (Germany)

    1968-05-15

    The Karlsruhe design of a steam-cooled fast reactor (Dl) has been the subject of a systems analysis. Here the dependence of fuel inventory, breeding ratio, rating, core geometry and plant efficiency on coolant pressure, and coolant temperature has been studied for two different rod powers. The effect of artificial surface roughness has been investigated. For some configurations the resulting fuel-cycle and capital costs have been determined and discussed. The main influence results from pressure. The lower pressure allows for higher breeding ratios, but lower efficiencies and vice versa. From this the fuel-cycle costs show an optimum at around 150 atm abs. The capital costs on the other side decrease with pressure. The over-all optimum of the power generating costs for the presently studied parameter range is at about 170 atm abs., a coolant outlet temperature of 540 Degree-Sign C and a rod power of 420 W/cm. Artificial roughness (boundary layer type) leads for a required system pressure and outlet temperature to a larger coolant volume fraction and, therefore, to reduced breeding ratios but higher efficiencies. As another part of the work some stability characteristics of the cores were studied. The dependence of the core stability on the varied parameters is shown. (author)

  3. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    EL Khatib, H.H.A.

    2013-01-01

    The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

  4. 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

  5. Thermionic nuclear reactor systems

    International Nuclear Information System (INIS)

    Kennel, E.B.

    1986-01-01

    Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

  6. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  7. TRAC-BD1: transient reactor analysis code for boiling-water systems

    International Nuclear Information System (INIS)

    Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

    1981-01-01

    The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

  8. Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

  9. 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

  10. Thermodynamic exergy analysis for small modular reactor in nuclear hybrid energy system

    Directory of Open Access Journals (Sweden)

    Boldon Lauren

    2016-01-01

    Full Text Available Small modular reactors (SMRs provide a unique opportunity for future nuclear development with reduced financial risks, allowing the United States to meet growing energy demands through safe, reliable, clean air electricity generation while reducing greenhouse gas emissions and the reliance on unstable fossil fuel prices. A nuclear power plant is comprised of several complex subsystems which utilize materials from other subsystems and their surroundings. The economic utility of resources, or thermoeconomics, is extremely difficult to analyze, particularly when trying to optimize resources and costs among individual subsystems and determine prices for products. Economics and thermodynamics cannot provide this information individually. Thermoeconomics, however, provides a method of coupling the quality of energy available based on exergy and the value of this available energy – “exergetic costs”. For an SMR exergy analysis, both the physical and economic environments must be considered. The physical environment incorporates the energy, raw materials, and reference environment, where the reference environment refers to natural resources available without limit and without cost, such as air input to a boiler. The economic environment includes market influences and prices in addition to installation, operation, and maintenance costs required for production to occur. The exergetic cost or the required exergy for production may be determined by analyzing the physical environment alone. However, to optimize the system economics, this environment must be coupled with the economic environment. A balance exists between enhancing systems to improve efficiency and optimizing costs. Prior research into SMR thermodynamics has not detailed methods on improving exergetic costs for an SMR coupled with storage technologies and renewable energy such as wind or solar in a hybrid energy system. This process requires balancing technological efficiencies and

  11. 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.

  12. Simplified Models for Analysis and Design of the Control System Main Loops of CAREM Reactor

    International Nuclear Information System (INIS)

    Etchepareborda, Andres; Flury, Celso

    2000-01-01

    The target of this work is to show a few models developed for control analysis and design of the reactor CAREM's main control loops within a broad range of power (between 40 % and 100%).By one side, it is shown the main features of a analytic model programed in MATLAB.This model is based on fitting steady state points at different power levels of the CAREM's RETRAN model.By the other side, it is shown linear models of black-box type denoting the perturbed behavior of the system for each level power point.These models are identified from temporal responses of CAREM's RETRAN model to perturbed input signals over the different steady power level points.Then the dynamics of these models are verified contrasting the temporal responses of the RETRAN model versus the responses of the MATLAB model and the identified models, in each steady power level point.Also are contrasting the frequency response of the linearization of MATLAB model versus the frequency response of the identified models, in each steady power level point.Either the MATLAB model as the identified models are good enough for the control analysis and design of the three main control loops.The MATLAB model has a few differences against the RETRAN model in the primary pressure output variable, that it must be taken into account in the design of this control loop if this model is used.The aim of these models is to represent in a satisfactory way the dynamics of the plant for a later control analysis and design of the control loops in a frequency range between 0.01 rad/seg and 0.3 rad/seg, and a power range between 40 % and 100 %

  13. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ballinger, Ronald G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Wang, Chun Yun [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kadak, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todreas, Neil [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mirick, Bradley [Concepts, Northern Engineering and Research, Woburn, MA (United States); Demetri, Eli [Concepts, Northern Engineering and Research, Woburn, MA (United States); Koronowski, Martin [Concepts, Northern Engineering and Research, Woburn, MA (United States)

    2004-08-30

    power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

  14. Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems

    International Nuclear Information System (INIS)

    Ballinger, Ronald G.; Chunyun Wang; Kadak, Andrew; Todreas, Neil

    2004-01-01

    power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%

  15. Reactor noise analysis applications and systems in WWER-440 and WWER-1000 type PWRs

    International Nuclear Information System (INIS)

    Por, G.

    1998-01-01

    This paper presents an introduction on different types of well selected noise diagnostic methods with their occurrence in WWER reactors with an analysis of their impact on operational safety and aging which affects the installations safety as well. The main objective is to attract the attention of NPP management staff dealing with safety, safety culture, maintenance, operation and quality assurance proving that such methods can give benefit not only to economy but impact safety of nuclear installations

  16. Probabilistic safety assessment of Tehran Research Reactor using systems analysis programs for hands-on integrated reliability evaluations

    International Nuclear Information System (INIS)

    Hosseini, M.H.; Nematollahi, M.R.; Sepanloo, K.

    2004-01-01

    Probabilistic safety assessment application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this document the application of the probabilistic safety assessment to the Tehran Research Reactor is presented. The level 1 practicabilities safety assessment application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantifications, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using systems analysis programs for hands-on integrated reliability evaluations software

  17. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    International Nuclear Information System (INIS)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K.; Kim, J. H.

    2015-01-01

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed

  18. Operational characteristics analysis of a 8 mH class HTS DC reactor for an LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. K.; Go, B. S.; Dinh, M. C.; Park, M.; Yu, I. K. [Changwon National University, Changwon (Korea, Republic of); Kim, J. H. [Daejeon University, Daejeon (Korea, Republic of)

    2015-03-15

    Many kinds of high temperature superconducting (HTS) devices are being developed due to its several advantages. In particular, the advantages of HTS devices are maximized under the DC condition. A line commutated converter (LCC) type high voltage direct current (HVDC) transmission system requires large capacity of DC reactors to protect the converters from faults. However, conventional DC reactor made of copper causes a lot of electrical losses. Thus, it is being attempted to apply the HTS DC reactor to an HVDC transmission system. The authors have developed a 8 mH class HTS DC reactor and a model-sized LCC type HVDC system. The HTS DC reactor was operated to analyze its operational characteristics in connection with the HVDC system. The voltage at both ends of the HTS DC reactor was measured to investigate the stability of the reactor. The voltages and currents at the AC and DC side of the system were measured to confirm the influence of the HTS DC reactor on the system. Two 5 mH copper DC reactors were connected to the HVDC system and investigated to compare the operational characteristics. In this paper, the operational characteristics of the HVDC system with the HTS DC reactor according to firing angle are described. The voltage and current characteristics of the system according to the types of DC reactors and harmonic characteristics are analyzed. Through the results, the applicability of an HTS DC reactor in an HVDC system is confirmed.

  19. Reactor safety systems

    International Nuclear Information System (INIS)

    Kafka, P.

    1975-01-01

    The spectrum of possible accidents may become characterized by the 'maximum credible accident', which will/will not happen. Similary, the performance of safety systems in a multitude of situations is sometimes simplified to 'the emergency system will/will not work' or even 'reactors are/ are not safe'. In assessing safety, one must avoid this fallacy of reducing a complicated situation to the simple black-and-white picture of yes/no. Similarly, there is a natural tendency continually to improve the safety of a system to assure that it is 'safe enough'. Any system can be made safer and there is usually some additional cost. It is important to balance the increased safety against the increased costs. (orig.) [de

  20. Report and analysis on 'PR and PP evaluation. Example sodium fast reactor full system case study'

    International Nuclear Information System (INIS)

    Sagara, Hiroshi; Inoue, Naoko; Kawakubo, Yoko; Watahiki, Masaru

    2011-01-01

    The Generation IV (GEN IV) Nuclear Energy Systems International Forum (GIF) Proliferation Resistance and Physical Protection Working Group (PRPP WG) was established in December 2002 in order to develop the PR and valuation methodology for GEN IV nuclear energy systems. In the final report of 'PR and PP Evaluation: Example Sodium Fast Reactor (ESFR) Full System Case Study,' issued in October 2009, the demonstration study of PR and PP evaluation with the qualitative approach are summarized using ESFR with four scenario threats. The present paper reviews and analyzes some results of the ESFR case study, and identifies the challenges and direction for the PR and PP evaluation methodology with quantitative approach. (author)

  1. Reactor cooling system

    International Nuclear Information System (INIS)

    Kato, Etsuji.

    1979-01-01

    Purpose: To eliminate cleaning steps in the pipelines upon reactor shut-down by connecting a filtrating and desalting device to the cooling system to thereby always clean up the water in the pipelines. Constitution: A filtrating and desalting device is connected to the pipelines in the cooling system by way of drain valves and a check valve. Desalted water is taken out from the exit of the filtrating and desalting device and injected to one end of the cooling system pipelines by way of the drain valve and the check valve and then returned by way of another drain valve to the desalting device. Water in the pipelines is thus always desalted and the cleaning step in the pipelines is no more required in the shut-down. (Kawakami, Y.)

  2. Scaling analysis for a Savannah River reactor scaled model integral system

    International Nuclear Information System (INIS)

    Boucher, T.J.; Larson, T.K.; McCreery, G.E.; Anderson, J.L.

    1990-11-01

    801The Savannah River Laboratory has requested that the Idaho National Engineering Laboratory perform an analysis to help define, examine, and assess potential concepts for the design of a scaled integral hydraulics test facility representative of the current Savannah River Plant reactor design. In this report the thermal-hydraulic phenomena of importance (based on the knowledge and experience of the authors and the results of the joint INEL/TPG/SRL phenomena identification and ranking effort) to reactor safety during the design basis loss-of-coolant accident were examined and identified. Established scaling methodologies were used to develop potential concepts for integral hydraulic testing facilities. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally support that a one-fourth (1/4) linear scale visual facility capable of operating at pressures up to 350 kPa (51 psia) and temperatures up to 330 K (134 degree F) will scale most hydraulic phenomena reasonably well. However, additional research will be necessary to determine the most appropriate method of simulating several of the reactor components, since the scaling methodology allows for several approaches which may only be assessed via appropriate research. 34 refs., 20 figs., 14 tabs

  3. Dynamic modeling of primary and secondary systems of IRIS reactor for transient analysis using SIMULINK

    International Nuclear Information System (INIS)

    Magalhaes, Mardson Alencar de Sa; Lira, Carlos Alberto Brayner de Oliveira; Silva, Mario Augusto Bezerra da

    2011-01-01

    The IRIS project has significantly advanced in the last few years in response to a demand for a new generation reactor, that could fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is a integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes in relation to a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics for power production. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were into the SIMULINK and the code was validated by the confrontation with RELAP simulations for a transient of feedwater reduction in the steam generators. (author)

  4. CFD and system analysis code investigations of the multidimensional flow mixing phenomena in the reactor pressure vessel

    International Nuclear Information System (INIS)

    Ceuca, S.C.; Herb, J.; Schoeffel, P.J.; Hollands, T.; Austregesilo, H.; Hristov, H.V.

    2017-01-01

    The realistic numerical prediction of transient fluid-dynamic scenarios including the complex, three-dimensional flow mixing phenomena occurring in the reactor pressure vessel (RPV) both in normal or abnormal operation are an important issue in today's reactor safety assessment studies. Both Computational Fluid Dynamics (CFD) tools as well as fluid-dynamic system analysis codes, each with its advantages and drawbacks, are commonly used to model such transients. Simulation results obtained with the open-source CFD tool-box OpenFOAM and the German thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients), the later developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design basis accidents anticipated for nuclear energy facilities, are compared against experimental data from the ROssendorf Coolant Mixing (ROCOM) test facility. In the case of the OpenFOAM CFD simulations the influence of various turbulence models and numerical schemes has been assessed while in the case of the system analysis code ATHLET a multidimensional nodalization recommended for real power plant applications has been employed. The simulation results show a good agreement with the experimental data, indicating that both OpenFOAM and ATHLET can capture the key flow features of the mixing processes in the Reactor Pressure Vessel (RPV). (author)

  5. Reactor system on barge

    International Nuclear Information System (INIS)

    Hayashi, Kingo; Yamada, Nobuyuki

    1987-01-01

    Floating electrical power plants or power plant barges add new dimensions to utility planners and agencies in the world. Intrinsically safe and economical reactors (ISER) employ steel reactor pressure vessels, which significantly reduce the weight as compared with PIUS, and provide siting versatility including barge-mounted plants. In this paper, the outline of power plant barges and barge-mounted ISERs is described. Besides their mobility, power plant barges have the salient advantages such as short delivery time and better quality control due to the outfitting in shipyards. These power plant barges may be temporarily moored or permanently grounded in shallow water at the centers of industrial complexes or the suitable areas adjacent to them, and satisfy the increasing needs for electric power. A cost-effective and technically perfect barge positioning system should be designed to meet the specific requirement for the location and its condition. Offshore siting away from coast may be applicable only to large plants of 1,000 MWe or more, and inshore siting and coastal or river siting are considered for an ISER-200 barge-mounted plant. The system of a barge-mounted ISER plant is discussed in the case of a floating type and the type on a seismic base isolator. (Kako, I.)

  6. Nuclear reactor safety system

    International Nuclear Information System (INIS)

    Sato, Takashi.

    1979-01-01

    Purpose: To allow sufficient removal of radioactive substance released in the reactor containment shell upon loss of coolants accidents thus to sufficiently decrease the exposure dose to human body. Constitution: A clean-up system is provided downstream of a heat exchanger and it is branched into a pipeway to be connected to a spray nozzle and further connected by way of a valve to a reactor container. After the end of sudden transient changes upon loss of coolants accidents, the pool water stored in the pressure suppression chamber is purified in the clean-up system and then sprayed in the dry-well by way of a spray nozzle. The sprayed water dissolves to remove water soluble radioactive substances floating in the dry-well and then returns to the pressure suppression chamber. Since radioactive substances in the dry-well can thus removed rapidly and effectively and the pool water can be reused, public hazard can also be decreased. (Horiuchi, T.)

  7. Thermophysical properties of thorium and uranium systems for use in reactor safety analysis

    International Nuclear Information System (INIS)

    Fink, J.K.; Chasanov, M.G.; Leibowitz, L.

    1977-06-01

    The data compilation is intended to serve as a preliminary set of thermophysical property values for use in reactor safety analyses of the Th-- 233 U reactor concept. The properties covered include mp, bp, enthalpy, heats of vaporization and fusion, heat capacity, thermal conductivity, density, thermal expansion, emissivity, viscosity, etc. The systems covered are Th, Th 0 . 9 U 0 . 1 , U, ThO 2 , Th 0 . 9 U 0 . 1 O 2 , UO 2 , U 0 . 8 Pu 0 . 2 O 2 , ThC, Th 0 . 9 U 0 . 1 C, UC, U 0 . 8 Pu 0 . 2 C, ThC 2 , Th 0 . 9 U 0 . 1 C 2 , and UC 2 . 5 figures, 10 tables, 150 references

  8. TSUNAMI analysis of the applicability of proposed experiments to reactor-grade and weapons-grade mixed-oxide systems

    International Nuclear Information System (INIS)

    Rearden, Bradley T.; Hopper, Calvin M.; Elam, Karla R.

    2005-01-01

    The applicability of proposed critical experiments for the criticality code validation of a series of prototypic reactor-grade and weapons-grade mixed-oxide systems has been assessed with the TSUNAMI methodology from SCALE 5. The application systems were proposed by the Nuclear Energy Agency (NEA) Organization for Economic Cooperation and Development (OECD) Working Party on Nuclear Criticality Safety MOX Experimental Needs Working Group. Forty-eight application systems were conceived to envelope the range of conditions in processing and fabrication of reactor-grade and weapons-grade MOX fuel. The applicability of 303 existing critical benchmarks to each of the 48 applications was assessed, and validation coverage was found to be lacking for certain applications. Two series of proposed critical experiments were also considered in this analysis. The TSUNAMI analysis has revealed that both series of proposed experiments are applicable to numerous configurations of the reactor-grade and weapons-grade systems. A detailed assessment of which experiments were revealed by TSUNAMI to be most applicable to specific prototypic fuel processing systems has been performed. (author)

  9. Probabilistic Analysis of Passive Safety System Reliability in Advanced Small Modular Reactors: Methodologies and Lessons Learned

    Energy Technology Data Exchange (ETDEWEB)

    Grabaskas, David; Bucknor, Matthew; Brunett, Acacia; Grelle, Austin

    2015-06-28

    Many advanced small modular reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize with a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper describes the most promising options: mechanistic techniques, which share qualities with conventional probabilistic methods, and simulation-based techniques, which explicitly account for time-dependent processes. The primary intention of this paper is to describe the strengths and weaknesses of each methodology and highlight the lessons learned while applying the two techniques while providing high-level results. This includes the global benefits and deficiencies of the methods and practical problems encountered during the implementation of each technique.

  10. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  11. Development of automatic reactor vessel inspection systems; development of data acquisition and analysis system for the nuclear vessel weld

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Po; Park, C. H.; Kim, H. T.; Noh, H. C.; Lee, J. M.; Kim, C. K.; Um, B. G. [Research Institute of KAITEC, Seoul (Korea)

    2002-03-01

    The objective of this project is to develop an automated ultrasonic data acquisition and data analysis system to examine heavy vessel welds. In order to examine nuclear vessel welds including reactor pressure vessel(RPV), huge amount of ultrasonic data from 6 channels should be able to be on-line processed. In addition, ultrasonic transducer scanning device should be remotely controlled, because working place is high radiation area. This kind of an automated ultrasonic testing equipment has not been developed domestically yet. In order to develop an automated ultrasonic testing system, RPV ultrasonic testing equipments developed in foreign countries were investigated and the capability of high speed ultrasonic signal processing hardwares was analyzed. In this study, ultrasonic signal processing system was designed. And also, ultrasonic data acquisition software was developed. The new systems were tested on the RPV welds of Ulchin Unit 6 to confirm their functions and capabilities. They worked very well as designed and the tests were successfully completed. 13 refs., 34 figs., 11 tabs. (Author)

  12. Comparative Neutronics Analysis of DIMPLE S06 Criticality Benchmark with Contemporary Reactor Core Analysis Computer Code Systems

    Directory of Open Access Journals (Sweden)

    Wonkyeong Kim

    2015-01-01

    Full Text Available A high-leakage core has been known to be a challenging problem not only for a two-step homogenization approach but also for a direct heterogeneous approach. In this paper the DIMPLE S06 core, which is a small high-leakage core, has been analyzed by a direct heterogeneous modeling approach and by a two-step homogenization modeling approach, using contemporary code systems developed for reactor core analysis. The focus of this work is a comprehensive comparative analysis of the conventional approaches and codes with a small core design, DIMPLE S06 critical experiment. The calculation procedure for the two approaches is explicitly presented in this paper. Comprehensive comparative analysis is performed by neutronics parameters: multiplication factor and assembly power distribution. Comparison of two-group homogenized cross sections from each lattice physics codes shows that the generated transport cross section has significant difference according to the transport approximation to treat anisotropic scattering effect. The necessity of the ADF to correct the discontinuity at the assembly interfaces is clearly presented by the flux distributions and the result of two-step approach. Finally, the two approaches show consistent results for all codes, while the comparison with the reference generated by MCNP shows significant error except for another Monte Carlo code, SERPENT2.

  13. Development of a coupled containment-reactor coolant system methodology for the analysis of IRIS small break LOCA

    International Nuclear Information System (INIS)

    Manfredini, Antonio; Oriolo, Francesco; Paci, Sandro; Oriani, Luca

    2003-01-01

    The main purpose of the present work is to identify the most relevant physical phenomena for the IRIS (International Reactor Innovative and Secure) containment system and the development of an integrated methodology for the simultaneous safety analysis of both the reactor and containment with available computer codes. Specific objectives are: (a) to assess the limitations of the lumped parameter codes on predictions of complex situations; (b) to identify alternatives to classical containment analysis techniques. The characteristic features of an integral reactor like IRIS present a much greater challenge to code developers than conventional, loop type PWRs. In particular, the integral primary system and the containment are strongly coupled during postulated accident conditions and thus an integrated simulation of both systems is required to obtain a reliable phenomenological representation. The comparison of the results obtained in the application of two containment codes (GOTHIC and integrated FUMO) on 'ad hoc' IRIS related benchmarks will also be described. These preliminary calculations were used to test the IRIS containment concept and cooling strategies, at the same time highlighting the most relevant issues that require a more refined investigation. Finally, this activity allowed to perform more refined calculations, in progress at the moment, aimed at showing that the IRIS safety systems and containment design solutions perform their intended functions. (author)

  14. ENFORM II: a calculational system for light water reactor logistics and effluent analysis

    International Nuclear Information System (INIS)

    Heeb, C.M.; Lewallen, M.A.; Purcell, W.L.; Cole, B.M.

    1979-09-01

    ENFORM is a computer-based information system that addresses the material logistics, environmental releases and economics of light water reactor (LWR) operation. The most important system inputs consist of electric energy generation requirements, details of plant construction scheduling, unit costs, and environmental release factors. From these inputs the ENFORM system computes the mass balances and generates the environmental release information for noxious chemicals and radionuclides from various fuel cycle facilities (except waste disposal). Fuel cycle costs and electric power costs are also computed. All code development subsequent to 1977 is summarized. Programming instructions are provided for the modules that are comprised in the ENFORM system. ENGEN, a code that uses a generation schedule specified by the user and isotopic data generated by ORIGEN, has been developed to produce a scenario-specific data base. Other codes (ENMAT, ENRAD, etc) have been developed to use data base information to estimate radioactive and nonradioactive release information

  15. Proceedings of workshop on reactor shutdown system

    International Nuclear Information System (INIS)

    1997-03-01

    India has gained considerable experience in design, development, construction and operation of research and power reactors during the last four decades. Reactor shutdown system (RSS) is the most important engineered safety system of any reactor. A lot of technological developments have taken place to improve the reactor shutdown systems, particularly with advancement in reliability analysis and instrumentation and control. If the reactor is not shutdown, the fuel may melt, releasing radioactivity and possibly reactivity addition as in the case of Fast Breeder Reactor (FBR). Apart from radiological safety consequences, large investment has to be written off. The function of the RSS is to stop fission chain reaction and prevent breach of fuel. The design of RSS is multidisciplinary. It requires reactor physics analysis, design of absorber rods, drive mechanisms, safety logic to order shutdown and instrumentation to detect unsafe conditions. High reliability is essential and this requires two independent shutdown systems. This book contains the proceedings of the workshop on reactor shutdown system and papers relevant to INIS are indexed separately

  16. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  17. Development of the GO-FLOW reliability analysis methodology for nuclear reactor system

    International Nuclear Information System (INIS)

    Matsuoka, Takeshi; Kobayashi, Michiyuki

    1994-01-01

    Probabilistic Safety Assessment (PSA) is important in the safety analysis of technological systems and processes, such as, nuclear plants, chemical and petroleum facilities, aerospace systems. Event trees and fault trees are the basic analytical tools that have been most frequently used for PSAs. Several system analysis methods can be used in addition to, or in support of, the event- and fault-tree analysis. The need for more advanced methods of system reliability analysis has grown with the increased complexity of engineered systems. The Ship Research Institute has been developing a new reliability analysis methodology, GO-FLOW, which is a success-oriented system analysis technique, and is capable of evaluating a large system with complex operational sequences. The research has been supported by the special research fund for Nuclear Technology, Science and Technology Agency, from 1989 to 1994. This paper describes the concept of the Probabilistic Safety Assessment (PSA), an overview of various system analysis techniques, an overview of the GO-FLOW methodology, the GO-FLOW analysis support system, procedure of treating a phased mission problem, a function of common cause failure analysis, a function of uncertainty analysis, a function of common cause failure analysis with uncertainty, and printing out system of the results of GO-FLOW analysis in the form of figure or table. Above functions are explained by analyzing sample systems, such as PWR AFWS, BWR ECCS. In the appendices, the structure of the GO-FLOW analysis programs and the meaning of the main variables defined in the GO-FLOW programs are described. The GO-FLOW methodology is a valuable and useful tool for system reliability analysis, and has a wide range of applications. With the development of the total system of the GO-FLOW, this methodology has became a powerful tool in a living PSA. (author) 54 refs

  18. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  19. Monitor for reactor feedwater systems

    International Nuclear Information System (INIS)

    Takizawa, Yoji; Tomizawa, Teruaki

    1983-01-01

    Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

  20. System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

    International Nuclear Information System (INIS)

    Harvego, E.A.; McKellar, M.G.; Sohal, M.S.; O'Brien, J.E.; Herring, J.S.

    2010-01-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540 C and 900 C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

  1. Distributed expert systems for nuclear reactor control

    International Nuclear Information System (INIS)

    Otaduy, P.J.

    1992-01-01

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

  2. MHD equilibrium methods for ITER [International Thermonuclear Experimental Reactor] PF [poloidal field] coil design and systems analysis

    International Nuclear Information System (INIS)

    Strickler, D.J.; Galambos, J.D.; Peng, Y.K.M.

    1989-03-01

    Two versions of the Fusion Engineering Design Center (FEDC) free-boundary equilibrium code designed to computer the poloidal field (PF) coil current distribution of elongated, magnetically limited tokamak plasmas are demonstrated and applied to the systems analysis of the impact of plasma elongation on the design point of the International Thermonuclear Experimental Reactor (ITER). These notes were presented at the ITER Specialists' Meeting on the PF Coil System and Operational Scenario, held at the Max Planck Institute for Plasma Physics in Garching, Federal Republic of Germany, May 24--27, 1988. 8 refs., 6 figs., 4 tabs

  3. Hydraulic analysis of emergency core cooling system of reactor RP-10; Analisis hidraulico del sistema de refrigeracion de emergencia del nucleo del reactor RP-10

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo Padilla, Alberto; Moreyra, Geraldo Lazaro; Nieto Malpartida, Manuel [Instituto Peruano de Energia Nuclear, Lima (Peru)]. E-mail: agallardo@ipen.gob.pe; glazaro@ipen.gob.pe; mnieto@ipen.gob.pe

    2002-07-01

    For design of the Emergency Core Cooling System (ECCS) of reactor RP-10 from Peru is very important the hydraulic analysis of this system. In this paper, based on a basic design of the ECCS are showed the conservation equations, the parabolic movement, being deduced from them the equations to evaluate regarding the time the variables to consider in the design: level of the emergency water in the reserve tank, flow, reaches of sprinkle, etc. In this analysis is considered a quasi-stationary flow for simplify the calculation. The developed model was implemented in a computer program denominated ECCSRP10, in language Fortran 77, whose results are shown in form graph. From analysis of results we can conclude that for the system of pipe of the ECCS the appropriate diameter is of 2 in., and that the maximum flow possible to give is of 5 m{sup 3}/h for to assure a minimum time of refrigeration of 150000 seconds. Experimental tests were made in a prototype of the pipe system being demonstrated that the obtained results of the simplified calculation agree with the values registered with a global approach of 10%. (author)

  4. Reliability analysis of the service water system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Tayt-Sohn, L.C.; Oliveira, L.F.S. de.

    1984-01-01

    A reliability analysis of the service water system is done aiming to use in the evaluation of the non reliability of the Component Cooling System (SRC) for great loss of cooling accidents in nuclear power plants. (E.G.) [pt

  5. Reliability analysis of the service water system of Angra 1 reactor

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Fleming, P.V.; Frutuoso e Melo, P.F.F.; Tayt-Sohn, L.C.

    1983-01-01

    A reliability analysis of the service water system is done aiming to use in the evaluation of the non reliability of the component cooling system (SRC) for great loss of cooling accidents in nuclear power plants. (E.G.) [pt

  6. VENTURE-PC 1.1, Reactor Analysis System with Sensitivity and Burnup

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: The VENTURE program solves the usual neutronics eigenvalue, adjoint, fixed source, and criticality search problems. It treats up to three dimensions, maps power density, and does first-order perturbation analysis at the macroscopic cross section level. The BURNER code solves the nuclide chain equations to estimate the nuclide concentrations and burnup at the end of an exposure time or after a shutdown period. This package is based on the CCC-459/BOLD VENTURE IV code system developed at Oak Ridge National Laboratory. In January 1989 the University of Cincinnati contributed the first VENTURE-PC package to RSICC's collection. It was a subset of the mainframe version consisting of the VENTURE and BURNER modules plus several processing modules. VENTURE-PC was distributed as CCC-459 until July 1997 when a new version (with updated source code compatible with newer FORTRAN-77 compilers, some revisions, and extensions to solve much larger problems) was contributed by Argonne National Laboratory. The principle code modules included in the VENTURE-PC system are: VENTURE: Multigroup neutronics finite-difference diffusion theory. BURNER: Depletion calculation for reactor core analysis. Other modules within VENTURE-PC are: DVENTR: Venture input processor; DCRSPR: Neutron cross section processor; DUTLIN: Control file (CNTRL) input processor; DCMACR: Citation format cross section input processor; CRXSPR: Cross section processor; DENMAN: Fuel repositioning module. In August of 1999, Argonne again contributed an updated version of the code which overcomes problem size constraints caused by binary record length limits inherent to the Fortran 90 compiler. The need for long records is detected and avoided by sub-blocking them. Also, the latest Fortran 95 compiler offers substantial speed gains on the newest processors. The source code is updated to be compatible with either Fortran 90 or Fortran 95. In August 2002, the package was updated with

  7. Complex risk analysis for loss of electric power in liquid metal nuclear reactor by system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2012-07-15

    The power stabilization of the nuclear power plants (NPPs) is investigated in the aspect of the liquid metal coolant. The quantification of the risk analysis is performed by the system dynamics (SD) method which is processed by the feedback and accumulation complex algorithms. The Vensim software package is used for the simulations, which is supported by the Monte-Carlo method. There are 2 kinds of considerations as the economic and safety properties. The result shows the stability of the operations when the power can be decided. This shows the higher efficiency of the reactor. The failure frequency is 16/60 = 27%. In the event of Power Stabilized, the failure event is in the quite lower frequency rate. The commercial use of the reactor is important in the operations. (orig.)

  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. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  10. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

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

  11. Reactor applications of quantitative diffraction analysis

    International Nuclear Information System (INIS)

    Feguson, I.F.

    1976-09-01

    Current work in quantitative diffraction analysis was presented under the main headings of: thermal systems, fast reactor systems, SGHWR applications and irradiation damage. Preliminary results are included on a comparison of various new instrumental methods of boron analysis as well as preliminary new results on Zircaloy corrosion, and materials transfer in liquid sodium. (author)

  12. Modeling, simulation, and analysis of a reactor system for the generation of white liquor of a pulp and paper industry

    Directory of Open Access Journals (Sweden)

    Ricardo Andreola

    2011-02-01

    Full Text Available An industrial system for the production of white liquor of a pulp and paper industry, Klabin Paraná Papéis, formed by ten reactors was modeled, simulated, and analyzed. The developed model considered possible water losses by the evaporation and reaction, in addition to variations in the volumetric flow of lime mud across the reactors due to the composition variations. The model predictions agreed well with the process measurements at the plant and the results showed that the slaking reaction was nearly complete at the third causticizing reactor, while causticizing ends by the seventh reactor. Water loss due to slaking reaction and evaporation occurred more pronouncedly in the slaker reactor than in the final causticizing reactors; nevertheless, the lime mud flow remained nearly constant across the reactors.

  13. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    International Nuclear Information System (INIS)

    Gallardo, J.; Marquino, W.; Mistreanu, A.; Yang, J.

    2015-09-01

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  14. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

    2015-09-15

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  15. Conceptual design analysis of an MHD power conversion system for droplet-vapor core reactors. Final report

    International Nuclear Information System (INIS)

    Anghaie, S.; Saraph, G.

    1995-01-01

    A nuclear driven magnetohydrodynamic (MHD) generator system is proposed for the space nuclear applications of few hundreds of megawatts. The MHD generator is coupled to a vapor-droplet core reactor that delivers partially ionized fissioning plasma at temperatures in range of 3,000 to 4,000 K. A detailed MHD model is developed to analyze the basic electrodynamics phenomena and to perform the design analysis of the nuclear driven MHD generator. An incompressible quasi one dimensional model is also developed to perform parametric analyses

  16. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  17. BWR type reactor system

    International Nuclear Information System (INIS)

    Morooka, Shin-ichi.

    1980-01-01

    Purpose: To reduce the internal structure in a reactor by rapidly and efficiently transferring heat generated in a reactor core out of the reactor and eliminating the danger of radiation exposure. Constitution: Steam generated in a pressure vessel is introduced into heat pipe group by inserting the heat pipe group into the steam dome of the pressure vessel. The introduced steam is condensed in the heat pipes to transfer the heat of the steam to the heat pipe group. The transferred heat is transmitted to a heat exchanger provided out of a containment vessel to generate steam to operate a turbine. Thus, it is not necessary to introduce the steam including radioactive substance externally and can remove only the heat so as to carry out the desired purpose. (Kamimura, M.)

  18. Development of long-lived radionuclide transmutation technology - Development of a code system for core analysis of the transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Zin; Kim, Yong Hee; Kim, Tae Hyung; Jo, Chang Keun; Park, Chang Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    The objective of this study is to develop a code system for core analysis= of the critical transmutation reactors utilizing fast neutrons. Core characteristics of the transmutation reactors were identified and four codes, HANCELL for pincell calculation, PRISM and AFEN-H3D for core calculation, and MA{sub B}URN for depletion calculation, were developed. The pincell calculation code is based on one-dimensional collision probability method and may provide homogenized/condensed parameters of a pincell and also can homogenize the control assembly via a nonlinear iterative method. The core calculation codes, PRISM and AFEN-H3D, solve the multi-group, multi-dimensional neutron diffusion equations for a hexagonal geometry and they are based on the finite difference method and analytic function expansion nodal (AFEN) method, respectively. The MA{sub B}URN code san analyze the behavior of actinides and fission products in a reactor core. Through benchmarking, we confirmed that the newly developed codes provide accurate solutions. 30 refs., 10 tabs., 8 figs. (author)

  19. Distributed computing and nuclear reactor analysis

    International Nuclear Information System (INIS)

    Brown, F.B.; Derstine, K.L.; Blomquist, R.N.

    1994-01-01

    Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations

  20. Reliability analysis of Diesel Generator power supply system of Prototype Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Pramod Kumar, E-mail: pramodks@igcar.gov.in; Bhuvana, V.; Ramakrishnan, M.

    2016-12-15

    Highlights: • The unavailability of DG success is 4.75E−3 for 2/4 and 1.47E−3 for 1/4. • Modeling includes sub systems like CB, SSWS, Fuel oil system & 220 V DC. • DG-FR, DG-FR-CCF and DG maintenance is major contributors of DG unavailability. • Uncertainty analysis has been carried out through Monte Carlo simulations. • Sensitivity analysis identifies DG mechanical FR as most sensitive part. - Abstract: The unavailability of Diesel Generator power supply system has been evaluated using Fault tree method with ISOGRAPH reliability software and is found to be 4.75E−3 for 2/4 (DG success) and 1.47E−3 for 1/4 (DG success). Common cause failures contribute significantly to the unavailability of the system. Statistical analysis indicates that the DG unavailability is uncertain by Error Factor 4.4 (90% confidence bound) for 2 out of 4 DG system (system success) and by Error Factor 4.1 (90% confidence bound) for 1 out of 4 DG system (system success). Support systems namely Safety related service water system, Fuel oil system and circuit breaker control power supply dependency have been modeled. Results of importance analysis and sensitivity study are used to identify significant contributors to unavailability. DG fails to run, DG fails to run due to CCF and DG maintenance out of service is identified as dominant and important contributors of DG unavailability. Uncertainty analysis has been carried out through Monte Carlo simulations.

  1. Reliability analysis of air recirculation and refrigeration systems of Angra-1 reactor contaiment: a reevaluation

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Fleming, P.V.; Frutuoso e Melo, P.F.F.

    1983-01-01

    A realiability analysis of the air refrigeration and recirculation containment systems (ARRCS) of Angra-1 nuclear power plants, were done, aiming to evaluate the probabilities of occurence of a several accident. The systems were analysed for a 24 hours accident, including time failures and demand failures [pt

  2. Isolation systems influence in the seismic loading propagation analysis applied to an innovative near term reactor

    International Nuclear Information System (INIS)

    Lo Frano, R.; Forasassi, G.

    2010-01-01

    Integrity of a Nuclear Power Plant (NPP) must be ensured during the plant life in any design condition and, particularly, in the event of a severe earthquake. To investigate the seismic resistance capability of as-built structures systems and components, in the event of a Safe Shutdown Earthquake (SSE), and analyse its related effects on a near term deployment reactor and its internals, a deterministic methodological approach, based on the evaluation of the propagation of seismic waves along the structure, was applied considering, also, the use of innovative anti-seismic techniques. In this paper the attention is focused on the use and influence of seismic isolation technologies (e.g. isolators based on passive energy dissipation) that seem able to ensure the full integrity and operability of NPP structures, to enhance the seismic safety (improving the design of new NPPs and if possible, to retrofit existing facilities) and to attain a standardization plant design. To the purpose of this study a numerical assessment of dynamic response/behaviour of the structures was accomplished by means of the finite element approach and setting up, as accurately as possible, a representative three-dimensional model of mentioned NPP structures. The obtained results in terms of response spectra (carried out from both cases of isolated and not isolated seismic analyses) are herein presented and compared in order to highlight the isolation technique effectiveness.

  3. Development of an automatic reactor inspection system

    International Nuclear Information System (INIS)

    Kim, Jae Hee; Eom, Heung Seop; Lee, Jae Cheol; Choi, Yoo Raek; Moon, Soon Seung

    2002-02-01

    Using recent technologies on a mobile robot computer science, we developed an automatic inspection system for weld lines of the reactor vessel. The ultrasonic inspection of the reactor pressure vessel is currently performed by commercialized robot manipulators. Since, however, the conventional fixed type robot manipulator is very huge, heavy and expensive, it needs long inspection time and is hard to handle and maintain. In order to resolve these problems, we developed a new automatic inspection system using a small mobile robot crawling on the vertical wall of the reactor vessel. According to our conceptual design, we developed the reactor inspection system including an underwater inspection robot, a laser position control subsystem, an ultrasonic data acquisition/analysis subsystem and a main control subsystem. We successfully carried out underwater experiments on the reactor vessel mockup, and real reactor ready for Ulchine nuclear power plant unit 6 at Dusan Heavy Industry in Korea. After this project, we have a plan to commercialize our inspection system. Using this system, we can expect much reduction of the inspection time, performance enhancement, automatic management of inspection history, etc. In the economic point of view, we can also expect import substitution more than 4 million dollars. The established essential technologies for intelligent control and automation are expected to be synthetically applied to the automation of similar systems in nuclear power plants

  4. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

  5. Human failure event analysis and precautionary methods and their application to reactor system

    International Nuclear Information System (INIS)

    Zhang Li; Huang Shudong; Wang Yiqun; Gao Wenyu; Zhang Jin

    2003-01-01

    Making use of human factor engineering, control science and safety science and adopting the method of systemically collection and doing research work factually, the authors analyze the problem and tendency of human factor science, the classification system, the formation, the quantitative appraisal, data collection and data bank, the effect and influence of organization management, the root cause analysis technology, and human error failure mode and effect and criticality analysis, the method and strategy of defense-in-depth for preventing human-initiated accident. The human factor accidents theory and mechanism are constructed. All of the above was successfully applied to Daya Bay Nuclear Power Station and Lingao Nuclear Power Station. (authors)

  6. Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

    International Nuclear Information System (INIS)

    Kalchev, B.

    1995-01-01

    The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E -3 , 1/RY to 1.07E -3 , 1/RY. 5 refs., 2 tabs., 5 figs

  7. Reliability analysis of the reconstructed safety systems of the Kozloduy-2 WWER-440/V-230 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalchev, B [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The Unit 2 of the Kozloduy NPP started operations in 1975. As it is designed according to safety standards of the middle sixties, it needs reconstruction in order to prolong its operational life up to the design age of 30 years, in agreement with the increased safety requirements in Bulgaria. The reliability analyses of front line systems of the unit are performed to this end. The approach taken in the study is the fault tree methodology to determine the unavailability of each system. Common mode failures are considered for the pumps and valves using the beta factor method. The mission time for each system is 24 hours and the test period is 720 hours. Support systems and human errors are also included. All the systems control and instrumentation signals are modelled explicitly in the fault trees. The generic IDEA reliability data base is used for all quantifications. The initiating events that would require the system operation are presented and on this basis the thermohydraulic analysis success criteria for each system are determined. The code for probabilistic safety assessment PSAPACK is used. Fault trees for the following front line safety systems are constructed: the high pressure injection system, the spray system and the auxiliary feed water system. The analysis consider some proposed decisions for reconstruction. The results show that the reliability of these systems has increased after reconstruction and the safety has been upgraded. This decrease the core damage frequency from 3.53E{sup -3}, 1/RY to 1.07E{sup -3}, 1/RY. 5 refs., 2 tabs., 5 figs.

  8. Laser fusion power reactor system (LFPRS)

    International Nuclear Information System (INIS)

    Kovacik, W.P.

    1977-01-01

    This report gives detailed information for each of the following areas: (1) reference concept description, (2) nuclear design, (3) structural design, (4) thermal and fluid systems design, (5) materials design and analysis, (6) reactor support systems and balance of plant, (7) instrumentation and control, (8) environment and safety, (9) economics assessment, and (10) development requirements

  9. Economic analysis of EBT reactor

    International Nuclear Information System (INIS)

    Woo, J.T.; Uckan, N.A.; Lidsky, L.M.

    1977-01-01

    In order to establish the economic potential of the Elmo Bumpy Torus (EBT) reactor, two independent system-costing models have been developed. Both models predict capital costs of approximately $400/kW(th). These relatively low costs reflect the simplicity of the EBTR design. In particular, the modular nature of the individual blanket-shield segments, the low costs ''accelerator style'' containment building, high beta, and steady-state operation lead to relatively low reactor costs. A detailed cost breakdown for subsystems is analyzed. High cost and high uncertainty subsystems are identified to direct further emphasis into those areas. The calculated capital costs for the EBT reactor are compared with those costs quoted for tokamak reactors

  10. Safety system for reactor container

    International Nuclear Information System (INIS)

    Shimizu, Miwako; Seki, Osamu; Mano, Takio.

    1995-01-01

    A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

  11. Study on development of virtual reactor core laboratory (1). Development of prototype coupled neutronic, thermal-hydraulic and structural analysis system

    International Nuclear Information System (INIS)

    Uto, Nariaki; Sugaya, Toshio; Tsukimori, Kazuyuki; Negishi, Hitoshi; Enuma, Yasuhiro; Sakai, Takaaki

    1999-09-01

    A study on development of virtual reactor core laboratory, which is to conduct numerical experiments representative of complicated physical phenomena in practical reactor core systems on a computational environment, has progressed at Japan Nuclear Cycle Development Institute (JNC). The study aims at systematic evaluation of these phenomena into which nuclear reactions, thermal-hydraulic characteristics, structural responses and fuel behaviors combine, and effective utilization of the obtained comprehension for core design. This report presents a production of a prototype computational system which is required to construct the virtual reactor core laboratory. This system is to evaluate reactor core performance under the coupled neutronic, thermal-hydraulic and structural phenomena, and is composed of two analysis tools connected by a newly developed interface program; 1) an existing space-dependent coupled neutronic and thermal-hydraulic analysis system arranged at JNC and 2) a core deformation analysis code. It acts on a cluster of several DEC/Alpha workstations. A specific library called MPI1 (Message Passing Interface 1) is incorporated as a tool for communicating among the analysis modules consisting of the system. A series of calculations for simulating a sequence of Unprotected Loss Of Heat Sink (ULOHS) coupled with rapid drop of some neutron absorber devices in a prototype fast reactor is tried to investigate how the system works. The obtained results show the core deformation behavior followed by the reactivity change that can be properly evaluated. The results of this report show that the system is expected to be useful for analyzing sensitivity of reactor core performance with respect to uncertainties of various design parameters and establishing a concept of passive safety reactor system, taking into account space distortion of neutron flux distribution during abnormal events as well as reactivity feedback from core deformation. (author)

  12. A simplified dynamic analysis for reactor piping systems under blowdown conditions

    International Nuclear Information System (INIS)

    Chen, M.M.

    1975-01-01

    In the design of pipelines in a nuclear power plant for blowdown conditions, is it customary to conduct dynamic analysis of the piping system to obtain the responses and the resulting stresses. Calculations are repeated for each design modification in piping geometry or supporting system until the design codes are met. The numerical calculations are, in general, very costly and time consuming. Until now, there have been no simple means for calculating the dynamic responses for the design. The proposed method reduces the dynamic calculation to a quasi-static one, and can be beneficially used for the preliminary design. The method is followed by a complete dynamical analysis to improve the final results. The new formulations greatly simplify the numerical computation and provide design guides. When used to design a given piping system, the method saved approximately one order of magnitude of computer time. The approach can also be used for other types of structures

  13. LOCA analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    Bajs, T.; Grgic, D.; Cavlina, N.

    2003-01-01

    The IRIS reactor (International Reactor Innovative and Secure) is an integral, light water cooled, medium power reactor. IRIS has been selected as an International Near Term Deployable (INTD) reactor, within the Generation IV International Forum activities. The IRIS concept addresses the key-requirements defined by the US DOE for next generation reactors, i.e. enhanced reliability and safety, and improved economics. It features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). An innovative safety approach has been developed to mitigate the IRIS response to small-to-medium Loss of Coolant Accident (LOCA). This strategy is based on the interaction of IRIS compact containment with the reactor vessel to limit initial blowdown, and on depressurization through the use of a passive Emergency Heat Removal System (EHRS). A small Automatic Depressurization System (ADS) provides supplementary depressurization capability. A pressure suppression system is provided to limit the pressure peak following the initial blowdown to well below the containment design limit. The ultimate result is that during a small-to-medium LOCA, the core remains covered for an extended period of time, without credit for emergency water injection or external core makeup. The IRIS LOCA response is based on 'maintaining water inventory' rather than on the principle of safety injection. This novel safety approach poses significant issues for computational and analysis methods since the IRIS vessel and containment are strongly coupled, and the system response is based on the interaction between the two. The small break LOCA was calculated using RELAP5/mod3.3 and GOTHIC codes. Break of the largest line connected to the IRIS Reactor Pressure Vessel (RPV) was analyzed. The results of the calculations confirmed good performance of the IRIS system during LOCA. (author)

  14. 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.)

  15. An advanced three-dimensional simulation system for safety analysis of gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lapins, Janis

    2016-07-01

    simulated simultaneously. TORT-TD and ATTICA{sup 3D} exchange data (power distributions or fuel and moderator temperature distribution, possibly hydrogen distribution) by means of a common interface that interpolates values that are exchanged on mutual computational grids by volumetric averaging. As verification for the proper operation of the interface, the steady state of the transient PBMR-400 benchmark was used. After obtaining a coupled steady state, the transient exercises are performed to test the proper working of the interface in time dependent cases. Here, the cold helium ingress, the total control rod withdrawal case and the total control rod ejection case were simulated and compared to results of other partakers of the benchmark. Also, the coupled system was validated for a full power temperature distribution experiment in the Chinese experimental reactor HTR-10 where good agreement could be reached with the measurements. The coupled HTR simulation system TORT-TD/ATTICA{sup 3D} was then applied for single control rod ejection cases for both the PBMR-400 and the HTR-PM. These cases require a 180 model of the reactor. As preparatory works, the control rod cross sections were adjusted to yield the same reactivity increase as the grey curtain model for the PBMR and with MCNP5 for the HTR-PM. Since there are strong shielding effects by neighbouring rods, the power increase was moderate due to strong Doppler and moderator feedbacks. For the HTR-PM, coupled calculations for water ingress cases are simulated. This also tested the whole computational sequence, i.e. steam transport into the core by ATTICA{sup 3D}, then transfer of hydrogen densities (from hydrogen or from steam) to TORT-TD via the interface, interpolation of the macroscopic cross sections which changes the power density, and the feedback to ATTICA{sup 3D}. Additionally, an anticipated transient without scram is simulated where shutdown of the reactor is achieved by the temperature feedback effects. For

  16. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  17. 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

  18. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  19. Probabilistic safety analysis of the containment spray system of Angra-1 reactor

    International Nuclear Information System (INIS)

    Gibelli, S.M.O.

    1981-02-01

    The calculation of the unavailability of the containment spray system of Angra-1, is done. The referred system has two different modes of operation (injection and recirculation) which were separately studied using the fault tree methodology. Besides equipment and human error failures, the contributions of test, maintenance and common-mode failures have also been considered. The quantitative evaluation was carried out by the computer code SAMPLE, which considers the uncertainties in the failures data and gives a distribution for the top event unavailability. The input data were obtained from the well-known Rasmussen Report. An importance analysis of the basic events of the trees was performed and a study of the viability of some suggestions for system design modification was also conducted. A comparison between the results obtained in this work and the corresponding ones in the Rasmussen Report has shown the fact that the unavailability of both systems are of the same order of magnitude. (Author) [pt

  20. Analysis of passive moderator cooling system of Candu-6A reactor at emergency condition

    International Nuclear Information System (INIS)

    Umar, Efrizon; Subki, M. Hadid; Vecchiarelli, Jack

    2001-01-01

    Analysis of passive moderator cooling system subject to in-core LOCA with no emergency core cooling injection has been done. In this study, the new model of passive moderator system has been tested for emergency conditions and CATHENA code Mod-3.5b/Rev1 is used to calculate some parameters of this passive moderator cooling system. This result of simulation show that the proposed moderator cooling system have given satisfactory result, especially for the case with 0.7 m riser diameter and the number of heat exchanger tubes 8100. For PEWS tank containing 3000 m3 of light water initially at 30 0C and a 3641 m2 moderator heat exchanger, the average long-term heat removed rate balances the moderator heat load and the flow through the passive moderator loop remains stable for over 72 hours with no saturated boiling in the calandria and flow instabilities do not develop during long-term period

  1. Reliability analysis of air recirculation and-refrigeration systems of Angra-1 reactor containment

    International Nuclear Information System (INIS)

    Fernandes Filho, T.L.

    1982-10-01

    A reliability analysis of the air refrigeration and recirculation containment systems (ARRCS) of Angra-1 nuclear power plants, were done. The fault tree analysis was used. The failure primary data were taken out of Wash-1400 and IEEE. These data were processed by these following computer codes : Prep-Kitt, Sample, Trebil, Cressex and Streusl for the two stages of ARRCS operation. The design bases accident studied was a LOCA (loss of coolant). The component that more contribution give to the non-availability of ARRCS is the motor of the ARRCS. (E.G.) [pt

  2. Probabilistic analysis of reactor safety - The auxiliary feedwater system of Angra I

    International Nuclear Information System (INIS)

    Oliveira, L.C.R. da L.C. de.

    1981-09-01

    The unavailability of the auxiliary feedwater system (AFWS) of Angra-1, was calculated. The fault tree analysis technique was used, considering two diferent types of contribution to system unavailability: The one due to hard-ware failure and the contribution due to test and maintenance which was separately analysed. The COMBO-and SAMPLE computer codes were used. The results have shown that the AFWS of Angra-1 contains enough redundancy to guarantee a safe operation under the conditions analysed, best values having been obtained for the unavailability of AFWS of Angra 1 with those codes than with the WASH-1400. (E.G.) [pt

  3. Safety Analysis Of Actinide Recycled Fast Power Reactor

    International Nuclear Information System (INIS)

    Taufik, Mohammad

    2001-01-01

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

  4. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  5. Power reactor information system (PRIS)

    International Nuclear Information System (INIS)

    1989-06-01

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

  6. Solvent refined coal reactor quench system

    Science.gov (United States)

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  7. A prototype expert system to support the development of a fault-tree analysis software for nuclear reactor safety

    International Nuclear Information System (INIS)

    Mesko, L.

    1990-01-01

    The project called EMERIS is designed to provide a material testing nuclear reactor and experimental loops with a software for the 'acquisition, evaluation and archivation of measured data during the operation of the experimental facility'. The project which gives job a team has a duration of two years and involves three Vax compatible TPA-type computers and many smaller computers for data digitalization and graphical workstations. The detailed description of the project is not the task of the paper. One of its modules, however, plays an important role in the considerations. Namely the module for distrubance analysis (DA) which is planned to perform a rule based on-line evaluation of numerous predefined fault trees in an expert system like environment

  8. CASTEM: a system of finite element computer programs for elastic and inelastic analysis of mechanical structures of reactors

    International Nuclear Information System (INIS)

    Hoffmann, A.; Livolant, M.; Roche, R.

    1978-01-01

    The nuclear research center at Saclay has developed the system of computer program CASTEM for the analysis of mechanical structures of reactors. This finite element system is designed specially to deal with nonlinear problems concerning both the material (plasticity, thermoplasticity, creep) and the geometry (nonlinear relationships between displacement and strain, buckling). Furthermore, a special effort has been devoted to the processing of dynamic problems (vibrations, natural modes, earthquakes, shock phenomena, etc..). The CASTEM system includes a large number of elementary modules corresponding to a total of over 80,000 Fortran instructions. Allowing the calculation of various structural geometries, including: axisymmetrical shells and liquids (with non axisymmetrical loading); pipes and frames; two-dimensional massive structures; three-dimensional shells; three-dimensional massive structures. Complex dynamic analysis can be made by combination of substructures natural mode shapes. Pre and post processors: automatic meshing, plotting of results, direct comparison of stresses to ASME limits make the use of the system easy and time saving

  9. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Pham Van Lam; Le Vinh Vinh; Huynh Ton Nghiem

    2004-01-01

    Requirements of neutron, thermohydraulic and safety analysis calculation are very important because of issuing new version of SAR for DNRR, research on construction of new research reactor and nuclear power plant. Research on application of system of neutron, thermohydraulic and safety analysis codes in order to simulation of the Dalat Nuclear Research Reactor has been done in the frame work of research theme in the year 2002-2003. The purposes of the research are maintaining safety operation of the DNRR and enhancement of man power and calculation and safety analysis tool potential. (author)

  10. Computer Aided Analysis and Prototype Testing of an Improved Biogas Reactor For Biomass System

    Directory of Open Access Journals (Sweden)

    Jeremy (Zheng Li

    2015-05-01

    Full Text Available The alternative fuel resources substituting for conventional fuels are required due to less availability of fuel resources than demand in the market. A large amount of crude oil and petroleum products are required to be imported in many countries over the world. Also the environmental pollution is another serious problem when use petroleum products. Biogas, with the composition of 54.5% CH4, 39.5% CO2, and 6% other elements (i.e., H2, N2, H2S, and O2, is a clear green fuel that can substitute the regular petroleum fuels to reduce the pollutant elements. Biogas can be produced by performing enriching, scrubbing, and bottling processes. The purification process can be further applied to take away the pollutants in biogas. The pure biogas process analyzed in this research is compressed to 2950 psi while being filled into gas cylinder. The daily produced biogas capacity is around 5480 ft3 and the processing efficacy is affected by surrounding environment and other factors. The design and development of this biogas system is assisted through mathematical analysis, 3D modeling, computational simulation, and prototype testing. Both computer aided analysis and prototype testing show close results which validate the feasibility of this biogas system in biomass applications.

  11. ORTAP: a nuclear steam supply system simulation for the dynamic analysis of high temperature gas cooled reactor transients

    International Nuclear Information System (INIS)

    Cleveland, J.C.; Hedrick, R.A.; Ball, S.J.; Delene, J.G.

    1977-01-01

    ORTAP was developed to predict the dynamic behavior of the high temperature gas cooled reactor (HTGR) Nuclear Steam Supply System for normal operational transients and postulated accident conditions. It was developed for the Nuclear Regulatory Commission (NRC) as an independent means of obtaining conservative predictions of the transient response of HTGRs over a wide range of conditions. The approach has been to build sufficient detail into the component models so that the coupling between the primary and secondary systems can be accurately represented and so that transients which cover a wide range of conditions can be simulated. System components which are modeled in ORTAP include the reactor core, a typical reheater and steam generator module, a typical helium circulator and circulator turbine and the turbine generator plant. The major plant control systems are also modeled. Normal operational transients which can be analyzed with ORTAP include reactor start-up and shutdown, normal and rapid load changes. Upset transients which can be analyzed with ORTAP include reactor trip, turbine trip and sudden reduction in feedwater flow. ORTAP has also been used to predict plant response to emergency or faulted conditions such as primary system depressurization, loss of primary coolant flow and uncontrolled removal of control poison from the reactor core

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

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

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

  13. Development of a system model for advanced small modular reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  14. Fault tree analysis of a research reactor

    International Nuclear Information System (INIS)

    Hall, J.A.; O'Dacre, D.F.; Chenier, R.J.; Arbique, G.M.

    1986-08-01

    Fault Tree Analysis Techniques have been used to assess the safety system of the ZED-2 Research Reactor at the Chalk River Nuclear Laboratories. This turned out to be a strong test of the techniques involved. The resulting fault tree was large and because of inter-links in the system structure the tree was not modularized. In addition, comprehensive documentation was required. After a brief overview of the reactor and the analysis, this paper concentrates on the computer tools that made the job work. Two types of tools were needed; text editing and forms management capability for large volumes of component and system data, and the fault tree codes themselves. The solutions (and failures) are discussed along with the tools we are already developing for the next analysis

  15. Guidelines for nuclear reactor equipments safety-analysis

    International Nuclear Information System (INIS)

    1978-01-01

    The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

  16. Sextant: an expert system for transient analysis of nuclear reactors and integral test facilities

    International Nuclear Information System (INIS)

    Barbet, N.; Dumas, M.; Mihelich, G.

    1987-01-01

    Expert systems provide a new way of dealing with the computer-aided management of nuclear plants by combining several knowledge bases and reasoning modes together with a set of numerical models for real-time analysis of transients. New development tools are required together with metaknowledge bases handling temporal hypothetical reasoning and planning. They have to be efficient and robust because during a transient, neither measurements nor models, nor scenarios are hold as absolute references. SEXTANT is a general purpose physical analyzer intended to provide a pattern and avoid duplication of general tools and knowledge bases for similar applications. It combines several knowledge bases concerning measurements, models and qualitative behavior of PWR with a mechanism of conjecture-refutation and a set of simplified models matching the current physical state. A prototype is under assessment by dealing with integral test facility transients. For its development, SEXTANT requires a powerful shell. SPIRAL is such a toolkit, oriented towards online analysis of complex processes and already used in several applications

  17. Innovations and enhancements in neutronic analysis of the Big-10 university research and training reactors based on the AGENT code system

    International Nuclear Information System (INIS)

    Hursin, M.; Shanjie, X.; Burns, A.; Hopkins, J.; Satvat, N.; Gert, G.; Tsoukalas, L. H.; Jevremovic, T.

    2006-01-01

    Introduction. This paper summarizes salient aspects of the 'virtual' reactor system developed at Purdue Univ. emphasizing efficient neutronic modeling through AGENT (Arbitrary Geometry Neutron Transport) a deterministic neutron transport code. DOE's Big-10 Innovations in Nuclear Infrastructure and Education (INIE) Consortium was launched in 2002 to enhance scholarship activities pertaining to university research and training reactors (URTRs). Existing and next generation URTRs are powerful campus tools for nuclear engineering as well as a number of disciplines that include, but are not limited to, medicine, biology, material science, and food science. Advancing new computational environments for the analysis and configuration of URTRs is an important Big-10 INIE aim. Specifically, Big-10 INIE has pursued development of a 'virtual' reactor, an advanced computational environment to serve as a platform on which to build operations, utilization (research and education), and systemic analysis of URTRs physics. The 'virtual' reactor computational system will integrate computational tools addressing the URTR core and near core physics (transport, dynamics, fuel management and fuel configuration); thermal-hydraulics; beam line, in-core and near-core experiments; instrumentation and controls; confinement/containment and security issues. Such integrated computational environment does not currently exist. The 'virtual' reactor is designed to allow researchers and educators to configure and analyze their systems to optimize experiments, fuel locations for flux shaping, as well as detector selection and configuration. (authors)

  18. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

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

  19. Tokamak reactor systems studies

    International Nuclear Information System (INIS)

    1992-01-01

    A summary of work completed on the ARIES project during this report period is given. The main areas of effort were: neutronics, shield optimization and design, safety, systems, startup and shutdown, and ripple loss

  20. Reactor limit control system

    International Nuclear Information System (INIS)

    Rubbel, F.E.

    1982-01-01

    The very extensive use of limitations in the operational field between protection system and closed-loop controls is an important feature of German understanding of operational safety. The design of limitations is based on very large activities in the computational field but mostly on the high level of the plant-wide own commissioning experience of a turnkey contractor. Limitations combine intelligence features of closed-loop controls with the high availability of protection systems. (orig.)

  1. Development of comprehensive and versatile framework for reactor analysis, MARBLE

    International Nuclear Information System (INIS)

    Yokoyama, Kenji; Hazama, Taira; Numata, Kazuyuki; Jin, Tomoyuki

    2014-01-01

    Highlights: • We have developed a neutronics code system for reactor analysis. • The new code system covers all five phases of the core design procedures. • All the functionalities are integrated and validated in the same framework. • The framework supports continuous improvement and extension. • We report results of validation and practical applications. - Abstract: A comprehensive and versatile reactor analysis code system, MARBLE, has been developed. MARBLE is designed as a software development framework for reactor analysis, which offers reusable and extendible functions and data models based on physical concepts, rather than a reactor analysis code system. From a viewpoint of the code system, it provides a set of functionalities utilized in a detailed reactor analysis scheme for fast criticality assemblies and power reactors, and nuclear data related uncertainty quantification such as cross-section adjustment. MARBLE includes five sub-systems named ECRIPSE, BIBLO, SCHEME, UNCERTAINTY and ORPHEUS, which are constructed of the shared functions and data models in the framework. By using these sub-systems, MARBLE covers all phases required in fast reactor core design prediction and improvement procedures, i.e. integral experiment database management, nuclear data processing, fast criticality assembly analysis, uncertainty quantification, and power reactor analysis. In the present paper, these functionalities are summarized and system validation results are described

  2. 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

  3. CCF analysis of BWR reactor shutdown systems based on the operating experience at the TVO I/II in 1981-1993

    International Nuclear Information System (INIS)

    Mankamo, T.

    1996-04-01

    The work constitutes a part of the project conducted within the research program of the Swedish Nuclear Power Inspectorate SKI, aimed to develop the methods and data base for the Common Cause Failure (CCF) analysis of highly redundant reactor scram systems. The data analysis for the TVO I/II plant is focused on the hydraulic scram system, and control rods and drives. It covers operating experiences from 1981 through 1993. (9 refs., 9 figs., 7 tabs.)

  4. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

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

    International Nuclear Information System (INIS)

    Yashima, Hideyuki

    1980-01-01

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

  6. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  7. 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.)

  8. Nuclear reactor power supply system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

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

  10. Power-balance analysis of muon-catalyzed fusion-fission hybrid reactor systems

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.

    1985-01-01

    A power-balance model of a muon-catalyzed fusion system in the context of a fission-fuel factory is developed and exercised to predict the required physics performance of systems competitive with either pure muon-catalyzed fusion systems or thermonuclear fusion-fission fuel factory hybrid systems

  11. The IAEA power reactor information system - PRIS

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  12. 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)

  13. Comprehensive safety analysis code system for nuclear fusion reactors III: Ex-vessel LOCA analyses considering passive safety

    International Nuclear Information System (INIS)

    Honda, T.; Okazaki, T.; Maki, K.; Uda, T.; Seki, Y.; Aoki, I.; Kunugi, T.

    1996-01-01

    Ex-vessel loss-of-coolant accidents (LOCAs) in a fusion reactor have been analyzed to investigate the possibility of passive plasma shutdown. For this purpose, a hybrid code of the plasma dynamics and thermal characteristics of the reactor structures, which has been modified to include the impurity emission from plasma-facing components (PFCs), has been developed. Ex-vessel LOCAs of the cooling system during the ignition operation in the International Thermonuclear Experimental Reactor (ITER), in which graphite PFCs were employed in conceptual design activity, were assumed. When double-ended break occurs at the cold leg of the divertor cooling system, the copper cooling tube begins to melt within 3 s after the LOCA, even though the plasma is passively shut down at nearly 4 s. An active plasma shutdown system will be needed for such rapid transient accidents. On the other hand, when a small (1%) break LOCA occurs there, the plasma is passively shut down at nearly 36 s, which happens before the copper cooling tube begins to melt. When the double-ended break LOCA occurs at the cold leg of the first-wall cooling system, there is enough time (nearly 100 s) to shut down the plasma with a controllable method before the reactor structures are damaged. 21 refs., 8 figs

  14. Introduction to Safety Analysis Approach for Research Reactors

    International Nuclear Information System (INIS)

    Park, Suki

    2016-01-01

    The research reactors have a wide variety in terms of thermal powers, coolants, moderators, reflectors, fuels, reactor tanks and pools, flow direction in the core, and the operating pressure and temperature of the cooling system. Around 110 research reactors have a thermal power greater than 1 MW. This paper introduces a general approach to safety analysis for research reactors and deals with the experience of safety analysis on a 10 MW research reactor with an open-pool and open-tank reactor and a downward flow in the reactor core during normal operation. The general approach to safety analysis for research reactors is described and the design features of a typical open-pool and open-tank type reactor are discussed. The representative events expected in research reactors are investigated. The reactor responses and the thermal hydraulic behavior to the events are presented and discussed. From the minimum CHFR and the maximum fuel temperature calculated, it is ensured that the fuel is not damaged in the step insertion of reactivity by 1.8 mk and the failure of all primary pumps for the reactor with a 10 MW thermal power and downward core flow

  15. Accident analysis in research reactors

    International Nuclear Information System (INIS)

    Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

    2007-01-01

    With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)

  16. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  17. The NAA system at the reactor BER II clinical analysis of fluorine, calcium, phosphorus and aluminium in bone biopsies, comparison with photon activation analysis and ion sensitive electrode

    International Nuclear Information System (INIS)

    Gawlik, D.; Gatschke, W.; Behne, D.; Braetter, P.

    1981-01-01

    Over the past few years an irradiation and measuring system has been developed for the reactor BER II at the Hahn-Meitner-Institute. It was to be suited in particular to the neutron activation analysis of biological material via short lived radionuclides. The general characteristics of this equipment and some experimental details as the recycling facility and the device used for measuring the irradiation time, the waiting time and the neutron fluency are described. This equipment was used to determine physiological concentrations of the elements sodium, chlorine, aluminium, phosphorus, fluorine, calcium, copper and selenium in biological material. The results have been compared with those obtained by photon activation analysis or ion sensitive electrode. As an example of the application of the apparatus the authors report on the clinical analysis of Ca, F, P and Al in the control of the treatment of patients suffering from osteoporosis. (author)

  18. Analysis of Russian transition scenarios to innovative nuclear energy system based on thermal and fast reactors with closed nuclear fuel cycle using INPRO methodology

    International Nuclear Information System (INIS)

    Kagramanyan, V.S.; Poplavskaya, E.V.; Korobeynikov, V.V.; Kalashnikov, A.G.; Moseev, A.L.; Korobitsyn, V.E.; Andreeva-Andrievskaya, L.N.

    2011-01-01

    This paper presents the results of the analysis of modeling of Russian nuclear energy (NE) scenarios on the basis of thermal and fast reactors with closed nuclear fuel cycle (NFC). Modeling has been carried out with use of CYCLE code (SSC RF IPPE's tool) designed for analysis of Nuclear Energy System (NES) with closed NFC taking into account plutonium and minor actinides (MA) isotopic composition change during multi-recycling of fuel in fast reactors. When considering fast reactor introduction scenarios, one of important questions is to define optimal time for their introduction and related NFC's facilities. Analysis of the results obtained has been fulfilled using the key INPRO indicators for sustainable energy development. It was shown that a delay in fast reactor introduction led to serious ecological, social and finally economic risks for providing energy security and sustainable development of Russia in long-term prospects and loss of knowledge and experience in mastering innovative technologies of fast reactors and related nuclear fuel cycle. (author)

  19. Kinetic analysis of sub-prompt-critical reactor assemblies

    International Nuclear Information System (INIS)

    Das, S.

    1992-01-01

    Neutronic analysis of safety-related kinetics problems in experimental neutron multiplying assemblies has been carried out using a sub-prompt-critical reactor model. The model is based on the concept of a sub-prompt-critical nuclear reactor and the concept of instantaneous neutron multiplication in a reactor system. Computations of reactor power, period and reactivity using the model show excellent agreement with results obtained from exact kinetics method. Analytic expressions for the energy released in a controlled nuclear power excursion are derived. Application of the model to a Pulsed Fast Reactor gives its sensitivity between 4 and 5. (author). 6 refs., 4 figs., 1 tab

  20. Consequence analysis for nuclear reactors, Yongbyon

    International Nuclear Information System (INIS)

    Kang, Taewook; Jae, Moosung

    2017-01-01

    Since the Fukushima nuclear power plant accidents in 2011, there have been an increased public anxiety about the safety of nuclear power plants in Korea. The lack of safeguards and facility aging issues at the Yongbyon nuclear facilities have increased doubts. In this study, the consequence analysis for the 5-MWe graphite-moderated reactor in North Korea was performed. Various accident scenarios including accidents at the interim spent fuel pool in the 5-MWe reactor have been developed and evaluated quantitatively. Since data on the design and safety system of nuclear facilities are currently insufficient, the release fractions were set by applying the alternative source terms made for utilization in the analysis of a severe accident by integrating the results of studies of severe accidents occurred before. The calculation results show the early fatality zero deaths and latent cancer fatality about only 13 deaths in Seoul. Thus, actual impacts of a radiological release will be psychological in terms of downwind perceptions and anxiety on the part of potentially exposed populations. Even considering the simultaneous accident occurrence in both 5-MWe graphite-moderated reactor and 100-MWt light water reactor, the consequence analysis using the MACCS2 code shows no significant damage to people in South Korea. (author)

  1. CRAW: An expert system for nuclear reactor cover gas alarm analysis

    International Nuclear Information System (INIS)

    Zimmerman, B.D.; Rawlins, J.A.

    1987-08-01

    CRAW is an expert system designed to distinguish fuel pin rupture at the Fast Flux Test Facility. The software is used to evaluate tag gases detected by the alarm system and access operational consequences of a rupture. The ability to distinguish fuel rod as opposed to fuel pin rupture is included. The software is implemented on an IBM-PC

  2. The analysis of mechanical behaviour of reactor coolant system layout scheme with 60 degree angle for the second phase project of Qinshan NPP

    International Nuclear Information System (INIS)

    Yu Ruhong

    1993-01-01

    For the reactor coolant system of the second phase project of Qinshan NPP, the layout scheme with two loops and an angle of 60 degree is adopted. In this scheme, two loops are connected to reactor pressure vessel (RPV), and the angle included between the inlet and outlet nozzles of the RPV is 60 degree in a same loop. The issues involved in the analysis of mechanical behaviour of piping system to demonstrate the validity of such a scheme are described briefly in the paper, including the modelling technique adopted in establishing mathematical model, the methods used for structural analysis of piping system, stress and fatigue analysis in piping fittings. A brief description of the calculation results are given and the feasibility and rationality are discussed

  3. Dosimetry system of the RB reactor

    International Nuclear Information System (INIS)

    Lolic, B.; Vukadin, D.

    1962-01-01

    Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

  4. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

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

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

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

  6. Singular system analysis of the Local Power Range Monitor (LPRM) readings of a Boiling Water Reactor (BWR) in an unstable event

    International Nuclear Information System (INIS)

    Ginestar Peiro, D.; Verdu, G.; Miro, R.

    2006-01-01

    Singular system analysis is a successful technique to separate oscillating components from a given signal. A methodology is proposed to apply this technique to the signals obtained from the LPRMs of a boiling water reactor core and extract the contributions of the in-phase oscillation and the out-of-phase oscillations from the LPRM readings during an unstable event. This methodology has been validated with synthetic signals and simulations of in-phase and out-of-phase oscillations of the Leibstadt reactor. Finally, one case of Ringhals I Stability Benchmark has been analysed. (author)

  7. Emergency system for nuclear reactors

    International Nuclear Information System (INIS)

    1976-01-01

    The invention concerns a circuit called 'of emergency help' intended to remove, in a safe and quick manner, the residual thermal power on the safety vessel of a fast neutron reactor cooled by a liquid metal flow, in the event of a failure occurring inside the main reactor vessel or on it. This system includes a network of spray nozzle tubes, distributed around and near the external surface of the safety vessel, to project on to the surface of the vessel a mist of a liquid having high latent vaporisation heat. The steam produced on contact with the safety vessel is collected in the space provided between the safety vessel and the external protection vessel by at least one collector pipe for dischaging this steam outside the vessel. Under a preferred design mode of the invention the liquid is water the use of which turns out to be particularly advantageous in practice owing to its favourable physical properties and its low cost [fr

  8. An analysis of the water-level monitoring system for a boiling-water reactor

    International Nuclear Information System (INIS)

    Carlson, R.W.; Belblidia, L.A.; Russell, J.L. Jr.

    1985-01-01

    The water-level instrumentation system is very important to the overall safety of a BWR. This system is being monitored by the Safety Parameter Display System (SPDS) that is being installed in Georgia Power Company's Plant Hatch. One of the most significant functions of the SPDS is the comparison of redundant instrument readings and formation of the best estimate of each parameter from those readings which are consistent. When comparing water-level instrument readings, it is necessary to correct the individual readings for differences between current and calibration conditions as well as for differences between calibration conditions for the multiple instruments. This paper documents the examination of the water-level instrumentation system at Plant Hatch and presents the development of the equations that were used to determine the differences between indicated and actual water levels. (author)

  9. The Optimization of power reactor control system

    International Nuclear Information System (INIS)

    Danupoyo, S.D.

    1997-01-01

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

  10. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Bolton, John A.

    1970-01-01

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  11. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  12. Reactor alarm system development and application issues

    Energy Technology Data Exchange (ETDEWEB)

    Drexler, J E; Oicese, G O [INVAP S.E. (Argentina)

    1997-09-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ``Reactor Alarm System`` (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: (a) Alarm Panel Display; (b) Alarm Page; (c) Alarm Masking and Inhibition; (d) Alarms Color and Attributes; (e) Condition Classification; and (f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: (a) Multiple-copy alarm summaries; (b) Improved alarm handling; (c) Extended dictionary; and (d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs.

  13. Reactor alarm system development and application issues

    International Nuclear Information System (INIS)

    Drexler, J.E.; Oicese, G.O.

    1997-01-01

    The new hardware and software technologies, and the need in research reactors for assistance systems in operation and maintenance, have given an appropriate background to develop a computer based system named ''Reactor Alarm System'' (RAS). RAS is a software package, user oriented, with emphasis on production, experiments and maintenance goals. It is designed to run on distributed systems conformed with microcomputers under QNX operating system. RAS main features are: a) Alarm Panel Display; b) Alarm Page; c) Alarm Masking and Inhibition; d) Alarms Color and Attributes; e) Condition Classification; and f) Arrangement Presentation. RAS design allows it to be installed as a part of a computer based Supervision and Control System in new installations or retrofit existing reactor instrumentation systems. The analysis of human factors during development stage and successive user feedback from different applications, brought out several RAS improvements: a) Multiple-copy alarm summaries; b) Improved alarm handling; c) Extended dictionary; and d) Enhanced hardware availability. It has proved successful in providing new capabilities for operators, and also has shown the continuous increase of user-demands, reflecting the expectations placed today on computer-based systems. (author). 6 figs, 1 tabs

  14. Analysis of tritium migration and deposition in fusion-reactor systems

    International Nuclear Information System (INIS)

    Holland, D.F.; Merrill, B.J.

    1981-01-01

    EG and G Idaho, Inc., is developing a safety analysis code, TMAP (Tritium Migration Analysis Program), to determine tritium loss into the environment and tritium buildup in components, coolants, and walls during normal and accident conditions. TMAP determines the thermal response of structures, solves equations for hydrogen movement through surface and in bulk materials, and also includes equations for chemical reactions. TMAP calculations of tritium movement through metal barriers at low tritium pressure agree closely with experimental measurements. The code has been used to predict inventory buildup and loss to the coolant of tritium implanted in the first wall of a fusion device, and concentrations during cleanup of tritium released into an enclosure

  15. A new balance-of-plant model for the SASSYS-1 LMR [liquid metal reactor] systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1989-01-01

    A balance-of-plant (BOP) model has been developed for use within the SASSYS-1 liquid-metal reactor systems analysis code. This model expands the scope of SASSYS-1 so that the code can explicitly model the waterside components of a nuclear power plant; previously, only the water side of the steam generators could be modeled, with the remainder of the water side represented by boundary conditions on the steam generator. The model represents the BOP a set of flow paths and path junctions; the mass and energy equations are solved at the junctions, and the momentum equation is solved along the flow paths. The junctions are thus mass and energy cells, and the paths are momentum cells. The various waterside component models (pumps, valves, etc.) are specialized types of energy or momentum cells, as appropriate. The solution scheme implicitly couples the energy cells through the momentum cells and solves simultaneously for pressures and enthalpies within the energy cells and for flows within the momentum cells

  16. Application of simple approximate system analysis methods for reliability and availability improvement of reactor WWER-1000

    International Nuclear Information System (INIS)

    Manchev, B.; Marinova, B.; Nenkova, B.

    2001-01-01

    The method described on this report provides a set of simple, easily understood 'approximate' models applicable to a large class of system architectures. Constructing a Markov model of each redundant subsystem and its replacement after that by a pseudo-component develops the approximation models. Of equal importance, the models can be easily understood even of non-experts, including managers, high-level decision-makers and unsophisticated consumers. A necessary requirement for their application is the systems to be repairable and the mean time to repair to be much smaller than the mean time to failure. This ia a case most often met in the real practice. Results of the 'approximate' model application on a technological system of Kozloduy NPP are also presented. The results obtained can be compared quite favorably with the results obtained by using SAPHIRE software

  17. Analysis of reactor noise; Analiza reaktorskih sumova

    Energy Technology Data Exchange (ETDEWEB)

    Velickovic, Lj [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1967-11-15

    This paper describes the theoretical model for interpretation of experimental results, experimental method for study of reactor noise at the RB reactor, numerical treatment of experimental results by correlation technique for analysis of reactor noise. A computer code was written to obtain autocorrelation function and spectral density function. Experimental results obtained by oscillator technique, pulse technique, and autocorrelation method are presented and discussed.

  18. Digital control system of advanced reactor

    International Nuclear Information System (INIS)

    Peng Huaqing; Zhang Rui; Liu Lixin

    2001-01-01

    This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

  19. Analysis and testing of W-DHR system for decay heat removal in the lead-cooled ELSY reactor

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Meloni, Paride; Polidori, Massimiliano; Gaggini, Piero; Labanti, Valerio; Tarantino, Mariano; Cinotti, Luciano; Presciuttini, Leonardo

    2009-01-01

    An innovative LFR system that complies with GEN IV goals is under design in the frame of ELSY European project. ELSY is a lead-cooled pool-type reactor of about 1500 MW thermal power which normally relies on the secondary system for decay heat removal. Since the secondary system is not safety-grade and must be fully depressurized in case of detection of a steam generator tube rupture, an independent and much reliable decay heat removal (DHR) system is foreseen on the primary side. Owing to the limited capability of the Reactor Vessel Air Cooling System (RVACS) in this large power reactor, additional safety-grade loops equipped with coolers immersed in the primary coolant are necessary for an efficient removal of decay heat. Some of these loops (W-DHR) are of innovative design and may operate with water at atmospheric pressure. In the frame of the ICE program to be performed on the integral facility CIRCE at ENEA/Brasimone research centre within the EUROTRANS European project, integral circulation experiments with core heat transport and heat removal by steam generator will be conducted in a reactor pool-type configuration. Taking advantage from this experimental program, a mock-up of W-DHR heat exchanger will be tested in order to investigate its functional behavior for decay heat removal. Some pre-test calculations of W-DHR heat exchanger operation in CIRCE have been performed with the RELAP5 thermal-hydraulic code in order to support the heat exchanger design and test conduct. In this paper the experimental activity to be conducted in CIRCE and main results from W-DHR pre-test calculations are presented, along with a preliminary investigation of the W-DHR system efficiency in ELSY configuration. (author)

  20. Methodology for time-dependent reliability analysis of accident sequences and complex reactor systems

    International Nuclear Information System (INIS)

    Paula, H.M.

    1984-01-01

    The work presented here is of direct use in probabilistic risk assessment (PRA) and is of value to utilities as well as the Nuclear Regulatory Commission (NRC). Specifically, this report presents a methodology and a computer program to calculate the expected number of occurrences for each accident sequence in an event tree. The methodology evaluates the time-dependent (instantaneous) and the average behavior of the accident sequence. The methodology accounts for standby safety system and component failures that occur (a) before they are demanded, (b) upon demand, and (c) during the mission (system operation). With respect to failures that occur during the mission, this methodology is unique in the sense that it models components that can be repaired during the mission. The expected number of system failures during the mission provides an upper bound for the probability of a system failure to run - the mission unreliability. The basic event modeling includes components that are continuously monitored, periodically tested, and those that are not tested or are otherwise nonrepairable. The computer program ASA allows practical applications of the method developed. This work represents a required extension of the presently available methodology and allows a more realistic PRA of nuclear power plants

  1. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

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

    2011-01-01

    Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

  2. Recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Braun, H. E.; Dollard, W. J.; Tower, S. N.

    1980-01-01

    A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

  3. A record/replay system for the harmonic analysis of nuclear reactor flux noise

    International Nuclear Information System (INIS)

    Lawrence, L.A.J.; Corran, E.R.

    1960-02-01

    The design of a record/replay system for the determination of neutron flux spectra is discussed and circuit details and performance figures are given. Frequency modulation is used with a carrier frequency of 1,000 cycles per second, the complete system having an overall bandwidth of 100 cycles per second. The noise is recorded on magnetic tape and when replayed is analysed into a spectrum by means of a selective amplifier used in conjunction with an infinitely variable speed control on the tape. It is shown that if the lowest spectral frequency is .01 cycles per second a recording time of many hours is necessary. The DIDO noise spectrum is analysed and shown to be contained in a bandwidth of a few cycles per second. (author)

  4. CFD analysis of bubble hydrodynamics in a fuel reactor for a hydrogen-fueled chemical looping combustion system

    International Nuclear Information System (INIS)

    Harichandan, Atal Bihari; Shamim, Tariq

    2014-01-01

    Highlights: • Computational study of the fuel reactor of chemical looping combustion technology. • The results yield better understanding of the bubble hydrodynamics in fuel reactor. • Increasing the reactor bed length increases the conversion rate. • Small oxygen carrier particles improves the conversion rate. - Abstract: This study investigates the temporal development of bubble hydrodynamics in the fuel reactor of a hydrogen-fueled chemical looping combustion (CLC) system by using a computational model. The model also investigates the molar fraction of products in gas and solid phases. The study assists in developing a better understanding of the CLC process, which has many advantages such as being a potentially promising candidate for an efficient carbon dioxide capture technology. The study employs the kinetic theory of granular flow. The reactive fluid dynamic system of the fuel reactor is customized by incorporating the kinetics of an oxygen carrier reduction into a commercial computational fluid dynamics (CFD) code. An Eulerian multiphase treatment is used to describe the continuum two-fluid model for both gas and solid phases. CaSO 4 and H 2 are used as an oxygen carrier and a fuel, respectively. The computational results are validated with the experimental and numerical results available in the open literature. The CFD simulations are found to capture the features of the bubble formation, rise and burst in unsteady and quasi-steady states very well. The results show a significant increase in the conversion rate with higher dense bed height, lower bed width, higher free board height and smaller oxygen carrier particles which upsurge an overall performance of the CLC plant

  5. A ''SuperCode'' for performing systems analysis of tokamak experiments and reactors

    International Nuclear Information System (INIS)

    Haney, S.W.; Barr, W.L.; Crotinger, J.A.; Perkins, L.J.; Solomon, C.J.; Chaniotakis, E.A.; Freidberg, J.P.; Wei, J.; Galambos, J.D.; Mandrekas, J.

    1992-01-01

    A new code, named the ''SUPERCODE,'' has been developed to fill the gap between currently available zero dimensional systems codes and highly sophisticated, multidimensional plasma performance codes. The former are comprehensive in content, fast to execute, but rather simple in terms of the accuracy of the physics and engineering models. The latter contain state-of-the-art plasma physics modelling but are limited in engineering content and time consuming to run. The SUPERCODE upgrades the reliability and accuracy of systems codes by calculating the self consistent 1 1/2 dimensional MHD-transport plasma evolution in a realistic engineering environment. By a combination of variational techniques and careful formation, there is only a modest increase in CPU time over O-D runs, thereby making the SUPERCODE suitable for use as a systems studies tool. In addition, considerable effort has been expended to make the code user- and programming-friendly, as well as operationally flexible, with the hope of encouraging wide usage throughout the fusion community

  6. Lead reactor strategy economical analysis

    International Nuclear Information System (INIS)

    Ciotti, Marco

    2013-01-01

    Conclusions: • A first attempt to evaluate LFR power plant electricity production cost has been performed; • Electricity price is similar to Gen III + plants; • The estimation accuracy is probably low; • Possible costs reduction could arise from coolant characteristics that may improve safety and simplicity by design; • Accident perception, not acceptable by public opinion, may be changed with low potential energy system (non exploding coolant); • Sustainability improvement could open to a better Public acceptance, depending on us. • Problems may arise in coupling a high capital cost low fuel cost plant in a grid with large amount of intermittent sources with priority dispatch. • Lead fast reactors can compete

  7. Swimming pool reactor reliability and safety analysis

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1997-01-01

    A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

  8. Nuclear characteristics analysis report for system-integrated modular advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Yoon; Lee, Chung Chan; Song, Jae Seung; Cho, Byung Oh; Zee, Sung Quun

    1998-11-01

    This report present nuclear characteristics analysis results for SMART. Information is given on fuel loading, power density distributions, reactivity coefficients and control rod worths. The core consists of 57 modified Korean Standard Fuel Assemblies (m-KOFAs), and all fuel assemblies contain burnable absorbers to control the power distribution and the excess reactivity that is required for soluble boron-free and ultra longer cycle operation. The cycle length of SMART amounts to 990 EFPD corresponding to a cycle burnup of 26,250 MWD/MTU. (author). 4 refs., 6 tabs., 85 figs.

  9. Nuclear characteristics analysis report for system-integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Park, Sang Yoon; Lee, Chung Chan; Song, Jae Seung; Cho, Byung Oh; Zee, Sung Quun

    1998-11-01

    This report present nuclear characteristics analysis results for SMART. Information is given on fuel loading, power density distributions, reactivity coefficients and control rod worths. The core consists of 57 modified Korean Standard Fuel Assemblies (m-KOFAs), and all fuel assemblies contain burnable absorbers to control the power distribution and the excess reactivity that is required for soluble boron-free and ultra longer cycle operation. The cycle length of SMART amounts to 990 EFPD corresponding to a cycle burnup of 26,250 MWD/MTU. (author). 4 refs., 6 tabs., 85 figs

  10. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  11. THYDE-P2 code: RCS (reactor-coolant system) analysis code

    International Nuclear Information System (INIS)

    Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo

    1986-12-01

    THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)

  12. Study, analysis, assess and compare the nuclear engineering systems of nuclear power plant with different reactor types VVER-1000, namely AES-91, AES-92 and AES-2006

    International Nuclear Information System (INIS)

    Le Van Hong; Tran Chi Thanh; Hoang Minh Giang; Le Dai Dien; Nguyen Nhi Dien; Nguyen Minh Tuan

    2015-01-01

    On November 25, 2009, in Hanoi, the National Assembly had been approved the resolution about policy for investment of nuclear power project in Ninh Thuan province which include two sites, each site has two units with power around 1000 MWe. For the nuclear power project at Ninh Thuan 1, Vietnam Government signed the Joint-Governmental Agreement with Russian Government for building the nuclear power plant with reactor type VVER. At present time, the Russian Consultant proposed four reactor technologies can be used for Ninh Thuan 1 project, namely: AES-91, AES-92, AES-2006/V491 and AES-2006/V392M. This report presents the main reactor engineering systems of nuclear power plants with VVER-1000/1200. The results from analysis, comparison and assessment between the designs of AES-91, AES-92 and AES-2006 are also presented. The obtained results show that the type AES-2006 is appropriate selection for Vietnam. (author)

  13. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  14. Breeding description for fast reactors and symbiotic reactor systems

    International Nuclear Information System (INIS)

    Hanan, N.A.

    1979-01-01

    A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  16. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  17. Subchannel analysis in nuclear reactors

    International Nuclear Information System (INIS)

    Ninokata, H.; Aritomi, M.

    1992-01-01

    This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)

  18. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    International Nuclear Information System (INIS)

    Lázaro, A.; Ammirabile, L.; Bandini, G.; Darmet, G.; Massara, S.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Mikityuk, K.; Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Stempniewicz, M.

    2014-01-01

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes

  19. Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

    Energy Technology Data Exchange (ETDEWEB)

    Lázaro, A., E-mail: aurelio.lazaro-chueca@ec.europa.eu [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); UPV—Universidad Politecnica de Valencia, Cami de vera s/n-46002, Valencia (Spain); Ammirabile, L. [JRC-IET European Commission—Westerduinweg 3, PO Box-2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Massara, S. [EDF, 1 avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2-28006 Madrid (Spain); Mikityuk, K. [PSI—Paul Scherrer Institut, 5232 Villigen Switzerland (Switzerland); Schikorr, M.; Bubelis, E.; Ponomarev, A.; Kruessmann, R. [KIT—Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen Germany (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, PO Box 9034 6800 ES, Arnhem (Netherlands)

    2014-01-15

    Highlights: • Ten system-code models of the ESFR were developed in the frame of the CP-ESFR project. • Eight different thermohydraulic system codes adapted to sodium fast reactor's technology. • Benchmarking exercise settled to check the consistency of the calculations. • Upgraded system codes able to simulate the reactivity feedback and key safety parameters. -- Abstract: The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes.

  20. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  1. Reliability analysis with the simulator S.ESCAF of a very complex sequential system: the electrical power supply system of a nuclear reactor

    International Nuclear Information System (INIS)

    Blot, M.

    1987-06-01

    The reliability analysis of complex sequential systems, in which the order of arrival of the events must be taken into account, can be very difficult, because the use of the classical modelling technique of Markov diagrams leads to an important limitation on the number of components which can be handled. The desk-top apparatus S.ESCAF, which electronically simulates very closely the behaviour of the system being studied, and is very easy to use, even by a non specialist in electronics, allows one to avoid these inconveniences and to enlarge considerably the analysis possibilities. This paper shows the application of the S.ESCAF method to the electrical power supply system of a nuclear reactor. This system requires the simulation of more than forty components with about sixty events such as failure, repair and refusal to start. A comparison of times necessary to perform the analysis by these means and by other methods is described, and the advantages of S.ESCAF are presented

  2. Uncertainty analysis of reactor safety systems with statistically correlated failure data

    International Nuclear Information System (INIS)

    Dezfuli, H.; Modarres, M.

    1985-01-01

    The probability of occurrence of the top event of a fault tree is estimated from failure probability of components that constitute the fault tree. Component failure probabilities are subject to statistical uncertainties. In addition, there are cases where the failure data are statistically correlated. Most fault tree evaluations have so far been based on uncorrelated component failure data. The subject of this paper is the description of a method of assessing the probability intervals for the top event failure probability of fault trees when component failure data are statistically correlated. To estimate the mean and variance of the top event, a second-order system moment method is presented through Taylor series expansion, which provides an alternative to the normally used Monte-Carlo method. For cases where component failure probabilities are statistically correlated, the Taylor expansion terms are treated properly. A moment matching technique is used to obtain the probability distribution function of the top event through fitting a Johnson Ssub(B) distribution. The computer program (CORRELATE) was developed to perform the calculations necessary for the implementation of the method developed. The CORRELATE code is very efficient and consumes minimal computer time. This is primarily because it does not employ the time-consuming Monte-Carlo method. (author)

  3. Study on statistical analysis of nonlinear and nonstationary reactor noises

    International Nuclear Information System (INIS)

    Hayashi, Koji

    1993-03-01

    For the purpose of identification of nonlinear mechanism and diagnosis of nuclear reactor systems, analysis methods for nonlinear reactor noise have been studied. By adding newly developed approximate response function to GMDH, a conventional nonlinear identification method, a useful method for nonlinear spectral analysis and identification of nonlinear mechanism has been established. Measurement experiment and analysis were performed on the reactor power oscillation observed in the NSRR installed at the JAERI and the cause of the instability was clarified. Furthermore, the analysis and data recording methods for nonstationary noise have been studied. By improving the time resolution of instantaneous autoregressive spectrum, a method for monitoring and diagnosis of operational status of nuclear reactor has been established. A preprocessing system for recording of nonstationary reactor noise was developed and its usability was demonstrated through a measurement experiment. (author) 139 refs

  4. Cryogenic system design for a compact tokamak reactor

    International Nuclear Information System (INIS)

    Slack, D.S.; Kerns, J.A.; Miller, J.R.

    1988-01-01

    The International Tokamak Engineering Reactor (ITER) is a program presently underway to design a next-generation tokamak reactor. The cryogenic system for this reactor must meet unusual and new requirements. Unusually high heat loads (100 kW at 4.5 K) must be handled because neutron shielding has been limited to save space in the reactor core. Also, large variations in the cryogenics loads occur over short periods of time because of the pulsed nature of some of the operating scenarios. This paper describes a workable cryogenic system design for a compact tokamak reactor such as ITER. A design analysis is presented dealing with a system that handles transient loads, coil quenches, reactor cool-down and the effect of variations in helium-supply temperatures on the cryogenic stability of the coils. 5 refs., 4 figs., 1 tab

  5. Reactor shutdown system of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

    2002-01-01

    Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

  6. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  7. Development of fault diagnostic technique using reactor noise analysis

    International Nuclear Information System (INIS)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B.

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  8. Development of fault diagnostic technique using reactor noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Ho; Kim, J. S.; Oh, I. S.; Ryu, J. S.; Joo, Y. S.; Choi, S.; Yoon, D. B

    1999-04-01

    The ultimate goal of this project is to establish the analysis technique to diagnose the integrity of reactor internals using reactor noise. The reactor noise analyses techniques for the PWR and CANDU NPP(Nuclear Power Plants) were established by which the dynamic characteristics of reactor internals and SPND instrumentations could be identified, and the noise database corresponding to each plant(both Korean and foreign one) was constructed and compared. Also the change of dynamic characteristics of the Ulchin 1 and 2 reactor internals were simulated under presumed fault conditions. Additionally portable reactor noise analysis system was developed so that real time noise analysis could directly be able to be performed at plant site. The reactor noise analyses techniques developed and the database obtained from the fault simulation, can be used to establish a knowledge based expert system to diagnose the NPP's abnormal conditions. And the portable reactor noise analysis system may be utilized as a substitute for plant IVMS(Internal Vibration Monitoring System). (author)

  9. Analysis of the Integral Response of CAREM Reactor and the Residual Heat Removal System During a Failure of the Steam Generators Feed Water System

    International Nuclear Information System (INIS)

    Gimenez, Marcelo; Zanocco, Pablo; Schlamp, Miguel

    2000-01-01

    A global analysis of the behavior of Carem-25 Reactor and Residual Heat Removal System (RHRS) to mitigate a loss of heat sink accident is done in the present work.The proposed RHRS removes 2 MW of power and is duplicated to fulfill the redundancy criteria.It consists of two condensers with two tubes in a parallel array.Each tube has 2 S CH 160 TP 347 SS and 2 m 2 of area.The RHRS design requierements (for this accidental sequence) are: Short-term: primary circuit pressure must remain below the safety valves opening set point and the condensers must not flood in order to avoid instabilities. Long-term: reach hot-shutdown condition (primary circuit pressure below 2.3 MPa) at least before 48 hrs. Short-term reactor behavior is simulated using RELAP5 with a detail nodalization of the primary circuit and RHRS.Long term performance is simulated with a simple and conservative model, assuming a saturated primary circuit. This condition is expected during RHRS operation

  10. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    International Nuclear Information System (INIS)

    Corradin, Michael; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-01-01

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  11. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  12. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  13. Failure Mode and Effects Analysis (FMEA) of the Emergency Core Cooling System (ECCS) for a Westinghouse type 312, three loop pressurized water reactor

    International Nuclear Information System (INIS)

    Shopsky, W.E.

    1977-01-01

    The Emergency Core Cooling System (ECCS) is a Safeguards System designed to cool the core in the unlikely event of a Loss-of-Coolant Accident (LOCA) in the primary reactor coolant system as well as to provide additional shutdown capability following a steam break accident. The system is designed for a high reliability of providing emergency coolant and shutdown reactivity to the core for all anticipated occurrences of such accidents. The ECCS by performing its intended function assures that fuel and clad damage is minimized during accident conditions thus reducing release of fission products from the fuel. The ECCS is designed to perform its function despite sustaining a single failure by the judicious use of equipment and flow path redundancy within and outside the containment structure. By the use of an analytic tool, a Failure Mode and Effects Analysis (FMEA), it is shown that the ECCS is in compliance with the Single Failure Criterion established for active failures of fluid systems during short and long term cooling of the reactor core following a LOCA or steam break accident. An analysis was also performed with regards to passive failure of ECCS components during long-term cooling of the core following an accident. The design of the ECCS was verified as being able to tolerate a single passive failure during long-term cooling of the reactor core following an accident. The FMEA conducted qualitatively demonstrates the reliability of the ECCS (concerning active components) to perform its intended safety function

  14. WRAP: a water reactor analysis package

    International Nuclear Information System (INIS)

    Anderson, M.M.

    1977-06-01

    The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

  15. Molecular ecology of anaerobic reactor systems

    DEFF Research Database (Denmark)

    Hofman-Bang, H. Jacob Peider; Zheng, D.; Westermann, Peter

    2003-01-01

    Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible for these ......Anaerobic reactor systems are essential for the treatment of solid and liquid wastes and constitute a core facility in many waste treatment plants. Although much is known about the basic metabolism in different types of anaerobic reactors, little is known about the microbes responsible...... to the abundance of each microbe in anaerobic reactor systems by rRNA probing. This chapter focuses on various molecular techniques employed and problems encountered when elucidating the microbial ecology of anaerobic reactor systems. Methods such as quantitative dot blot/fluorescence in-situ probing using various...

  16. Modeling and analysis of the disk MHD generator component of a gas core reactor/MHD Rankine cycle space power system

    International Nuclear Information System (INIS)

    Welch, G.E.; Dugan, E.T.; Lear, W.E. Jr.; Appelbaum, J.G.

    1990-01-01

    A gas core nuclear reactor (GCR)/disk magnetohydrodynamic (MHD) generator direct closed Rankine space power system concept is described. The GCR/disk MHD generator marriage facilitates efficient high electric power density system performance at relatively high operating temperatures. The system concept promises high specific power levels, on the order of 1 kW e /kg. An overview of the disk MHD generator component magnetofluiddynamic and plasma physics theoretical modeling is provided. Results from a parametric design analysis of the disk MHD generator are presented and discussed

  17. Seal analysis technology for reactor pressure vessel

    International Nuclear Information System (INIS)

    Zheng Liangang; Zhang Liping; Yang Yu; Zang Fenggang

    2009-01-01

    There is the coolant with radiation, high temperature and high pressure in the reactor pressure vessel (RPV). It is closely correlated to RPV sealing capability whether the whole nuclear system work well or not. The aim of this paper is to study the seal analysis method and technology, such as the pre-tensioning of the bolt, elastoplastic contact and coupled technology of thermal and structure. The 3 D elastoplastic seal analysis method really and generally consider the loads and model the contact problem with friction between the contact plates. This method is easier than the specialized seal program and used widely. And it is more really than the 2 D seal analysis method. This 3 D elastoplastic seal analysis method has been successfully used in the design and analysis of RPV. (authors)

  18. RSAS: a Reactor Safety Assessment System

    International Nuclear Information System (INIS)

    Sebo, D.E.; Dixon, B.W.; Bray, M.A.

    1985-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (NRC). RSAS is being developed for use at the NRC's Operations Center in the event of a serious incident at a licensed nuclear power plant. The system generates situation assessments for the NRC Reactor Safety Team based on a limited number of plant parameters, known operator actions, and plant status data. The RSAS rule base currently covers one reactor type. The extension of the rule base to other reactor types is also discussed

  19. Development of a core follow calculational system for research reactors

    International Nuclear Information System (INIS)

    Muller, E.Z.; Ball, G.; Joubert, W.R.; Schutte, H.C.; Stoker, C.C.; Reitsma, F.

    1994-01-01

    Over the last few years a comprehensive Pressurized Water Reactor and Materials Testing Reactor core analysis code system based on modern reactor physics methods has been under development by the Atomic Energy Corporation of South Africa. This system, known as OSCAR-3, will incorporate a customized graphical user interface and data management system to ensure user-friendliness and good quality control. The system has now reached the stage of development where it can be used for practical MTR core analyses. This paper describes the current capabilities of the components of the OSCAR-3 package, their integration within the package, and outlines future developments. 10 refs., 1 tab., 1 fig

  20. Code systems for effective and precise calculation of the basic neutron characteristics, core loading optimization, analysis and estimation of the operation regimes of WWER type reactors

    International Nuclear Information System (INIS)

    Apostolov, T.; Ivanov, K.; Prodanova, R.; Manolova, M.; Petrova, T.; Alekova, G.

    1993-01-01

    Two directions for investigations are suggested: 1) Analysis and evaluation of the real loading patterns and operational regimes for Kozloduy NPP WWER-440 and WWER-1000 in the frame of the recent safety criteria and nuclear power plant operating limits. 2) Development of modern code system for WWER type reactor core analysis with advanced features: new design and materials for fuel and control rods, increasing the fuel enrichment, using the integral and discrete burnable absorbers etc. The fuel technology design evolution maximizes the fuel utilization efficiency, improves operation performance and enhances safety margins. By the joint efforts of specialists from INRNE, Sofia (BG) and KAB, Berlin (GE), the codes NESSEL-IV-EC, PYTHIA and DERAB have been developed and verified. In the frame of the PHARE programme the joint project ASPERCA has been proposed intended for reactor physics calculations with PHYBER-WWER code for safety enhancement and operation reliability improvement. In-core fuel management benchmarks for 4 cycles of unit 2 (WWER-440) and 2 cycles of unit 5 (WWER-1000) have been performed. The coordination of burnable absorber design implementation, low leakage loadings usage, reloading enrichment increase and steel content reduction in the core have made the reactor core analysis more demanding and the definition of loading patterns - more difficult. This complexity requires routine use of three-dimensional fast accurate core model with extended and updated cross section libraries. To meet the needs of WWER advanced loading patterns and in-core fuel management improvements the HEXANES code systems is being developed and qualified. Some test calculations have been carried out by the HEXANES code system investigating the influence of Gd in the fuel on the main reactor physics parameters. For reevaluation of the core safety-related design limits forming the basis of licensing procedure, the code DYN3D/M2 is used. 16 refs., 3 figs. (author)

  1. Development of the coupled 'system thermal-hydraulics, 3D reactor kinetics, and hot channel' analysis capability of the MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. J.; Chung, B. D.; Lee, W.J

    2005-02-01

    The subchannel analysis capability of the MARS 3D module has been improved. Especially, the turbulent mixing and void drift models for flow mixing phenomena in rod bundles have been assessed using some well-known rod bundle test data. Then, the subchannel analysis feature was combined to the existing coupled 'system Thermal-Hydraulics (T/H) and 3D reactor kinetics' calculation capability of MARS. These features allow the coupled 'system T/H, 3D reactor kinetics, and hot channel' analysis capability and, thus, realistic simulations of hot channel behavior as well as global system T/H behavior. In this report, the MARS code features for the coupled analysis capability are described first. The code modifications relevant to the features are also given. Then, a coupled analysis of the Main Steam Line Break (MSLB) is carried out for demonstration. The results of the coupled calculations are very reasonable and realistic, and show these methods can be used to reduce the over-conservatism in the conventional safety analysis.

  2. 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)

  3. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    International Nuclear Information System (INIS)

    1998-01-01

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS; Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  4. A Level 1+ Probabilistic Safety Assessment of the high flux Australian reactor. Vol. 2. Appendix C: System analysis models and results

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-01-01

    This section contains the results of the quantitative system/top event analysis. Section C. 1 gives the basic event coding scheme. Section C.2 shows the master frequency file (MFF), which contains the split fraction names, the top events they belong to, the mean values of the uncertainty distribution that is generated by the Monte Carlo quantification in the System Analysis module of RISKMAN, and a brief description of each split fraction. The MFF is organized by the systems modeled, and within each system, the top events associated with the system. Section C.3 contains the fault trees developed for the system/top event models and the RISKMAN reports for each of the system/top event models. The reports are organized under the following system headings: Compressed/Service Air Supply (AIR); Containment Isolation System (CIS); Heavy Water Cooling System (D20); Emergency Core Cooling System (ECCS); Electric Power System (EPS); Light Water Cooling system (H20); Helium Gas System (HE); Mains Water System (MW); Miscellaneous Top Events (MISC); Operator Actions (OPER) Reactor Protection System (RPS); Space Conditioner System (SCS); Condition/Status Switch (SWITCH); RCB Ventilation System (VENT); No. 1 Storage Block Cooling System (SB)

  5. Application of reliability-centered maintenance to boiling water reactor emergency core cooling systems fault-tree analysis

    International Nuclear Information System (INIS)

    Choi, Y.A.; Feltus, M.A.

    1995-01-01

    Reliability-centered maintenance (RCM) methods are applied to boiling water reactor plant-specific emergency core cooling system probabilistic risk assessment (PRA) fault trees. The RCM is a technique that is system function-based, for improving a preventive maintenance (PM) program, which is applied on a component basis. Many PM programs are based on time-directed maintenance tasks, while RCM methods focus on component condition-directed maintenance tasks. Stroke time test data for motor-operated valves (MOVs) are used to address three aspects concerning RCM: (a) to determine if MOV stroke time testing was useful as a condition-directed PM task; (b) to determine and compare the plant-specific MOV failure data from a broad RCM philosophy time period compared with a PM period and, also, compared with generic industry MOV failure data; and (c) to determine the effects and impact of the plant-specific MOV failure data on core damage frequency (CDF) and system unavailabilities for these emergency systems. The MOV stroke time test data from four emergency core cooling systems [i.e., high-pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), low-pressure core spray (LPCS), and residual heat removal/low-pressure coolant injection (RHR/LPCI)] were gathered from Philadelphia Electric Company's Peach Bottom Atomic Power Station Units 2 and 3 between 1980 and 1992. The analyses showed that MOV stroke time testing was not a predictor for eminent failure and should be considered as a go/no-go test. The failure data from the broad RCM philosophy showed an improvement compared with the PM-period failure rates in the emergency core cooling system MOVs. Also, the plant-specific MOV failure rates for both maintenance philosophies were shown to be lower than the generic industry estimates

  6. Linear regression and sensitivity analysis in nuclear reactor design

    International Nuclear Information System (INIS)

    Kumar, Akansha; Tsvetkov, Pavel V.; McClarren, Ryan G.

    2015-01-01

    Highlights: • Presented a benchmark for the applicability of linear regression to complex systems. • Applied linear regression to a nuclear reactor power system. • Performed neutronics, thermal–hydraulics, and energy conversion using Brayton’s cycle for the design of a GCFBR. • Performed detailed sensitivity analysis to a set of parameters in a nuclear reactor power system. • Modeled and developed reactor design using MCNP, regression using R, and thermal–hydraulics in Java. - Abstract: The paper presents a general strategy applicable for sensitivity analysis (SA), and uncertainity quantification analysis (UA) of parameters related to a nuclear reactor design. This work also validates the use of linear regression (LR) for predictive analysis in a nuclear reactor design. The analysis helps to determine the parameters on which a LR model can be fit for predictive analysis. For those parameters, a regression surface is created based on trial data and predictions are made using this surface. A general strategy of SA to determine and identify the influential parameters those affect the operation of the reactor is mentioned. Identification of design parameters and validation of linearity assumption for the application of LR of reactor design based on a set of tests is performed. The testing methods used to determine the behavior of the parameters can be used as a general strategy for UA, and SA of nuclear reactor models, and thermal hydraulics calculations. A design of a gas cooled fast breeder reactor (GCFBR), with thermal–hydraulics, and energy transfer has been used for the demonstration of this method. MCNP6 is used to simulate the GCFBR design, and perform the necessary criticality calculations. Java is used to build and run input samples, and to extract data from the output files of MCNP6, and R is used to perform regression analysis and other multivariate variance, and analysis of the collinearity of data

  7. A customized digital monitoring and display system for nonpower reactors

    International Nuclear Information System (INIS)

    Ficaro, E.P.; Wehe, D.K.

    1989-01-01

    A digital data acquisition system for monitoring plant variables has been designed and implemented at the University of Michigan's Ford Nuclear Reactor (FNR), a 2-MW open-pool research reactor. The digital data provided by this system will be useful for: improved operator training, real-time experimental calculations, noise analysis, closed-loop control, and expert system applications. This paper describes the analog-to-digital (A/D) transitions and the associated applications and benefits experienced

  8. PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

    International Nuclear Information System (INIS)

    Bednyakov, Sergey

    2000-12-01

    The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

  9. A numerical technique for reactor subchannel analysis

    International Nuclear Information System (INIS)

    Fath, Hassan E.S.

    1983-01-01

    A numerical technique is developed for the solution of the transient boundary layer equations with a moving liquid-vapour interface boundary. The technique uses the finite difference method with the velocity components defined over an Eulerian mesh. A system of interface massless markers is defined where the markers move with the flow field according to a simple kinematic relation between the interface geometry and the fluid velocity. Different applications of nuclear engineering interest are reported with some available results. The present technique is capable of predicting the interface profile near the wall which is important in the reactor subchannel analysis

  10. Research reactor job analysis - A project description

    International Nuclear Information System (INIS)

    Yoder, John; Bessler, Nancy J.

    1988-01-01

    Addressing the need of the improved training in nuclear industry, nuclear utilities established training program guidelines based on Performance-Based Training (PBT) concepts. The comparison of commercial nuclear power facilities with research and test reactors owned by the U.S. Department of Energy (DOE), made in an independent review of personnel selection, training, and qualification requirements for DOE-owned reactors pointed out that the complexity of the most critical tasks in research reactors is less than that in power reactors. The U.S. Department of Energy (DOE) started a project by commissioning Oak Ridge Associated Universities (ORAU) to conduct a job analysis survey of representative research reactor facilities. The output of the project consists of two publications: Volume 1 - Research Reactor Job Analysis: Overview, which contains an Introduction, Project Description, Project Methodology,, and. An Overview of Performance-Based Training (PBT); and Volume 2 - Research Reactor Job Analysis: Implementation, which contains Guidelines for Application of Preliminary Task Lists and Preliminary Task Lists for Reactor Operators and Supervisory Reactor Operators

  11. Analysis of proposed gamma-ray detection system for the monitoring of core water inventory in a pressurized water reactor

    International Nuclear Information System (INIS)

    Markoff, D.M.

    1987-12-01

    An initial study has been performed of the feasibility of employing an axial array of gamma detectors located outside the pressure vessel to monitor the coolant in a PWR. A one-dimensional transport analysis model is developed for the LOFT research reactor and for a mock-PWR geometry. The gamma detector response to coolant voiding in the core and downcomer has been determined for both geometries. The effects of various conditions (for example, time after shutdown, materials in the transport path, and the relative void fraction in different water regions) on the detector response are studied. The calculational results have been validated by a favorable comparison with LOFT experimental data. Within the limitations and approximations considered in the analysis, the results indicate that the gamma-ray detection scheme is able to unambiguously respond to changes in the coolant inventory within any vessel water region

  12. Nuclear reactor auxiliary heat removal system

    International Nuclear Information System (INIS)

    Thompson, R.E.; Pierce, B.L.

    1977-01-01

    An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

  13. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung-Kyu, E-mail: power@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Kim, Kwangmin; Park, Minwon [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Yu, In-Keun, E-mail: yuik@changwon.ac.kr [Changwon National University, 55306 Sarim-dong, Changwon 641-773 (Korea, Republic of); Lee, Sangjin [Uiduk University, Gyeongju 780-713 (Korea, Republic of)

    2015-11-15

    Highlights: • A 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC transmission system. • The 400 mH class HTS DC reactor was connected to real power network via the HVDC system. • The DC current flowed in HTS DC reactor has several harmonic components and it was analyzed using FFT. - Abstract: High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  14. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    International Nuclear Information System (INIS)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-01-01

    Highlights: • A 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC transmission system. • The 400 mH class HTS DC reactor was connected to real power network via the HVDC system. • The DC current flowed in HTS DC reactor has several harmonic components and it was analyzed using FFT. - Abstract: High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  15. Operating characteristic analysis of a 400 mH class HTS DC reactor in connection with a laboratory scale LCC type HVDC system

    Science.gov (United States)

    Kim, Sung-Kyu; Kim, Kwangmin; Park, Minwon; Yu, In-Keun; Lee, Sangjin

    2015-11-01

    High temperature superconducting (HTS) devices are being developed due to their advantages. Most line commutated converter based high voltage direct current (HVDC) transmission systems for long-distance transmission require large inductance of DC reactor; however, generally, copper-based reactors cause a lot of electrical losses during the system operation. This is driving researchers to develop a new type of DC reactor using HTS wire. The authors have developed a 400 mH class HTS DC reactor and a laboratory scale test-bed for line-commutated converter type HVDC system and applied the HTS DC reactor to the HVDC system to investigate their operating characteristics. The 400 mH class HTS DC reactor is designed using a toroid type magnet. The HVDC system is designed in the form of a mono-pole system with thyristor-based 12-pulse power converters. In this paper, the investigation results of the HTS DC reactor in connection with the HVDC system are described. The operating characteristics of the HTS DC reactor are analyzed under various operating conditions of the system. Through the results, applicability of an HTS DC reactor in an HVDC system is discussed in detail.

  16. Advanced thermionic reactor systems design code

    International Nuclear Information System (INIS)

    Lewis, B.R.; Pawlowski, R.A.; Greek, K.J.; Klein, A.C.

    1991-01-01

    An overall systems design code is under development to model an advanced in-core thermionic nuclear reactor system for space applications at power levels of 10 to 50 kWe. The design code is written in an object-oriented programming environment that allows the use of a series of design modules, each of which is responsible for the determination of specific system parameters. The code modules include a neutronics and core criticality module, a core thermal hydraulics module, a thermionic fuel element performance module, a radiation shielding module, a module for waste heat transfer and rejection, and modules for power conditioning and control. The neutronics and core criticality module determines critical core size, core lifetime, and shutdown margins using the criticality calculation capability of the Monte Carlo Neutron and Photon Transport Code System (MCNP). The remaining modules utilize results of the MCNP analysis along with FORTRAN programming to predict the overall system performance

  17. Expert system for fast reactor diagnostic

    International Nuclear Information System (INIS)

    Parcy, J.P.

    1982-09-01

    A general description of expert systems is given. The operation of a fast reactor is reviewed. The expert system to the diagnosis of breakdowns limited to the reactor core. The structure of the system is described: specification of the diagnostics; structure of the data bank and evaluation of the rules; specification of the prediagnostics and evaluation; explanation of the diagnostics; time evolution of the system; comparison with other expert systems. Applications to some cases of faults are finally presented [fr

  18. Accident analysis for PRC-II reactor

    International Nuclear Information System (INIS)

    Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng

    1997-12-01

    The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward

  19. Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

    International Nuclear Information System (INIS)

    Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

    2007-01-01

    the heat transfer to the water splitting cycle. A preliminary exergy analysis of the proposed heat pump is conducted and a coefficient of performance (COP), taking into account a decrease in electricity generation in the nuclear power generation cycle, is evaluated. The calculated per unit mass flow heat supply rate to the thermochemical cycle increases by a factor of 3 to 5 depending on the steam temperature. A combined system comprising a SCW nuclear plant, a chemical heat pump and a lower temperature UT-3 thermochemical water splitting cycle is presented. Despite a decrease in electricity generation, the calculated exergy efficiency of hydrogen production in the considered system appears to be competitive with that for low temperature water electrolysis

  20. Reactor accident analysis and evaluation

    International Nuclear Information System (INIS)

    Chang, J.W.

    1983-01-01

    Reactor Management Division of Korea Advanced Energy Research Institute has, so far, adopted, modified and developed quite a number of large programs for nuclear core analysis. During the course of this work, it was found necessary to employ some standard subroutines for handling data, input procedures, core memory management and search files. Many programs share lots of common subroutines and/or functions with other programs. Above all, some of them are in lack of transmittal. During the installation of big codes for CYBER computer, it has drawn our keen attention that many elementary subroutines are heavily machine-dependent and that their conversion is extremely difficult. After having collected and modified the subroutines to fit in different codes, it was finally named KINEP (KAERI Improved Nuclear Environmental Package). KINEP has been proved to be convenient even for smaller programs for general purpose. The KINEP includes about one hundred subroutines to facilitate data handling, operator communications, storage allocation, decimal input, file maintence and scratch I/O. (Author)

  1. PWR systems transient analysis

    International Nuclear Information System (INIS)

    Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

    1985-01-01

    Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

  2. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  3. Innovative inspection system for reactor pressure vessels

    International Nuclear Information System (INIS)

    Mertens, K.; Trautmann, H.

    1999-01-01

    The versatile, compact and modern underwater systems described, the DELPHIN manipulators and MIDAS submarines, are innovative systems enabling RPV inspections at considerably reduced efforts and time, thus reducing the total time required for ISI of reactors. (orig./CB) [de

  4. Sensitivity analysis on the effect of software-induced common cause failure probability in the computer-based reactor trip system unavailability

    International Nuclear Information System (INIS)

    Kamyab, Shahabeddin; Nematollahi, Mohammadreza; Shafiee, Golnoush

    2013-01-01

    Highlights: ► Importance and sensitivity analysis has been performed for a digitized reactor trip system. ► The results show acceptable trip unavailability, for software failure probabilities below 1E −4 . ► However, the value of Fussell–Vesley indicates that software common cause failure is still risk significant. ► Diversity and effective test is founded beneficial to reduce software contribution. - Abstract: The reactor trip system has been digitized in advanced nuclear power plants, since the programmable nature of computer based systems has a number of advantages over non-programmable systems. However, software is still vulnerable to common cause failure (CCF). Residual software faults represent a CCF concern, which threat the implemented achievements. This study attempts to assess the effectiveness of so-called defensive strategies against software CCF with respect to reliability. Sensitivity analysis has been performed by re-quantifying the models upon changing the software failure probability. Importance measures then have been estimated in order to reveal the specific contribution of software CCF in the trip failure probability. The results reveal the importance and effectiveness of signal and software diversity as applicable strategies to ameliorate inefficiencies due to software CCF in the reactor trip system (RTS). No significant change has been observed in the rate of RTS failure probability for the basic software CCF greater than 1 × 10 −4 . However, the related Fussell–Vesley has been greater than 0.005, for the lower values. The study concludes that consideration of risk associated with the software based systems is a multi-variant function which requires compromising among them in more precise and comprehensive studies

  5. Application of autoregressive moving average model in reactor noise analysis

    International Nuclear Information System (INIS)

    Tran Dinh Tri

    1993-01-01

    The application of an autoregressive (AR) model to estimating noise measurements has achieved many successes in reactor noise analysis in the last ten years. The physical processes that take place in the nuclear reactor, however, are described by an autoregressive moving average (ARMA) model rather than by an AR model. Consequently more correct results could be obtained by applying the ARMA model instead of the AR model to reactor noise analysis. In this paper the system of the generalised Yule-Walker equations is derived from the equation of an ARMA model, then a method for its solution is given. Numerical results show the applications of the method proposed. (author)

  6. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  7. DEPTH-CHARGE static and time-dependent perturbation/sensitivity system for nuclear reactor core analysis. Revision I

    International Nuclear Information System (INIS)

    White, J.R.

    1985-04-01

    This report provides the background theory, user input, and sample problems required for the efficient application of the DEPTH-CHARGE system - a code black for both static and time-dependent perturbation theory and data sensitivity analyses. The DEPTH-CHARGE system is of modular construction and has been implemented within the VENTURE-BURNER computational system at Oak Ridge National Laboratory. The DEPTH module (coupled with VENTURE) solves for the three adjoint functions of Depletion Perturbation Theory and calculates the desired time-dependent derivatives of the response with respect to the nuclide concentrations and nuclear data utilized in the reference model. The CHARGE code is a collection of utility routines for general data manipulation and input preparation and considerably extends the usefulness of the system through the automatic generation of adjoint sources, estimated perturbed responses, and relative data sensitivity coefficients. Combined, the DEPTH-CHARGE system provides, for the first time, a complete generalized first-order perturbation/sensitivity theory capability for both static and time-dependent analyses of realistic multidimensional reactor models. This current documentation incorporates minor revisions to the original DEPTH-CHARGE documentation (ORNL/CSD-78) to reflect some new capabilities within the individual codes

  8. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  9. Preliminary conceptual design and analysis on KALIMER reactor structures

    International Nuclear Information System (INIS)

    Kim, Jong Bum

    1996-10-01

    The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

  10. 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

  11. Reactor core design aiding system

    International Nuclear Information System (INIS)

    Kanazawa, Nobuhiro; Hamaguchi, Yukio; Nakao, Takashi; Kondo, Yasuhide

    1995-01-01

    A two-dimensional radial power distribution and an axial one-dimensional power distribution are determined based on a distribution of a three-dimensional infinite multiplication factor, to obtain estimated power distribution estimation values. The estimation values are synthesized to obtain estimated three-dimensional power distribution values. In addition, the distribution of a two-dimensional radial multiplication factor and the distribution of an one-dimensional axial multiplication factor are determined based on the three-dimensional power distribution, to obtain estimated values for the multiplication factor distribution. The estimated values are synthesized to form estimated values for the three-dimensional multiplication factor distribution. Further, estimated fuel loading pattern value is determined based on the three-dimensional power distribution or the two-dimensional radial power distribution. Since the processes for determining the estimated values comprise only additive and multiplying operations, processing time can be remarkably saved compared with calculation based on a detailed physical models. Since the estimation is performed on every fuel assemblies, a nervous circuit network not depending on the reactor core system can be constituted. (N.H.)

  12. Heat transfer capability analysis of heat pipe for space reactor

    International Nuclear Information System (INIS)

    Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang

    2015-01-01

    To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)

  13. Thermo-mechanical analysis of an acceleration grid for the international thermonuclear experimental reactor-neutral beam injection system

    International Nuclear Information System (INIS)

    Fujiwara, Yukio; Hanada, Masaya; Okumura, Yoshikazu; Suzuki, Satoshi; Watanabe, Kazuhiro

    2001-01-01

    In the engineering design of a negative-ion beam source for a high-power neutral beam injection (NBI) system, one of the most important issues is thermo-mechanical design of acceleration grids for producing several tens of MW ion beams. An acceleration grid for the international thermonuclear experimental reactor-neutral beam injection (ITER-NBI) system will be subjected to the heat loading as high as 1.5 MW. In the present paper, thermo-mechanical characteristics of the acceleration grid for the ITER-NBI system were analyzed. Numerical simulation indicated that maximum aperture-axis displacement of the acceleration grid due to thermal expansion would be about 0.7 mm for the heat loading of 1.5 MW. From the thin lens theory of beam optics, beamlet deflection angle by the aperture-axis displacement was estimated to be about 2 mrad, which is within the requirement of the engineering design of the ITER-NBI system. Numerical simulation also indicated that no melting on the acceleration grid would occur for a heat loading of 1.5 MW, while local plastic deformation would happen. To avoid the plastic deformation, it is necessary to reduce the heat loading onto the acceleration grid to less than 1 MW

  14. 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)

  15. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Owen, P.S.; Parker, M.B.; Omberg, R.P.

    1979-05-01

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U 3 O 8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  16. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  17. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  18. Thermal hydraulics analysis of the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Dean, E-mail: Dean_Wang@uml.edu [University of Massachusetts Lowell, One University Avenue, Lowell, MA 01854 (United States); Yoder, Graydon L.; Pointer, David W.; Holcomb, David E. [Oak Ridge National Laboratory, 1 Bethel Valley RD #6167, Oak Ridge, TN 37831 (United States)

    2015-12-01

    Highlights: • The TRACE AHTR model was developed and used to define and size the DRACS and the PHX. • A LOFF transient was simulated to evaluate the reactor performance during the transient. • Some recommendations for modifying FHR reactor system component designs are discussed. - Abstract: The Advanced High Temperature Reactor (AHTR) is a liquid salt-cooled nuclear reactor design concept, featuring low-pressure molten fluoride salt coolant, a carbon composite fuel form with embedded coated particle fuel, passively triggered negative reactivity insertion mechanisms, and fully passive decay heat rejection. This paper describes an AHTR system model developed using the Nuclear Regulatory Commission (NRC) thermal hydraulic transient code TRAC/RELAP Advanced Computational Engine (TRACE). The TRACE model includes all of the primary components: the core, downcomer, hot legs, cold legs, pumps, direct reactor auxiliary cooling system (DRACS), the primary heat exchangers (PHXs), etc. The TRACE model was used to help define and size systems such as the DRACS and the PHX. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. Some initial recommendations for modifying system component designs are also discussed. The TRACE model will be used as the basis for developing more detailed designs and ultimately will be used to perform transient safety analysis for the reactor.

  19. Passive Decay Heat Removal System for Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Jangsik; Lee, Jeong Ik; Jeong, Yong Hoon [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Dry cooling system is applied as waste heat removal system therefore it is able to consider wide construction site. Schematic figure of the reactor is shown in Fig. 1. In safety features, the reactor has double containment and passive decay heat removal (PDHR) system. The double containment prevents leakage from reactor coolant system to be emitted into environment. The passive decay heat removal system copes with design basis accidents (DBAs). Micros Modular Reactor (MMR) which has been being developed in KAIST is S-CO{sub 2} gas cooled reactor and shows many advantages. The S-CO{sub 2} power cycle reduces size of compressor, and it makes small size of power plant enough to be transported by trailer.The passive residual heat removal system is designed and thermal hydraulic (TH) analysis on coolant system is accomplished. In this research, the design process and TH analysis results are presented. PDHR system is designed for MMR and coolant system with the PDHR system is analyzed by MARS-KS code. Conservative assumptions are applied and the results show that PDHR system keeps coolant system under the design limitation.

  20. Seismic attenuation system for a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liszkai, Tamas; Cadell, Seth

    2018-01-30

    A system for attenuating seismic forces includes a reactor pressure vessel containing nuclear fuel and a containment vessel that houses the reactor pressure vessel. Both the reactor pressure vessel and the containment vessel include a bottom head. Additionally, the system includes a base support to contact a support surface on which the containment vessel is positioned in a substantially vertical orientation. An attenuation device is located between the bottom head of the reactor pressure vessel and the bottom head of the containment vessel. Seismic forces that travel from the base support to the reactor pressure vessel via the containment vessel are attenuated by the attenuation device in a direction that is substantially lateral to the vertical orientation of the containment vessel.

  1. Gaseous fuel reactors for power systems

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  2. Human Reliability Analysis for Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ronald L. Boring; David I. Gertman

    2012-06-01

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  3. All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

    International Nuclear Information System (INIS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

    2009-01-01

    Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

  4. Activation analysis with small mobile reactors

    International Nuclear Information System (INIS)

    Chung, C.

    1990-01-01

    A small nuclear reactor (a low-power reactor without heat removal devices) usually has thermal power output under 100 W and an average in-core thermal neutron flux below 10 9 n/cm 2 s. Conventional activation analysis is restricted to determination of specific elements with large neutron capture cross sections in sizable samples. In-vivo prompt gamma activation analysis (IVPGAA) can be used for diagnosis of elemental composition of the human body, particularly the essential elements Ca, Cl, N, and P in the whole body, and toxic Cd and Hg in contaminated organs. In this chapter, activation analysis using an external neutron beam from the Tsing Hua Mobile Educational Reactor (THMER) for in vivo activation is described. Characteristics of the mobile reactor, in-vivo medical diagnosis, and radiation safety are emphasized. 17 refs, 12 figs, 3 tabs

  5. Development of a sensitivity analysis systems in nuclear reactors through generalized perturbation theory at first order in 2 D geometries

    International Nuclear Information System (INIS)

    Garcia, Juan Matias

    2005-01-01

    Perturbation Methods represent a powerful tool to do sensitivity analysis, and they found many aplications in nuclear engineering.As an introduction to this kind of analysis, we develope a program that apply the Generalized Perturbation Theory or GPT Method to bidimensional system of rectangular geometry.We first consider an homogeneous system of non-multiplying material and then an heterogeneous system with region of multiplying material, with the intention of make concret aplications of perturbation method to nuclear engineering problems.The program, that we called Pert, determines neutron fluxes and importance functions applying the Multigroup Diffusion Theory; and also solves the integrals required to calculate sensitivity coefficients.Using this perturbation methods we could verify the low computational cost required to make this kind of analysis and the simplicity of the equations systems involved, allowing us to make elaborates sensitivity analysis for the responses of our interest

  6. Criteria for the CAREM reactor's expert system design conduction

    International Nuclear Information System (INIS)

    Furman, A.; Delgado, R.

    1990-01-01

    The present work describes the analysis made to start with the development of an Expert System for the CAREM (SE) reactor's conduction. The following tasks are presented: a) purpose of the Expert System; b) Decision Making structure; c) Architecture of the Expert System; d) Description of Subsystems and e) Licensing. (Author) [es

  7. Stack Monitoring System At PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Zamrul Faizad Omar; Mohd Sabri Minhat; Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Izhar Abu Hussin

    2014-01-01

    This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

  8. Sampling system for a boiling reactor NPP

    International Nuclear Information System (INIS)

    Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

    1976-01-01

    Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

  9. ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

    Directory of Open Access Journals (Sweden)

    Alexander Labutin

    2017-02-01

    Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

  10. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  11. 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.

  12. Analysis of water cooled reactors stability; Analiza stabilnosti reaktorskih sistema hladjenih vodom

    Energy Technology Data Exchange (ETDEWEB)

    Marinkovic, P; Pesic, M [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1980-07-01

    A model for stability analysis of non-boiling water cooled nuclear system is developed. The model is based on linear reactor kinetics and space averaged heat transfer in reactor and heat-exchanger. The transfer functions are defined and the analysis was applied to nuclear reactor RA at 'Boris Kidric' Institute - Vinca. (author)

  13. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  14. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    International Nuclear Information System (INIS)

    Neil Todreas; Pavel Hejzlar

    2008-01-01

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

  15. System for unattended surveillance of nuclear reactor behavior

    International Nuclear Information System (INIS)

    Gonzalez, R.C.; Howington, L.C.

    1977-01-01

    A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and dimensionality reduction capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns and to recognize deviations from these patterns were evaluated by experiments at the ORNL High-Flux Isotope Reactor. Power perturbations of less than 0.1% of the mean value in selected frequency ranges were readily detected by the pattern recognition system

  16. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  17. Reactor core operation management system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Tomomi.

    1992-05-28

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

  18. Reactor core operation management system

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1992-01-01

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

  19. Cooling system upon reactor isolation

    International Nuclear Information System (INIS)

    Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

    1992-01-01

    A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

  20. Dynamic operator actions analysis for inherently safe fast reactors and light water reactors

    International Nuclear Information System (INIS)

    Ho, V.; Apostolakis, G.

    1988-01-01

    A comparative dynamic human actions analysis of inherently safe fast reactors (ISFRs) and light water reactors (LWRs) in terms of systems response and estimated human error rates is presented. Brief overviews of the ISFR and LWR systems are given to illustrate the design differences. Key operator actions required by the ISFR reactor shutdown and decay heat removal systems are identified and are compared with those of the LWR. It is observed that, because of the passive nature of the ISFR safety-related systems, a large time window is available for operator actions during transient events. Furthermore, these actions are fewer in number, are less complex, and have lower error rates and less severe consequences than those of the LWRs. We expect the ISFR operator errors' contribution to risk is smaller (at least in the context of the existing human reliability models) than that of the LWRs. (author)

  1. Finite element application to global reactor analysis

    International Nuclear Information System (INIS)

    Schmidt, F.A.R.

    1981-01-01

    The Finite Element Method is described as a Coarse Mesh Method with general basis and trial functions. Various consequences concerning programming and application of Finite Element Methods in reactor physics are drawn. One of the conclusions is that the Finite Element Method is a valuable tool in solving global reactor analysis problems. However, problems which can be described by rectangular boxes still can be solved with special coarse mesh programs more efficiently. (orig.) [de

  2. EMERIS: an advanced information system for a materials testing reactor

    International Nuclear Information System (INIS)

    Adorjan, F.; Buerger, L.; Lux, I.; Mesko, L.; Szabo, K.; Vegh, J.; Ivanov, V.V.; Mozhaev, A.A.; Yakovlev, V.V.

    1990-06-01

    The basic features of the Materials Testing Reactor of IAE, Moscow (MR) Information System (EMERIS) are outlined. The purpose of the system is to support reactor and experimental test loop operators by a flexible, fully computerized and user-friendly tool for the aquisition, analysis, archivation and presentation of data obtained during operation of the experimental facility. High availability of EMERIS services is ensured by redundant hardware and software components, and by automatic configuration procedure. A novel software feature of the system is the automatic Disturbance Analysis package, which is aimed to discover primary causes of irregularities occurred in the technology. (author) 2 refs.; 2 figs

  3. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  4. Reliabitity study of the accumulator system for Angra-1 reactor

    International Nuclear Information System (INIS)

    Santos Maciel, C.C.R.

    1980-01-01

    The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

  5. Study, design and evaluation of nuclear reactor computer control system

    International Nuclear Information System (INIS)

    Menacer, S.

    1988-01-01

    Nuclear reactor control is a complex process that varies with each reactor and there is no universal agreement as to the best type of control system. After the use of conventional systems for a long time, attention turned towards digital techniques in the reactor control system. This interest emerged because of the difficulties faced in the data manipulation, mainly for post-incident analysis. However, it is not sufficient to insert a computer in a system to solve all the data-handling problems and also the insertion of a computer in a real-time system is not without any effect on the overall system. The scope of this thesis is to show the important parameters that have to be taken into account when choosing and evaluate the performances of the selected system

  6. Technical Readiness and Gaps Analysis of Commercial Optical Materials and Measurement Systems for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anheier, Norman C.; Suter, Jonathan D.; Qiao, Hong (Amy); Andersen, Eric S.; Berglin, Eric J.; Bliss, Mary; Cannon, Bret D.; Devanathan, Ramaswami; Mendoza, Albert; Sheen, David M.

    2013-08-06

    This report intends to support Department of Energy’s Office of Nuclear Energy (DOE-NE) Nuclear Energy Research and Development Roadmap and industry stakeholders by evaluating optical-based instrumentation and control (I&C) concepts for advanced small modular reactor (AdvSMR) applications. These advanced designs will require innovative thinking in terms of engineering approaches, materials integration, and I&C concepts to realize their eventual viability and deployability. The primary goals of this report include: 1. Establish preliminary I&C needs, performance requirements, and possible gaps for AdvSMR designs based on best available published design data. 2. Document commercial off-the-shelf (COTS) optical sensors, components, and materials in terms of their technical readiness to support essential AdvSMR in-vessel I&C systems. 3. Identify technology gaps by comparing the in-vessel monitoring requirements and environmental constraints to COTS optical sensor and materials performance specifications. 4. Outline a future research, development, and demonstration (RD&D) program plan that addresses these gaps and develops optical-based I&C systems that enhance the viability of future AdvSMR designs. The development of clean, affordable, safe, and proliferation-resistant nuclear power is a key goal that is documented in the Nuclear Energy Research and Development Roadmap. This roadmap outlines RD&D activities intended to overcome technical, economic, and other barriers, which currently limit advances in nuclear energy. These activities will ensure that nuclear energy remains a viable component to this nation’s energy security.

  7. Recycling systems for BWR type reactors

    International Nuclear Information System (INIS)

    Takagi, Akio; Yamamoto, Fumiaki; Fukumoto, Ryuji.

    1986-01-01

    Purpose: To stabilize the coolant flowing characteristics and reactor core reactivity. Constitution: The recycling system in a BWR type reactor comprises a recycling pump disposed to the outside of a reactor pressure vessel, a ring header connected to the recycling pump through main pipe ways, and a plurality of pipes branched from and connected with the ring header and connected to a plurality of jet pumps within the pressure vessel. Then, by making the diameter for the pipeways of each of the branched pipes different from each other, the effective cross-sectional area is varied to thereby average the coolant flow rate supplied to each of the jet pumps. (Seki, T.)

  8. Study on effects of development of reactor constant in fast reactor analysis

    International Nuclear Information System (INIS)

    Chiba, Gou

    2002-12-01

    Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

  9. Probabilistic safety analysis applied to RBMK reactors

    International Nuclear Information System (INIS)

    Gerez Martin, L.; Fernandez Ramos, P.

    1995-01-01

    The project financed by the European Union ''Revision of RBMK Reactor Safety was divided into nine Topic Groups dealing with different aspects of safety. The area covered by Topic Group 9 was Probabilistic Safety Analysis. TG9 will have touched on some of the problems discussed by other groups, although in terms of the systematic quantification of the impact of design characteristics and RBMK reactor operating practices on the risk of core damage. On account of the reduced time scale and the resources available for the project, the analysis was made using a simplified method based on the results of PSAs conducted in Western countries and on the judgement of the group members. The simplifies method is based on the concepts of Qualification, Redundancy and Automatic Actuation of the systems considered. PSA experience shows that systems complying with the above-mentioned concepts have a failure probability of 1.0E-3 when redundancy is simple, ie two similar equipment items capable of carrying out the same function. In general terms, this value can be considered to be dominated by potential common cause failures. The value considered above changes according to factors that have a positive effect upon it, such as an additional redundancy with a different equipment item (eg a turbo pumps and a motor pump), individual trains with good separations, etc, or a negative effect, such as the absence of suitable periodical tests, the need for operators to perform manual operations, etc. Similarly, possible actions required by the operator during accident sequences are assigned failure probability values between 1 and 1.0E-4, according to the complexity of the action (including local actions to be performed outside the control room) and the time available

  10. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Iwasaki, Takashi; Sakai, Takaaki; Enuma, Yasuhiro; Mizuno, Tomoyasu

    2002-03-01

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

  11. Gaseous fuel reactors for power systems

    Science.gov (United States)

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

    1977-01-01

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

  12. Nuclear reactor machine refuelling system

    International Nuclear Information System (INIS)

    Cashen, W.S.; Erwin, D.

    1977-01-01

    Part of an on-line fuelling machine for a CANDU pressure-tube reactor is described. The present invention provides a refuelling machine wherein the fuelling components, including the fuel carrier and the closure adapter, are positively positioned and retained within the machine magazine or positively secured to the machine charge tube head, and cannot be accidentally disengaged as in former practice. The positive positioning devices include an arcuate keeper plate. Simplified hooked fingers are used. (NDH)

  13. Reactor vessel stud closure system

    International Nuclear Information System (INIS)

    Spiegelman, S.R.; Salton, R.B.; Beer, R.W.; Malandra, L.J.; Cognevich, M.L.

    1982-01-01

    A quick-acting stud tensioner apparatus for enabling the loosening or tightening of a stud nut on a reactor vessel stud. The apparatus is adapted to engage the vessel stud by closing a gripper around an upper end of the vessel stud when the apparatus is seated on the stud. Upon lifting the apparatus, the gripper releases the vessel stud so that the apparatus can be removed

  14. Progress of electromagnetic analysis for fusion reactors

    International Nuclear Information System (INIS)

    Takagi, T.; Ruatto, P.; Boccaccini, L.V.

    1998-01-01

    This paper describes the recent progress of electromagnetic analysis research for fusion reactors including methods, codes, verification tests and some applications. Due to the necessity of the research effort for the structural design of large tokamak devices since the 1970's with the help of the introduction of new numerical methods and the advancement of computer technologies, three-dimensional analysis methods have become as practical as shell approximation methods. The electromagnetic analysis is now applied to the structural design of new fusion reactors. Some more modeling and verification tests are necessary when the codes are applied to new materials with nonlinear material properties. (orig.)

  15. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  16. Nuclear reactor system for ABWR

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Kitagawa, Koji

    1997-01-01

    Various tests and measurements were performed during the pre-operational test run of Unit No. 6 of The Tokyo Electric Power Co., Inc.'s Kashiwazaki-Kariwa Nuclear Power Station, the first advanced boiling water reactor (ABWR) unit in the world, and the design and performance adequacy of the ABWR were confirmed. The realization of the ABWR in Japan took about 20 years. It was decided that technologies for the reactor internal pump (RIP) and the fine-motion control rod drive (FMCRD), which had been applied in Europe, would be incorporated in the ABWR aiming at simplification of its structure and operation. These main components were evaluated, modified and verified in consideration of the unique Japanese environment, such as seismic conditions, through a joint study program with Japanese utilities as well as an improvement and standardization program in cooperation with the government. In addition to incorporating RIP and FMCRD technologies, the ABWR also has improved features in terms of the design of the reactor pressure vessel and internals, as well as automated servicing equipment for the RIP, FMCRD, and primary containment vessel. (author)

  17. Safety-oriented global analysis of reactor dynamics

    International Nuclear Information System (INIS)

    Belhadj, M.; Aldemir, T.

    1992-01-01

    It is well known that the asymptotic solutions of the non-linear systems encountered in reactor dynamics can change from stable to periodic or from periodic to chaotic with a very small change in system parameters and/or initial conditions. In that respect, determination of the domains of attraction (DOAs) in the state-space that contains the asymptotic solutions and the identification of the basins of attraction (BOAs) and lead to these DOAs usually requires a global analysis of reactor dynamics (as opposed to a local analysis through perturbation theory). From the standpoint of safety, the DOAs indicate whether the reactor behavior remains within the imposed constraints or not, and the BOAs show which initial conditions lead to safe operation. Due to the lack of a general theory, often the only feasible method for the global analysis of nonlinear systems is the direct integration of governing equations. However, direct integration can be computationally prohibitive, particularly if there is uncertainty on the values of the system parameters to be used in the analysis, and/or asymptotic system behavior is chaotic. In a recent study, a global analysis algorithm was presented to determine the structure of DOAs (and their probability distribution when there is uncertainty on the system parameters) more quickly than by direct integration. This paper shows how the new algorithm can be expanded to determine the BOAs of reactor dynamics equations as well as their DOAs

  18. Pressurizer pump reliability analysis high flux isotope reactor

    International Nuclear Information System (INIS)

    Merryman, L.; Christie, B.

    1993-01-01

    During a prolonged outage from November 1986 to May 1990, numerous changes were made at the High Flux Isotope Reactor (HFIR). Some of these changes involved the pressurizer pumps. An analysis was performed to calculate the impact of these changes on the pressurizer system availability. The analysis showed that the availability of the pressurizer system dropped from essentially 100% to approximately 96%. The primary reason for the decrease in availability comes because off-site power grid disturbances sometimes result in a reactor trip with the present pressurizer pump configuration. Changes are being made to the present pressurizer pump configuration to regain some of the lost availability

  19. Breeder reactor fuel fabrication system development

    International Nuclear Information System (INIS)

    Bennett, D.W.; Fritz, R.L.; McLemore, D.R.; Yatabe, J.M.

    1981-01-01

    Significant progress has been made in the design and development of remotely operated breeder reactor fuel fabrication and support systems (e.g., analytical chemistry). These activities are focused by the Secure Automated Fabrication (SAF) Program sponsored by the Department of Energy to provide: a reliable supply of fuel pins to support US liquid metal cooled breeder reactors and at the same time demonstrate the fabrication of mixed uranium/plutonium fuel by remotely operated and automated methods

  20. Renewal of reactor cooling system of JMTR. Reactor building site

    International Nuclear Information System (INIS)

    Onoue, Ryuji; Kawamata, Takanori; Otsuka, Kaoru; Sekine, Katsunori; Koike, Sumio; Gorai, Shigeru; Nishiyama, Yutaka; Fukasaku, Akitomi

    2012-03-01

    The Japan Materials Testing Reactor (JMTR) is a light water moderated and cooled tank-type reactor, and its thermal power is 50 MW. The JMTR is categorized as high flux testing reactors in the world. The JMTR has been utilized for irradiation experiments of nuclear fuels and materials, as well as for radioisotope productions since the first criticality in March 1968 until August 2006. JAEA is decided to refurbish the JMTR as an important fundamental infrastructure to promote the nuclear research and development. And The JMTR refurbishment work is carried out for 4 years from 2007. Before refurbishment work, from August 2006 to March 2007, all concerned renewal facilities were selected from evaluation on their damage and wear in terms of aging. Facilities which replacement parts are no longer manufactured or not likely to be manufactured continuously in near future, are selected as renewal ones. Replace priority was decided with special attention to safety concerns. A monitoring of aging condition by the regular maintenance activity is an important factor in selection of continuous using after the restart. In this report, renewal of the cooling system within refurbishment facilities in the JMTR is summarized. (author)

  1. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  2. A program for dynamic noise investigations of reactor systems

    International Nuclear Information System (INIS)

    Antonov, N.A.; Yaneva, N.B.

    1980-01-01

    A stochastic process analysis in nuclear reactors is used for the state diagnosis and dynamic characteristic investigation of the reactor system. A program DENSITY adapted and tested on an IBM 360 ES type computer is developed. The program is adjusted for fast processing of long series exploiting a relatively small memory. The testing procedure is discussed and the method of the periodic sequences corresponding to characteristic reactivity perturbations of the reactor systems is considered. The program is written for calculating the auto-power spectral density and the cross-power spectral density, as well as the coherence function of stationary statistical time series using the advantages of the fast Fourier transformation. In particular, it is shown that the multi-frequency binary sequences are very useful with respect to the signal-to-noise ratio and the frequency distribution in view of the frequency reactor test

  3. Seismic analysis of fast breeder reactor block

    International Nuclear Information System (INIS)

    Gantenbein, F.

    1990-01-01

    Seismic analysis of LMFBR reactor block is complex due mainly to the fluid structure interaction and the 3D geometry of the structure. Analytical methods which have been developed for this analysis will be briefly described in the paper and applications to a geometry similar to SPX1 will be shown

  4. Choice of thermal reactor systems: a report

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    This is a report by the UK National Nuclear Corporation published by the UK Secretary of State for Energy (Mr. Benn) on 29th July 1977. It is concerned with the advantages and disadvantages of three thermal reactor systems -the AGR (advanced gas cooled reactor), the PWR (pressurised water reactor), and the SGHWR (steam generating heavy water reactor). The object was to help in the future choice of a thermal system for the UK to cover the next 25 years. The matter of export potential is also considered. A programme of four stations of 1100 to 1300 MW each over six years starting from 1979 was assumed. It is emphasised that a decision must be taken now both about reactor systems and actual orders. Headings are as follows: Extract from conclusions reached; Summary of main features of assessment; General conclusions regarding the following - safety, security of the investment, operational characteristics, development and launching requirements, effect on industry, and capital and generation costs. It is stated that in order to make an overall judgement on reactor choice the technical, commercial and social issues involved must be weighed in conjunction with cost differentials.

  5. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

    2001-01-01

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  6. Thermodynamic analysis of behaviour of boiling water reactor coolant on the basis of solubility in Fe3O4-H2O-O2 system

    International Nuclear Information System (INIS)

    Zarembo, V.I.; Slobodov, A.A.; Kritskij, V.G.; Puchkov, L.V.; Sedov, V.M.

    1986-01-01

    The thermodynamic analysis of the behaviour of boiling water reactor coolant on the basis of solubility in Fe 3 O 4 -H 2 O-O 2 system is performed for the purpose of establishing the iron existence forms in non-sedimentated suspended corrosion product particles as well as iron concentration of corrosion origin in power plants. It is shown that the iron solubility in the considered system with temperature variation occurs through the maximum at 423 K. Below this temperature the crystal Fe(OH) 3 is responsible for its value, at higher temperatures - magnetite. The growth of equilibrium oxygen concentration from 0.1 to 1000 μg/kg H 2 O only slightly increases the magnetite solubility

  7. Analysis of reactor trips involving balance-of-plant failures

    International Nuclear Information System (INIS)

    Seth, S.; Skinner, L.; Ettlinger, L.; Lay, R.

    1986-01-01

    The relatively high frequency of plant transients leading to reactor trips at nuclear power plants in the US is of economic and safety concern to the industry. A majority of such transients is due to failures in the balance-of-plant (BOP) systems. As a part of a study conducted for the US Nuclear Regulatory Commission, Mitre has carried out a further analysis of the BOP failures associated with reactor trips. The major objectives of the analysis were to examine plant-to-plant variations in BOP-related trips, to understand the causes of failures, and to determine the extent of any associated safety system challenges. The analysis was based on the Licensee Event Reports submitted on all commercial light water reactors during the 2-yr period, 1984-1985

  8. Comparative analysis of operation and safety of subcritical nuclear systems and innovative critical reactors; Analyse comparative du fonctionnement et de la surete de systemes sous-critiques et de reacteurs critiques innovants

    Energy Technology Data Exchange (ETDEWEB)

    Bokov, P.M

    2005-05-01

    The main goal of this thesis work is to investigate the role of core subcriticality for safety enhancement of advanced nuclear systems, in particular, molten salt reactors, devoted to both energy production and waste incineration/transmutation. The inherent safety is considered as ultimate goal of this safety improvement. An attempt to apply a systematic approach for the analysis of the subcriticality contribution to inherent properties of hybrid system was performed. The results of this research prove that in many cases the subcriticality may improve radically the safety characteristics of nuclear reactors, and in some configurations it helps to reach the 'absolute' intrinsic safety. In any case, a proper choice of subcriticality level makes all analyzed transients considerably slower and monotonic. It was shown that the weakest point of the independent-source systems with respect to the intrinsic safety is thermohydraulic unprotected transients, while in the case of the coupled-source systems the excess reactivity/current insertion events remain a matter of concern. To overcome these inherent drawbacks a new principle of realization of a coupled sub-critical system (DENNY concept) is proposed. In addition, the ways to remedy some particular safety-related problems with the help of the core sub-criticality are demonstrated. A preliminary safety analysis of the fast-spectrum molten salt reactor (REBUS concept) is also carried out in this thesis work. Finally, the potential of the alternative (to spallation) neutron sources for application in hybrid systems is examined. (author)

  9. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume I. RELAP4/MOD5 description. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    RELAP4 is a computer program written in FORTRAN IV for the digital computer analysis of nuclear reactors and related systems. It is primarily applied in the study of system transient response to postulated perturbations such as coolant loop rupture, circulation pump failure, power excursions, etc. The program was written to be used for water-cooled (PWR and BWR) reactors and can be used for scale models such as LOFT and SEMISCALE. Additional versatility extends its usefulness to related applications, such as ice condenser and containment subcompartment analysis. Specific options are available for reflood (FLOOD) analysis and for the NRC Evaluation Model.

  10. A powerful methodology for reactor vessel pressurized thermal shock analysis

    International Nuclear Information System (INIS)

    Boucau, J.; Mager, T.

    1994-01-01

    The recent operating experience of the Pressurized Water Reactor (PWR) Industry has focused increasing attention on the issue of reactor vessel pressurized thermal shock (PTS). More specifically, the review of the old WWER-type of reactors (WWER 440/230) has indicated a sensitive behaviour to neutron embrittlement. This led already to some remedial actions including safety injection water preheating or vessel annealing. Such measures are usually taken based on the analysis of a selected number of conservative PTS events. Consideration of all postulated cooldown events would draw attention to the impact of operator action and control system effects on reactor vessel PTS. Westinghouse has developed a methodology which couples event sequence analysis with probabilistic fracture mechanics analyses, to identify those events that are of primary concern for reactor vessel integrity. Operating experience is utilized to aid in defining the appropriate event sequences and event frequencies of occurrence for the evaluation. Once the event sequences of concern are identified, detailed deterministic thermal-hydraulic and structural evaluations can be performed to determine the conditions required to minimize the extension of postulated flaws or enhance flaw arrest in the reactor vessel. The results of these analyses can then be used to better define further modifications in vessel and plant system design and to operating procedures. The purpose of the present paper will be to describe this methodology and to show its benefits for decision making. (author). 1 ref., 3 figs

  11. Function analysis of nuclear power plants for developing of man-machine interface system for Korean next generation reactor