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Sample records for reactors transients identification

  1. Nuclear reactors transients identification and classification system

    International Nuclear Information System (INIS)

    Bianchi, Paulo Henrique

    2008-01-01

    This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

  2. Nuclear reactors transients identification and classification system; Sistema de identificacao e classificacao de transientes em reatores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Bianchi, Paulo Henrique

    2008-07-01

    This work describes the study and test of a system capable to identify and classify transients in thermo-hydraulic systems, using a neural network technique of the self-organizing maps (SOM) type, with the objective of implanting it on the new generations of nuclear reactors. The technique developed in this work consists on the use of multiple networks to do the classification and identification of the transient states, being each network a specialist at one respective transient of the system, that compete with each other using the quantization error, that is a measure given by this type of neural network. This technique showed very promising characteristics that allow the development of new functionalities in future projects. One of these characteristics consists on the potential of each network, besides responding what transient is in course, could give additional information about that transient. (author)

  3. Development of a transient identification and classification system to an integral pressurized water reactor

    International Nuclear Information System (INIS)

    Aronne, Ivan Dionysio

    2009-01-01

    The demand for energy in the modern world is growing, particularly in the developing countries. The nuclear options has been deserving prominence for their qualities of not impacting the environment through emissions of greenhouse gases and nor to demand great areas.. However society requests improvement in the safety of new reactors and the utilities request larger availability of the power plants. The IRIS project of an integral nuclear pressurized water reactor proposes to fulfill those requirements. A system for identification and classification of transients would help to improve the safety and to increase the availability of the IRIS increasing its competitiveness. In order to contribute to the development o such a system it was developed in this work a System for Identification and Classification of Transients - SDICT - capable of monitoring the operation of the reactor and of providing information on its operational state. SICT was developed using the technique of neural networks, more specifically the Self-Organizing Maps. Results of IRIS simulation with RELAP5 code were used to train the neural network of SICT. To demonstrate the correctness of the methodology of using simulations results, whose values have characteristics different from the measured ones, it was made a version of SICT for an experimental installation, the The Circuit no. 1 - CT1. Experiments were run in this test facility and simulations of its operation were done with RELAP5. This CT1 version of SICT was then checked against the simulation and experimental data validating the methodology. As a result of the activities to develop SICT, it is now available for futures studies: the developed application, SICT, a database of experiments in CT1, a validate nodalization of CT1, a database of results of CT1 simulations , a nodalization of the IRIS tested for several normal and abnormal transients and a database with the results of IRIS simulations. Attached to this thesis is a CD with the

  4. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

    Gaber, F.A.; El Messiry, A.M.

    1988-01-01

    This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors

  5. 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)

  6. Reactor transients tests for SNR fuel elements in HFR reactor

    International Nuclear Information System (INIS)

    Plitz, H.

    1989-01-01

    In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

  7. The economic impact of reactor transients

    International Nuclear Information System (INIS)

    Rossin, A.D.; Vine, G.L.

    1984-01-01

    This chapter discusses the cost estimation of transients and the causal relationship between transients and accidents. It is suggested that the calculation of the actual cost of a transient that has occurred is impossible without computerized records. Six months of operating experience reports, based on a survey of pressurized water reactors (PWRs) and boiling water reactors (BWRs) conducted by the Nuclear Safety Analysis Center (NSAC), are analyzed. The significant costs of a reactor transient are the repair costs resulting from severe damage to plant equipment, the cost of scrams (the actions the system is designed to take to avoid safety risks), US NRC fines, negative publicity, utility rates and revenues. It is estimated that the Three Mile Island-2 accident cost the US over $100 billion in nuclear plant delays and cancellations, more expensive fuel, oil imports, backfits, bureaucratic, administrative and legal costs, and lost productivity

  8. Advanced Instrumentation for Transient Reactor Testing

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

    2018-01-31

    Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

  9. PC-Reactor-core transient simulation code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

    PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt

  10. 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

  11. Incidental transients problems in reactor. Application examples

    International Nuclear Information System (INIS)

    Marbach, G.

    1988-03-01

    The fast neutron reactor fuel element qualification should be made not only for nominal operation but also for incidental and accidental transients. Different studies and tests permit to bring this justification such as simulation in hot laboratory after irradiation of irradiated pins or specific tests interpretation [fr

  12. Reactivity transient calculatios in research reactor

    International Nuclear Information System (INIS)

    Santos, R.S. dos

    1986-01-01

    A digital program for reactivity transient analysis in research reactor and cylindrical geometry was showed quite efficient when compared with methods and programs of the literature, as much in the solution of the neutron kinetics equation as in the thermohydraulic. An improvement in the representation of the feedback reactivity adopted on the program reduced markedly the computation time, with some accuracy. (Author) [pt

  13. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

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

  14. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  15. Genetic and neural approaches to nuclear transient identification

    International Nuclear Information System (INIS)

    Almeida, Jose Carlos Soares de; Mol, Antonio Carlos de Abreu; Pereira, Claudio Marcio Nascimento Abreu; Lapa, Celso Marcelo Franklin

    2005-01-01

    This work presents two approaches for pattern recognition to the same set of reactor signals. The first one describes a possibilistic approach optimized by genetic algorithm. The use of a possibilistic classification provides a natural and consistent classification rules, leading naturally to a good heuristic to handle the 'don't know' response, in case of unrecognized transient, which is fairly desirable in transient classification systems where safety is critical, since wrong or not reliable classifications can be catastrophic. Application of the proposed approach to a nuclear transient identification problem reveals good capability of the genetic algorithm in learning optimized possibilistic classification rules for efficient diagnosis including 'don't know' response. The second one uses two multilayer neural networks (NN). The first NN is responsible for the dynamic identification. This NN uses, as input, a short set (in a moving time window) of recent measurements of each variable avoiding the necessity of using starting events. The second NN is used to validate the instantaneous identification (from the first net) through the validation of each variable. This net is responsible for allowing the system to provide a 'don't know' response. In order to validate both methods, a Nuclear Power Plant (NPP) transient identification problem comprising postulated accidents, simulated for a pressurized water reactor, was proposed in the validation process it has been considered noisy data in order to evaluate the method robustness. Obtained results reveal the ability of the methods in dealing with both dynamic identification of transients and correct 'don't know' response. (author)

  16. RETRAN sensitivity studies of light water reactor transients. Final report

    International Nuclear Information System (INIS)

    Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

    1977-06-01

    This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

  17. Transient analysis of the IRIS reactor

    International Nuclear Information System (INIS)

    Bajs, T.; Oriani, L.; Ricotti, M.E.; Barroso, A.C.

    2002-01-01

    An international consortium of industry, laboratory, university and utility establishments, led by Westinghouse, is developing a modular, integral, light water cooled, small to medium power reactor, the International Reactor Innovative and Secure (IRIS). IRIS features innovative, advanced engineering, but it is firmly based on the proven technology of pressurized water reactors (PWR). Given the large number of organizations involved in the IRIS design, the RELAP5/MOD 3.3 code has been selected as the main system code. A nodalization of the reference IRIS design has been developed with a basic set of protective functions and controls. Engineered Safety Features of the concept are being also implemented, and in particular the Emergency Heat Removal System that is used for safety grade decay heat removal and in the small break LOCA response of IRIS (Large break LOCAs are eliminated in IRIS by the adoption of the Integral layout) This paper discusses developed model and transient behavior of the system for representative transient sequences.(author)

  18. Transient identification system with noising data and 'don't know' response

    International Nuclear Information System (INIS)

    Mol, Antonio C. de A.; Martinez, Aquilino S.; Schirru, Roberto

    2002-01-01

    In the last years, many different approaches based on neural network (NN) has been proposed for transient identification in nuclear power plants (NPP). Some of them focus the dynamic identification using recurrent neural networks however, they are not able to deal with unrecognized transients. Other kind of solution uses competitive learning in order to allow the 'don't know' response. In this case dynamic, dynamic features are not well represented. This work presents a new approach for neural network based transient identification which allows either dynamic identification and 'don't know'response. Such approach uses two multilayer neural networks trained with backpropagation algorithm. The first one is responsible for the dynamic identification. This NN uses, a short set (in a movable time window) of recent measurements of each variable avoiding the necessity of using starting events. The other one is used to validate the instantaneous identification (from the first net) through the validation of each variable. This net is responsible for allowing the system to provide 'don't know' response. In order to validate the method a NPP transient identification problem comprising 15 postulated accidents, simulated for a pressurized water reactor, was proposed in the validation process it has been considered noising data in other to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know' response. In order to validate the method, a NPP transient identification problem comprising 15 postulated accidents simulated for a pressurized water reactor, was proposed in the validation process it has been considered noising data in order to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know' response. (author)

  19. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    Energy Technology Data Exchange (ETDEWEB)

    Laureau, A., E-mail: laureau.axel@gmail.com; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-05-15

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  20. Transient coupled calculations of the Molten Salt Fast Reactor using the Transient Fission Matrix approach

    International Nuclear Information System (INIS)

    Laureau, A.; Heuer, D.; Merle-Lucotte, E.; Rubiolo, P.R.; Allibert, M.; Aufiero, M.

    2017-01-01

    Highlights: • Neutronic ‘Transient Fission Matrix’ approach coupled to the CFD OpenFOAM code. • Fission Matrix interpolation model for fast spectrum homogeneous reactors. • Application for coupled calculations of the Molten Salt Fast Reactor. • Load following, over-cooling and reactivity insertion transient studies. • Validation of the reactor intrinsic stability for normal and accidental transients. - Abstract: In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.

  1. Transient analysis for PWR reactor core using neural networks predictors

    International Nuclear Information System (INIS)

    Gueray, B.S.

    2001-01-01

    In this study, transient analysis for a Pressurized Water Reactor core has been performed. A lumped parameter approximation is preferred for that purpose, to describe the reactor core together with mechanism which play an important role in dynamic analysis. The dynamic behavior of the reactor core during transients is analyzed considering the transient initiating events, wich are an essential part of Safety Analysis Reports. several transients are simulated based on the employed core model. Simulation results are in accord the physical expectations. A neural network is developed to predict the future response of the reactor core, in advance. The neural network is trained using the simulation results of a number of representative transients. Structure of the neural network is optimized by proper selection of transfer functions for the neurons. Trained neural network is used to predict the future responses following an early observation of the changes in system variables. Estimated behaviour using the neural network is in good agreement with the simulation results for various for types of transients. Results of this study indicate that the designed neural network can be used as an estimator of the time dependent behavior of the reactor core under transient conditions

  2. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  3. Control of operational transients in power reactors - Methodology

    International Nuclear Information System (INIS)

    Vukovic, D.

    1983-01-01

    By introducing the nuclear power stations in the electric power system, questions of their possibilities to satisfy system's demand arise. Control of operational transients (temperature and Xe 135 ) in power reactors by determining the optimal control rod strategy is given. Ti optimize the Xe 135 transients, the Pantryagin theorem of optimal processes is applied. For solving three dimensional, two-group diffusion equations the heterogeneous Feinberg-Galanin method with axial flux harmonics is adopted. An application of this formalism to three-dimensional, finite cylindrical pressurised water reactor radially reflected is presented. (author)

  4. Classifier-ensemble incremental-learning procedure for nuclear transient identification at different operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Baraldi, Piero, E-mail: piero.baraldi@polimi.i [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Razavi-Far, Roozbeh [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Zio, Enrico [Dipartimento di Energia - Sezione Ingegneria Nucleare, Politecnico di Milano, via Ponzio 34/3, 20133 Milano (Italy); Ecole Centrale Paris-Supelec, Paris (France)

    2011-04-15

    An important requirement for the practical implementation of empirical diagnostic systems is the capability of classifying transients in all plant operational conditions. The present paper proposes an approach based on an ensemble of classifiers for incrementally learning transients under different operational conditions. New classifiers are added to the ensemble where transients occurring in new operational conditions are not satisfactorily classified. The construction of the ensemble is made by bagging; the base classifier is a supervised Fuzzy C Means (FCM) classifier whose outcomes are combined by majority voting. The incremental learning procedure is applied to the identification of simulated transients in the feedwater system of a Boiling Water Reactor (BWR) under different reactor power levels.

  5. Classifier-ensemble incremental-learning procedure for nuclear transient identification at different operational conditions

    International Nuclear Information System (INIS)

    Baraldi, Piero; Razavi-Far, Roozbeh; Zio, Enrico

    2011-01-01

    An important requirement for the practical implementation of empirical diagnostic systems is the capability of classifying transients in all plant operational conditions. The present paper proposes an approach based on an ensemble of classifiers for incrementally learning transients under different operational conditions. New classifiers are added to the ensemble where transients occurring in new operational conditions are not satisfactorily classified. The construction of the ensemble is made by bagging; the base classifier is a supervised Fuzzy C Means (FCM) classifier whose outcomes are combined by majority voting. The incremental learning procedure is applied to the identification of simulated transients in the feedwater system of a Boiling Water Reactor (BWR) under different reactor power levels.

  6. Analysis of short-term reactor cavity transient

    International Nuclear Information System (INIS)

    Cheng, T.C.; Fischer, S.R.

    1981-01-01

    Following the transient of a hypothetical loss-of-coolant accident (LOCA) in a nuclear reactor, peak pressures are reached within the first 0.03 s at different locations inside the reactor cavity. Due to the complicated multidimensional nature of the reactor cavity, the short-term analysis of the LOCA transient cannot be performed by using traditional containment codes, such as CONTEMPT. The advanced containment code, BEACON/MOD3, developed at the Idaho National Engineering Laboratory (INEL), can be adapted for such analysis. This code provides Eulerian, one and two-dimensional, nonhomogeneous, nonequilibrium flow modeling as well as lumped parameter, homogeneous, equilibrium flow modeling for the solution of two-component, two-phase flow problems. The purpose of this paper is to demonstrate the capability of the BEACON code to analyze complex containment geometry such as a reactor cavity

  7. Design criteria of integrated reactors based on transients

    International Nuclear Information System (INIS)

    Zanocco, P.; Gimenez, M.; Delmastro, D.

    1999-01-01

    A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author)

  8. Transients and burn dynamics in advanced tokamak fusion reactors

    International Nuclear Information System (INIS)

    Mantsinen, M.J.; Salomaa, R.R.E.

    1994-01-01

    Transient behavior of D 3 He-tokamak reactors is investigated numerically using a zero-dimensional code with prescribed profiles. Pure D 3 He start-up is compared to DT-assisted and DT-ignited start-ups. We have considered two categories of transients which could extinguish steady fusion burn: fuelling interruptions and sudden confinement changes similar to the L → H transients occurring in present-day tokamaks. Shutdown with various current and density ramp-down scenarios are studied, too. (author)

  9. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

    An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

  10. An efficient Neuro-Fuzzy approach to nuclear power plant transient identification

    Energy Technology Data Exchange (ETDEWEB)

    Gomes da Costa, Rafael [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Abreu Mol, Antonio Carlos de, E-mail: mol@ien.gov.br [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Instituto Nacional de C and T de Reatores Nucleares Inovadores (Brazil); Carvalho, Paulo Victor R. de, E-mail: paulov@ien.gov.br [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Lapa, Celso Marcelo Franklin, E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear - CNEN, Programa de Pos-Graduacao em Ciencia e Tecnologia Nucleares, Via Cinco, s/no, Cidade Universitaria, Rua Helio de Almeida, 75, Postal Box 68550, Zip Code 21941-906 Rio de Janeiro (Brazil); Instituto Nacional de C and T de Reatores Nucleares Inovadores (Brazil)

    2011-06-15

    Highlights: > We investigate a Neuro-Fuzzy modeling tool use for able transient identification. > The prelusive transient type identification is done by an artificial neural network. > After, the fuzzy-logic system analyzes the results emitting reliability degree of it. > The research support was made in a PWR simulator at the Brazilian Nuclear Engineering Institute. > The results show the potential to help operators' decisions in a nuclear power plant. - Abstract: Transient identification in nuclear power plants (NPP) is often a computational very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Recently, several works have been developed for transient identification. These works frequently present a non reliable response, using the 'don't know' as the system output. In this work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A validation of this identification system was made at the three loops Pressurized Water Reactor (PWR) simulator of the Human-System Interface Laboratory (LABIHS) of the Nuclear Engineering Institute

  11. An efficient Neuro-Fuzzy approach to nuclear power plant transient identification

    International Nuclear Information System (INIS)

    Gomes da Costa, Rafael; Abreu Mol, Antonio Carlos de; Carvalho, Paulo Victor R. de; Lapa, Celso Marcelo Franklin

    2011-01-01

    Highlights: → We investigate a Neuro-Fuzzy modeling tool use for able transient identification. → The prelusive transient type identification is done by an artificial neural network. → After, the fuzzy-logic system analyzes the results emitting reliability degree of it. → The research support was made in a PWR simulator at the Brazilian Nuclear Engineering Institute. → The results show the potential to help operators' decisions in a nuclear power plant. - Abstract: Transient identification in nuclear power plants (NPP) is often a computational very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Recently, several works have been developed for transient identification. These works frequently present a non reliable response, using the 'don't know' as the system output. In this work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A validation of this identification system was made at the three loops Pressurized Water Reactor (PWR) simulator of the Human-System Interface Laboratory (LABIHS) of the Nuclear Engineering Institute (IEN

  12. Transient neutrons flux behaviour in a spherical reactor core

    International Nuclear Information System (INIS)

    Souza, A.W.A. de.

    1978-11-01

    This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

  13. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  14. A neural model for transient identification in dynamic processes with 'don't know' response

    International Nuclear Information System (INIS)

    Mol, Antonio C. de A.; Martinez, Aquilino S.; Schirru, Roberto

    2003-01-01

    This work presents an approach for neural network based transient identification which allows either dynamic identification or a 'don't know' response. The approach uses two 'jump' multilayer neural networks (NN) trained with the backpropagation algorithm. The 'jump' network is used because it is useful to dealing with very complex patterns, which is the case of the space of the state variables during some abnormal events. The first one is responsible for the dynamic identification. This NN uses, as input, a short set (in a moving time window) of recent measurements of each variable avoiding the necessity of using starting events. The other one is used to validate the instantaneous identification (from the first net) through the validation of each variable. This net is responsible for allowing the system to provide a 'don't know' response. In order to validate the method, a Nuclear Power Plant (NPP) transient identification problem comprising 15 postulated accidents, simulated for a pressurized water reactor (PWR), was proposed in the validation process it has been considered noisy data in order to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know' response. Another important point studied in this work is that the system has shown to be independent of a trigger signal which indicates the beginning of the transient, thus making it robust in relation to this limitation

  15. Review of advanced reactor transient analysis capabilities and applications for Savannah River Plant reactors

    International Nuclear Information System (INIS)

    Buckner, M.R.; Hostetler, D.E.; Anderson, M.M.; Dodds, H.L.

    1977-01-01

    GRASS is a three-dimensional, coupled neutronic and engineering code for analysis of the radioisotope production reactors at the Savannah River Plant. The capabilities of GRASS are reviewed with emphasis on recent additions to model accident conditions involving the transport of molten fuel material and to accurately characterize neutronic and engineering feedback. The general application of GRASS to the Savannah River reactors is discussed, and results are presented for the analyses of severla reactor transient calculations

  16. A fast reactor transient analysis methodology for personal computers

    International Nuclear Information System (INIS)

    Ott, K.O.

    1993-01-01

    A simplified model for a liquid-metal-cooled reactor (LMR) transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All 30 differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes a new form, i.e., the quadratic dynamics equation. In this integral formulation, the initial value problem of typical LMR transients can be solved with large item steps (initially 1 s, later up to 256 s). This then makes transient problems amenable to a treatment on personal computer. The resulting mathematical model forms the basis for the GW-BASIC program LMR transient calculation (LTC) program. The LTC program has also been converted to QuickBASIC. The running time for a 10-h transient overpower transient is then ∼40 to 10 s, depending on the hardware version (286, 386, or 486 with math coprocessors)

  17. Accident transient processes at NPPs with the WWER type reactors

    International Nuclear Information System (INIS)

    Bukrinskij, A.M.

    1982-01-01

    Thermal-physical and nuclear-physical transient processes at NPPs with the WWER type reactors during accidents with the main technological equipment failures and the accidents with loss of coolant in the primary and secondary coolant circuits are considered. Mathematical methods used for these processes modelling is described. Examples of concrete calculations for accidents with different failures are given. Comparative analysis of the results of dynamic tests at the Novo-Voronezh-3 reactor is presented. It is concluded that the modern NPP design is impossible without application of mathematical modelling methods. The mathematical modelling of transients is also necessary for proper and safe NPP operation. Mathematical modelling of accidents at NPPs is a comparatively new method of investigation. Its success and development are completely based on the progress in modern computer development. With their improvement the mathematical models will become more complicate and adequacy of real physical process representation by their means will increase

  18. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    Carew, J.F.; Neogy, P.

    1983-01-01

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  19. Taipower's transient analysis methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Pinghue

    1998-01-01

    The methodology presented in this paper is a part of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors' developed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research. This methodology utilizes four computer codes developed or sponsored by Electric Power Research institute: system transient analysis code RETRAN-02, core thermal-hydraulic analysis code COBRAIIIC, three-dimensional spatial kinetics code ARROTTA, and fuel rod evaluation code FREY. Each of the computer codes was extensively validated. Analysis methods and modeling techniques were conservatively established for each application using a systematic evaluation with the assistance of sensitivity studies. The qualification results and analysis methods were documented in detail in TPC topical reports. The topical reports for COBRAIIIC, ARROTTA. and FREY have been reviewed and approved by the Atomic Energy Council (ABC). TPC 's in-house transient methodology have been successfully applied to provide valuable support for many operational issues and plant improvements for TPC's Maanshan Units I and 2. Major applications include the removal of the resistance temperature detector bypass system, the relaxation of the hot-full-power moderator temperature coefficient design criteria imposed by the ROCAEC due to a concern on Anticipated Transient Without Scram, the reduction of boron injection tank concentration and the elimination of the heat tracing, and the reduction of' reactor coolant system flow. (author)

  20. Parameter identification in a nonlinear nuclear reactor model using quasilinearization

    International Nuclear Information System (INIS)

    Barreto, J.M.; Martins Neto, A.F.; Tanomaru, N.

    1980-09-01

    Parameter identification in a nonlinear, lumped parameter, nuclear reactor model is carried out using discrete output power measurements during the transient caused by an external reactivity change. In order to minimize the difference between the model and the reactor power responses, the parameter promt neutron generation time and a parameter in fuel temperature reactivity coefficient equation are adjusted using quasilinearization. The influences of the external reactivity disturbance, the number and frequency of measurements and the measurement noise level on the method accuracy and rate of convergence are analysed through simulation. Procedures for the design of the identification experiments are suggested. The method proved to be very effective for low level noise measurements. (Author) [pt

  1. AIREKMOD-RR, Reactivity Transients in Nuclear Research Reactors

    International Nuclear Information System (INIS)

    Baggoura, B.; Mazrou, H.

    2001-01-01

    1 - Description of program or function: AIREMOD-RR is a point kinetics code which can simulate fast transients in nuclear research reactor cores. It can also be used for theoretical reactor dynamics studies. It is used for research reactor kinetic analysis and provides a point neutron kinetic capability. The thermal hydraulic behavior is governed by a one-dimensional heat balance equation. The calculations are restricted to a single equivalent unit cell which consists of fuel, clad and coolant. 2 - Method of solution: For transient reactor kinetic calculations a modified Runge Kutta numerical method is used. The external reactivity insertion, specified as a function of time, is converted in dollar ($) unit. The neutron density, energy release and feedback variables are given at each time step. The two types of reactivity feedback considered are: Doppler effect and moderator effect. A new expression for the reactivity dependence on the feedback variables has been introduced in the present version of the code. The feedback reactivities are fitted in power series expression. 3 - Restrictions on the complexity of the problem: The number of delayed neutron groups and the total number of equations are limited only by computer storage capabilities. - Coolant is always in liquid phase. - Void reactivity feedback is not considered

  2. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Solomon, K.A.

    1979-07-01

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  3. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  4. Transient Analysis Needs for Generation IV Reactor Concepts

    International Nuclear Information System (INIS)

    Siefken, L.J.; Harvego, E.A.; Coryell, E.W.; Davis, C.B.

    2002-01-01

    The importance of nuclear energy as a vital and strategic resource in the U. S. and world's energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE's Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors. (authors)

  5. Transient Analysis of Generation IV quick reactors; Analisis de Transitorios en Reactores Rapidos de Generacion IV

    Energy Technology Data Exchange (ETDEWEB)

    Vazquez, M.; Martin-Fuertes, F.

    2013-07-01

    As a complement to the attached code 3D neutron-CIEMAT thermohydraulic added a module to simulate transient. Temporary kinetics is resolved by factoring flow in a spatial part and another storm. MCNP provides the reactivity and updated spatial function and COBRA-IV calculates the temperature distribution. Temporary dependence of amplitude is calculated using time delayed neutron Kinetic equations. As an example of application, examines a transient loss of flow in MYRRHA, a lead-cooled experimental reactor.

  6. Transient classification for the IRIS reactor using self-organized maps built in free platform

    International Nuclear Information System (INIS)

    Doraskevicius Junior, Waldemar

    2005-01-01

    Kohonen's Self Organized Maps (SOM) were tested with data from several operational conditions of the nuclear reactor IRIS (International Reactor Innovative and Secure) to develop an effective tool in the classification and transient identification in nuclear reactors. The data were derived from 56 simulations of the operation of IRIS, from steady-state conditions to accidents. The digital system built for the tests was based on the JAVA platform for the portability and scalability, and for being one of the free development platforms. Satisfactory results of operation classification were obtained with reasonable processing time in personal computers; about two to five minutes were spent for ordination and convergence of the learning on the data base. The methodology of this work was extended to the supervision of logistics of natural gas for Brazilian pipelines, showing satisfactory results for the classification of deliveries for simultaneous measurement in several points. (author)

  7. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  8. Transient response of Babcock and Wilcox-designed reactors

    International Nuclear Information System (INIS)

    1980-05-01

    On February 26, 1980, the Crystal River Unit No. 3 Nuclear Generating Plant, designed by the Babcock and Wilcox Company (B and W), experienced an incident involving a malfunction in an instrumentation and control system power supply. Faced with the Crystal River Unit 3 incident and the apparently high frequency of such near similar types of transients in other B and W designed plants, a special Task Force was established within the Office of Nuclear Reactor Regulation to provide an assessment of the apparent sensitivity of the B and W designed plants to such transients and the consequences of malfunctions and failures of the integrated control system and non-nuclear instrumentation. This report provides an assessment of these issues

  9. The Dynamic Monte Carlo Method for Transient Analysis of Nuclear Reactors

    NARCIS (Netherlands)

    Sjenitzer, B.L.

    2013-01-01

    In this thesis a new method for the analysis of power transients in a nuclear reactor is developed, which is more accurate than the present state-of-the-art methods. Transient analysis is important tool when designing nuclear reactors, since they predict the behaviour of a reactor during changing

  10. PUMP: analog-hybrid reactor coolant hydraulic transient model

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

    The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

  11. Transient performance of flow in circuits of PWR type reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.; Carajilescov, P.

    1988-09-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  12. Transient performance of flow in PWR reactor circuits

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.; Carajilescov, P.

    1988-12-01

    Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

  13. Qualitative diagnosis for transients analysis on nuclear reactors

    International Nuclear Information System (INIS)

    Lorre, J.P.; Dorlet, E.; Evrard, J.M.

    1995-01-01

    One of the major aims of an intelligent monitoring system, is the supervision task which assist the operator in understanding what occurs on a process. Failures hypotheses must be located and the inferring process must be explained. This paper demonstrate a second generation expert system (SEXTANT) decided to the transients analysis on PWR nuclear reactors. This system detects failures by simulating the process with a numerical model. A diagnosis module uses an even graph built from a causal graph model of the plant to generate hypotheses, and a numerical model to validate these hypotheses. Hypotheses are stored into scenarios which are concurrent possible interpretations of the process evolution. The approach is illustrated by an application for the analysis of the house load operation on a pressurized water reactor. (authors). 9 refs., 10 figs

  14. Modelling of an ULOF transient in a sodium fast reactor

    International Nuclear Information System (INIS)

    Droin, Jean-Baptiste

    2016-01-01

    Within the framework of the Generation IV Sodium-cooled Fast Reactor (SFR) R and D program of CEA (French Commissariat a l'Energie Atomique et aux Energies Alternatives), safety in case of severe accidents is assessed.Such transients are usually simulated with mechanistic codes (such as SAS-SFR and SIMMER III). as a complement to these codes, which give reference accidental transient calculations, a new physico-statistical approach is currently followed by the CEA; its final objective being to derive the variability of the main results of interest for safety. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena to advanced statistical analysis techniques. It enables to perform a large number of simulations in a reasonable computational time and to describe all the possible bifurcations of the accident transient.In this context, this PhD work presents the physical tool (models and results assessment) dedicated to the initiation and primary phases of an Unprotected Loss Of Flow accident (i.e. until the end of sub-assemblies degradation and before large molten pools formation). The accident phenomenology during these phases is described and illustrated by numerous experimental evidences.It is underlined that the features of the new heterogeneous core concept (called CFV of the French ASTRID prototype) leads to different kinds of ULOF transients than those occurring in the previous past homogeneous cores (SuperPhenix, Phenix...). Indeed, its negative void effect drops the nuclear power when sodium heats-up and possibly boils. This enables three types of ULOF transients characterized by various core final states; the first two types leading to final coolable core states in natural circulation flow (the first one in single phase, the second one in stabilized two-phase flow) whereas the core undergoes a flow excursion followed by sub-assemblies degradation in the last type. In this study, a

  15. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  16. Temporal neural network for the identification of nuclear power plant transients

    International Nuclear Information System (INIS)

    Uluyol, O.; Ragheb, M.

    1993-01-01

    In this paper a layered spatiotemporal neural network is proposed for the identification of nuclear power plant transients. The developed layered spatiotemporal network is inspired by the formal avalanche structure developed by S. Grossberg and offers advantages compared with the stationary pattern approach using the perceptron paradigm. Each layer in the network is trained to recognize a separate time-dependent accident scenario. Within each scenario, the temporal behavior of the relevant parameters such as pressurizer pressure, pressurizer water volume, cold and hot legs temperatures, vessel flow, and power, are considered. Numerical cases are considered where the proposed methodology is applied to two nuclear power plant anticipated transient scenarios: the Station Blackout and the Anticipated Transient without Scram transients in a pressurized water reactor . The transient signatures used were generated by modeling the accidents using RELAP5/MOD2, a best-estimate thermal-hydraulics numerical code. The ability of the proposed layered spatiotemporal network to operate at different noise levels is investigated. Its incorporation within an Insightful Algorithm and Anticipatory Systems context for identifying and in predicting the course of nuclear transients is discussed

  17. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    International Nuclear Information System (INIS)

    Wright, A.E.; Dutt, D.S.; Harrison, L.J.

    1990-01-01

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

  18. Gas-core reactor power transient analysis. Final report

    International Nuclear Information System (INIS)

    Kascak, A.F.

    1972-01-01

    The gas core reactor is a proposed device which features high temperatures. It has applications in high specific impulse space missions, and possibly in low thermal pollution MHD power plants. The nuclear fuel is a ball of uranium plasma radiating thermal photons as opposed to gamma rays. This thermal energy is picked up before it reaches the solid cavity liner by an inflowing seeded propellant stream and convected out through a rocket nozzle. A wall-burnout condition will exist if there is not enough flow of propellant to convect the energy back into the cavity. A reactor must therefore operate with a certain amount of excess propellant flow. Due to the thermal inertia of the flowing propellant, the reactor can undergo power transients in excess of the steady-state wall burnout power for short periods of time. The objective of the study was to determine how long the wall burnout power could be exceeded without burning out the cavity liner. The model used in the heat-transfer calculation was one-dimensional, and thermal radiation was assumed to be a diffusion process. (auth)

  19. Rapid thermal transient in a reactor coolant channel

    International Nuclear Information System (INIS)

    Cherubini, A.

    1986-01-01

    This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow

  20. Anticipated transients without scram for light water reactors

    International Nuclear Information System (INIS)

    1978-12-01

    In the first two volumes of this report, Anticipated Transients without Scram for Light Water Reactors NUREG-0460, dated April 1978, the NRC staff reviewed the information on this subject that had been developed in the past and evaluated the susceptibility of current nuclear plants to ATWS events using fault tree/event tree analysis techniques. Based on that evaluation, the staff concluded that some corrective measures were required to reduce the risk of severe consequences arising from possible ATWS events. Since the issuance of NUREG-0460, new safety and cost information has become available on ATWS. Also, new insights have been developed on the general subject of quantitative risk assessment. The purpose of this supplement to NUREG-0460 is to summarize the important additions to the information base and to propose a course of action from among a variety of alternatives for resolving the ATWS concern

  1. Tightly coupled simulation of nuclear reactor transients with artificial intelligence

    International Nuclear Information System (INIS)

    Makowitz, H.; Ragheb, M.; Laats, E.T.; Bray, M.A.

    1985-01-01

    The authors' current efforts are directed toward exploring new avenues of research in simulation of nuclear reactor kinetics transients with artificial intelligence (AI). Being examined are advanced graphics systems such as the Nuclear Plant Analyzer designed to run in parallel with the RELAP5 code, faster than real-time best-estimate simulations, the utilization of the multi-CPU super computers, and simulation as knowledge by attempting to develop new assessment methodologies for artificial intelligence systems and their associated interfaces. This new and fertile area of research should be viewed by the educational and university community as an indication of the future possibilities for AI developments in a number of academic and engineering disciplines

  2. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  3. Transient response of level instruments in a research reactor

    International Nuclear Information System (INIS)

    Cheng, Lap Y.

    1989-01-01

    A numerical model has been developed to simulate the dynamics of water level instruments in a research nuclear reactor. A bubble device, with helium gas as the working fluid, is used to monitor liquid level by sensing the static head pressure due to the height of liquid in the reactor vessel. A finite-difference model is constructed to study the transient response of the water level instruments to pressure perturbations. The field equations which describe the hydraulics of the helium gas in the bubbler device are arranged in the form of a tridiagonal matrix and the field variables are solved at each time step by the Thomas algorithm. Simulation results indicate that the dynamic response of the helium gas depends mainly on the volume and the inertia of the gas in the level instrument tubings. The anomalies in the simulated level indication are attributed to the inherent lag in the level instrument due to the hydraulics of the system. 1 ref., 5 figs

  4. ATWS analyses. Analysis of anticipated transients without reactor scram in Combustion Engineering NSSS's

    International Nuclear Information System (INIS)

    1976-05-01

    Results are presented of analyses of the transient thermal-hydraulic conditions and radiological release consequences which would occur in power plants which employ a Combustion Engineering Nuclear Steam Supply System during Anticipated Transients Without Scram due to a lack of insertion of the Control Element Assemblies upon signals for automatic or manual reactor shutdown. The transients analyzed include all events which meet the criterion to be considered as anticipated at least once in the plant lifetime with automatic reactor shutdown

  5. Models for transient analyses in advanced test reactors

    International Nuclear Information System (INIS)

    Gabrielli, Fabrizio

    2011-01-01

    Several strategies are developed worldwide to respond to the world's increasing demand for electricity. Modern nuclear facilities are under construction or in the planning phase. In parallel, advanced nuclear reactor concepts are being developed to achieve sustainability, minimize waste, and ensure uranium resources. To optimize the performance of components (fuels and structures) of these systems, significant efforts are under way to design new Material Test Reactors facilities in Europe which employ water as a coolant. Safety provisions and the analyses of severe accidents are key points in the determination of sound designs. In this frame, the SIMMER multiphysics code systems is a very attractive tool as it can simulate transients and phenomena within and beyond the design basis in a tightly coupled way. This thesis is primarily focused upon the extension of the SIMMER multigroup cross-sections processing scheme (based on the Bondarenko method) for a proper heterogeneity treatment in the analyses of water-cooled thermal neutron systems. Since the SIMMER code was originally developed for liquid metal-cooled fast reactors analyses, the effect of heterogeneity had been neglected. As a result, the application of the code to water-cooled systems leads to a significant overestimation of the reactivity feedbacks and in turn to non-conservative results. To treat the heterogeneity, the multigroup cross-sections should be computed by properly taking account of the resonance self-shielding effects and the fine intra-cell flux distribution in space group-wise. In this thesis, significant improvements of the SIMMER cross-section processing scheme are described. A new formulation of the background cross-section, based on the Bell and Wigner correlations, is introduced and pre-calculated reduction factors (Effective Mean Chord Lengths) are used to take proper account of the resonance self-shielding effects of non-fuel isotopes. Moreover, pre-calculated parameters are applied

  6. Transient identification system with noising data and 'don't know' response; Sistema de identificacao de transientes com inclusao de ruidos e indicacao de eventos desconhecidos

    Energy Technology Data Exchange (ETDEWEB)

    Mol, Antonio C. de A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil). Div. de Confiabilidade Humana; Martinez, Aquilino S.; Schirru, Roberto [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    In the last years, many different approaches based on neural network (NN) has been proposed for transient identification in nuclear power plants (NPP). Some of them focus the dynamic identification using recurrent neural networks however, they are not able to deal with unrecognized transients. Other kind of solution uses competitive learning in order to allow the 'don't know' response. In this case dynamic, dynamic features are not well represented. This work presents a new approach for neural network based transient identification which allows either dynamic identification and 'don't know'response. Such approach uses two multilayer neural networks trained with backpropagation algorithm. The first one is responsible for the dynamic identification. This NN uses, a short set (in a movable time window) of recent measurements of each variable avoiding the necessity of using starting events. The other one is used to validate the instantaneous identification (from the first net) through the validation of each variable. This net is responsible for allowing the system to provide 'don't know' response. In order to validate the method a NPP transient identification problem comprising 15 postulated accidents, simulated for a pressurized water reactor, was proposed in the validation process it has been considered noising data in other to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know' response. In order to validate the method, a NPP transient identification problem comprising 15 postulated accidents simulated for a pressurized water reactor, was proposed in the validation process it has been considered noising data in order to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know

  7. Transient thermal creep of nuclear reactor pressure vessel type concretes

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1983-01-01

    The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

  8. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    International Nuclear Information System (INIS)

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250 0 C)

  9. Thermal-hydraulic transient characteristics of ship-propulsion reactor investigated through safety analysis

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

    Thermal-hydraulic behaviors in the reactor of Nuclear Ship ''Mutsu'' were investigated through safety evaluation of operational transients by using RETRAN and COBRA-IV codes. The results were compared to the transient behaviors of typical commercial PWR and the characteristics of transient thermal-hydraulic behaviors in ship-loaded reactor were figured out. ''Mutsu'' reactor has larger thermal margin than commercial PWR because it is designed to be used as ship-propulsion power source in the load-following operation mode. This margin makes transient behavior in general milder than in commercial PWR but high opening pressure set point of main-steam safety valves leads poor heat-sink condition after reactor trip. The effects of other small-sized components are also investigated. The findings in the paper will be helpful in the design of future advanced reactor for nuclear ship. (author)

  10. Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes

    International Nuclear Information System (INIS)

    Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.

    2002-01-01

    A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)

  11. Study on the transient behaviours of MNSR reactor for control rod withdrawal

    International Nuclear Information System (INIS)

    Yang Shunhai

    1995-10-01

    The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  14. Thermohydraulic characteristics under some transient conditions of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang; Khang, Ngo Phu; An, Tran Khac; Nghiem, Huynh Ton [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Some experimental and theoretical thermal hydraulic characteristics of the Dalat Nuclear Research Reactor are presented, together with some general assessments, from a thermal hydraulic point of view, of its safety under transient conditions. (author). 3 refs., 9 figs., 7 tabs.

  15. Development of a computer code for Dalat research reactor transient analysis

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong

    2003-01-01

    DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)

  16. Development of refined MCNPX-PARET multi-channel model for transient analysis in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, S.; Koonen, E. [SCK-CEN, BR2 Reactor Dept., Boeretang 200, 2400 Mol (Belgium); Olson, A. P. [RERTR Program, Nuclear Engineering Div., Argonne National Laboratory, Cass Avenue, Argonne, IL 60439 (United States)

    2012-07-01

    Reactivity insertion transients are often analyzed (RELAP, PARET) using a two-channel model, representing the hot assembly with specified power distribution and an average assembly representing the remainder of the core. For the analysis of protected by the reactor safety system transients and zero reactivity feedback coefficients this approximation proves to give adequate results. However, a more refined multi-channel model representing the various assemblies, coupled through the reactivity feedback effects to the whole reactor core is needed for the analysis of unprotected transients with excluded over power and period trips. In the present paper a detailed multi-channel PARET model has been developed which describes the reactor core in different clusters representing typical BR2 fuel assemblies. The distribution of power and reactivity feedback in each cluster of the reactor core is obtained from a best-estimate MCNPX calculation using the whole core geometry model of the BR2 reactor. The sensitivity of the reactor response to power, temperature and energy distributions is studied for protected and unprotected reactivity insertion transients, with zero and non-zero reactivity feedback coefficients. The detailed multi-channel model is compared vs. simplified fewer-channel models. The sensitivities of transient characteristics derived from the different models are tested on a few reactivity insertion transients with reactivity feedback from coolant temperature and density change. (authors)

  17. Measurement of transient hydrodynamic characteristics of the reactor RA primary cooling system

    International Nuclear Information System (INIS)

    Jovic, L.; Majstorovic, D.; Zeljkovic, I.

    1987-01-01

    Experimental study of transient hydrodynamic characteristics of the research nuclear reactor RA by simultaneous measurements of fluid flow and pressure on several locations of the RA primary coolant system is done. Loss of electric power transient on the main circulation pumps is simulated. measurement methodology, data processing and results of measured data analysis are given. (author)

  18. Theoretical basis for a transient thermal elastic-plastic stress analysis of nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hsu, T.R.; Bertels, A.W.M.; Banerjee, S.; Harrison, W.C.

    1976-07-01

    This report presents the theoretical basis for a transient thermal elastic-plastic stress analysis of a nuclear reactor fuel element subject to severe transient thermo-mechanical loading. A finite element formulation is used for both the non-linear stress analysis and thermal analysis. These two major components are linked together to form an integrated program capable of predicting fuel element transient behaviour in two dimensions. Specific case studies are presented to illustrate capabilities of the analysis. (author)

  19. Transient simulation of an endothermic chemical process facility coupled to a high temperature reactor: Model development and validation

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Seker, Volkan; Revankar, Shripad T.; Downar, Thomas J.

    2012-01-01

    Highlights: ► Models for PBMR and thermochemical sulfur cycle based hydrogen plant are developed. ► Models are validated against available data in literature. ► Transient in coupled reactor and hydrogen plant system is studied. ► For loss-of-heat sink accident, temperature feedback within the reactor core enables shut down of the reactor. - Abstract: A high temperature reactor (HTR) is a candidate to drive high temperature water-splitting using process heat. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Three thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. The models and coupling scheme are presented here, as well as a transient test case initiated within the chemical plant. The 50% feed flow failure within the chemical plant results in a slow loss-of-heat sink (LOHS) accident in the nuclear reactor. Due to the temperature feedback within the reactor core the nuclear reactor partially shuts down over 1500 s. Two distinct regions are identified within the coupled plant response: (1) immediate LOHS due to the loss of the sulfuric

  20. Identification of nuclear power plant transients using the Particle Swarm Optimization algorithm

    International Nuclear Information System (INIS)

    Canedo Medeiros, Jose Antonio Carlos; Schirru, Roberto

    2008-01-01

    In order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition, transient identification systems have been devised to help operators identify possible plant transients and take fast and right corrective actions in due time. In the design of classification systems for identification of nuclear power plants transients, several artificial intelligence techniques, involving expert systems, neuro-fuzzy and genetic algorithms have been used. In this work we explore the ability of the Particle Swarm Optimization algorithm (PSO) as a tool for optimizing a distance-based discrimination transient classification method, giving also an innovative solution for searching the best set of prototypes for identification of transients. The Particle Swarm Optimization algorithm was successfully applied to the optimization of a nuclear power plant transient identification problem. Comparing the PSO to similar methods found in literature it has shown better results

  1. Identification of nuclear power plant transients using the Particle Swarm Optimization algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Canedo Medeiros, Jose Antonio Carlos [Universidade Federal do Rio de Janeiro, PEN/COPPE, UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE, UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br

    2008-04-15

    In order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition, transient identification systems have been devised to help operators identify possible plant transients and take fast and right corrective actions in due time. In the design of classification systems for identification of nuclear power plants transients, several artificial intelligence techniques, involving expert systems, neuro-fuzzy and genetic algorithms have been used. In this work we explore the ability of the Particle Swarm Optimization algorithm (PSO) as a tool for optimizing a distance-based discrimination transient classification method, giving also an innovative solution for searching the best set of prototypes for identification of transients. The Particle Swarm Optimization algorithm was successfully applied to the optimization of a nuclear power plant transient identification problem. Comparing the PSO to similar methods found in literature it has shown better results.

  2. Transient thermal characteristics of a core channel in a molten salt reactor

    International Nuclear Information System (INIS)

    Sakashita, H.; Ishiguro, R.; Sugiyama, K.

    1987-01-01

    The present paper deals with the thermal characteristics of Molten Salt Reactor (MSR). Analyses of the fundamental behavior of internal heat generating fluid and graphite contiguous to the fluid are performed. As a result, it is known that the transient thermal characteristics of MSR differ fundamentally from those of a solid-fuel reactor, and the simplified method of thermal analysis which is commonly used for solid-fuel reactors gives optimistic predictions than the actual phenomena. (author)

  3. Analytical prediction and experimental verification of reactor safety system injection transient

    International Nuclear Information System (INIS)

    Roy, B.N.; Nomm, E.

    1991-01-01

    This paper describes the computer code that was developed for thermal hydraulic transient analysis of mixed phase fluid system and the flow tests that were carried out to validate the Code. A full scale test facility was designed to duplicate the Supplementary Shutdown System (SSS) of Savannah River Production Reactors. Several steady state and dynamic flow tests were conducted simulating the actual reactor injection transients. A dynamic multiphase fluid flow code was developed and validated with experimental results and utilized for system performance predictions and development of technical specifications for reactors. 3 refs

  4. Estimation of power feedback parameters of pulse reactor IBR-2M on transients

    International Nuclear Information System (INIS)

    Pepyolyshev, Yu.N.; Popov, A.K.

    2013-01-01

    Parameters of the IBR-2M reactor power feedback (PFB) on a model of the reactor dynamics by mathematical treatment of two registered transients are estimated. Frequency characteristics and the pulse transient characteristics corresponding to these PFB parameters are calculated. PFB parameters received thus can be considered as their express tentative estimation as real measurements in this case occupy no more than 30 minutes. Total PFB is negative at 1 and 2 MW. At the received estimations of PFB parameters in a self-regulation mode it is possible to consider the stability margins of the IBR-2M reactor satisfactory

  5. Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

    Science.gov (United States)

    Hoogenboom, J. Eduard; Sjenitzer, Bart L.

    2014-06-01

    To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

  6. Simulation of hot-channel transients for PHWR reactors

    International Nuclear Information System (INIS)

    Masriera, N.A.

    1988-01-01

    For the simulation of transients a whole-plant code is needed. These codes model the core in a very simplified way. When local variables have to be calculated a different kind of code is needed: a subchannel-code. This report studies the use of the cobra code as a subchannel-code, for the simulation of a PHWR fuel channel, considering that this code was developed for PWR cores calculation. A special effort is made to obtain optimized models for different calculations: steady state, soft transients and severe transients. These models differ in number of subchannels, axial nodes, and the choice of the most important variables. (Author) [es

  7. Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

    International Nuclear Information System (INIS)

    Yulianti, Yanti; Su'ud, Zaki; Waris, Abdul; Khotimah, S. N.; Shafii, M. Ali

    2010-01-01

    The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

  8. RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

    This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

  9. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  11. Comparison of the transient behavior of lead-based advanced critical and sub-critical reactors

    International Nuclear Information System (INIS)

    Wang Gang; Gu Zhixing; Wang Zhen; Jin Ming; Bai Yunqing

    2014-01-01

    A lead-based reactor developed by FDS Team is proposed in 2011 and designed to be 10 MW. It is a pool type reactor and the primary coolant is driven by natural circulation. The reactor has two operation modes, which are a lead-based critical fast reactor mode and a lead-based sub-critical reactor mode. The conceptual designs of the two modes are both completed by 2013. In this paper, four transient accidents were simulated for both the critical and sub-critical reactors above by NTC-2D code, which is developed by FDS Team for advanced reactor safety analysis. The four accidents were protected and unprotected loss of heat sink accidents (PLOHS and ULOHS), protected and unprotected transient overpower accidents (PTOP and UTOP). The simulation results of the two reactors were compared and analyzed. The results showed that during PLOHS and PTOP accidents for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the sub-critical reactor increased bigger than those of the critical one. For UTOP, the parameters above of the critical fast reactor were much bigger than those of the sub-critical one. The analysis results showed different safety advantages of the lead-based critical fast and sub-critical reactors during different transient accidents. (author)

  12. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  13. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  14. Simulation of Safety and Transient Analysis of a Pressurized Water Reactor using the Personal Computer Transient Analyzer

    Directory of Open Access Journals (Sweden)

    Sunday J. IBRAHIM

    2013-06-01

    Full Text Available Safety and transient analyses of a pressurised water reactor (PWR using the Personal Computer Transient Analyzer (PCTRAN simulator was carried out. The analyses presented a synergistic integration of a numerical model; a full scope high fidelity simulation system which adopted point reactor neutron kinetics model and movable boundary two phase fluid models to simplify the calculation of the program, so it could achieve real-time simulation on a personal computer. Various scenarios of transients and accidents likely to occur at any nuclear power plant were simulated. The simulations investigated the change of signals and parameters vis a vis loss of coolant accident, scram, turbine trip, inadvertent control rod insertion and withdrawal, containment failure, fuel handling accident in auxiliary building and containment, moderator dilution as well as a combination of these parameters. Furthermore, statistical analyses of the PCTRAN results were carried out. PCTRAN results for the loss of coolant accident (LOCA caused a rapid drop in coolant pressure at the rate of 21.8KN/m2/sec triggering a shutdown of the reactor protection system (RPS, while the turbine trip accident showed a rapid drop in total plant power at the rate of 14.3 MWe/sec causing a downtime in the plant. Fuel handling accidents mimic results showed release of radioactive materials in unacceptable doses. This work shows the potential classes of nuclear accidents likely to occur during operation in proposed reactor sites. The simulations are very appropriate in the light of Nigeria’s plan to generate nuclear energy in the region of 1000 MWe from reactors by 2017.

  15. A neural model for transient identification in dynamic processes with 'don't know' response

    Energy Technology Data Exchange (ETDEWEB)

    Mol, Antonio C. de A. E-mail: mol@ien.gov.br; Martinez, Aquilino S. E-mail: aquilino@lmp.ufrj.br; Schirru, Roberto E-mail: schirru@lmp.ufrj.br

    2003-09-01

    This work presents an approach for neural network based transient identification which allows either dynamic identification or a 'don't know' response. The approach uses two 'jump' multilayer neural networks (NN) trained with the backpropagation algorithm. The 'jump' network is used because it is useful to dealing with very complex patterns, which is the case of the space of the state variables during some abnormal events. The first one is responsible for the dynamic identification. This NN uses, as input, a short set (in a moving time window) of recent measurements of each variable avoiding the necessity of using starting events. The other one is used to validate the instantaneous identification (from the first net) through the validation of each variable. This net is responsible for allowing the system to provide a 'don't know' response. In order to validate the method, a Nuclear Power Plant (NPP) transient identification problem comprising 15 postulated accidents, simulated for a pressurized water reactor (PWR), was proposed in the validation process it has been considered noisy data in order to evaluate the method robustness. Obtained results reveal the ability of the method in dealing with both dynamic identification of transients and correct 'don't know' response. Another important point studied in this work is that the system has shown to be independent of a trigger signal which indicates the beginning of the transient, thus making it robust in relation to this limitation.

  16. Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Wang Yan; Li Fu; Zheng Yanhua

    2014-01-01

    In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

  17. Applying a neuro-fuzzy approach for transient identification in a nuclear power plant

    International Nuclear Information System (INIS)

    Costa, Rafael G.; Mol, Antonio C.A.; Pereira, Claudio M.N.A.; Carvalho, Paulo V.R.

    2009-01-01

    Transient identification in Nuclear Power Plant (NPP) is often a very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Several systems based on specialist systems, neural networks, and fuzzy logic have been developed for transient identification. In the work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A preliminary evaluation of the developed system was made at the Human-System Interface Laboratory (LABIHS). The obtained results show that the system can help the operators to take decisions during transients/accidents in the plant. (author)

  18. A fast reactor transient analysis methodology for PCs

    International Nuclear Information System (INIS)

    Ott, K.O.

    1991-10-01

    This Manual describes a PC program for LMR Transient Calculations, LTC, written in GW-BASIC. It calculates the power and temperature trajectories for unscrammed TOP and LOHS transients. The LOF transient treatment is not operational in the GW-BASIC program because of storage limitations. The corresponding mathematical model, which allows a rapid treatment of the kinetics and the various feedback effects, is described in Ref. 1. It is briefly reviewed in Sec. 1. The program structure is outlined in Sec. 2, followed by a more detailed description in Sec. 3. Computational details are presented in Appendix A. A complete listing of the GW-BASIC program is given in Appendix B. Appendix C shows input-echo and output for a TOP sample problem, and Appendix D is a Glossary of all quantities used in the LTC program. The limitations of the GW-BASIC storage (to about 60K) are removed if it is run within Quick-BASIC. This then allows the extension of this program to treat LOF transients. Running LTC in Quick-BASIC permits also larger ''Dimensions'' for TOP and LOHS transients

  19. Study on transient of fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Streck, E.E.

    1988-01-01

    The point kinetic equations for a Fluidized-Bed Nuclear Reactor are solved by the method of Hansen. Due to the time varying nature of the reactor volume, the equations have a non-conventional formulation (moving boundary problem), but the method of solution preserves its asymptotic convergence and efficiency characteristics under this formulation. A one dimensional and linearized thermal hydraulics feedback model was coupled to the point kinetic equations in order to obtain a more realistic representation of the reactor power. The resulting equations are solved by the Euler explicit method. (author)

  20. Nuclear power plant transient identification using a neuro-fuzzy inference system

    International Nuclear Information System (INIS)

    Mol, Antonio Carlos de Abreu; Oliveira, Mauro Vitor de; Santos, Isaac Jose Antonio Luchetti dos; Carvalho, Paulo Victor Rodrigues de; Grecco, Claudio Henrique dos Santos; Auguto, Silas Cordeiro

    2005-01-01

    Transient identification in Nuclear Power Plant (NPP) is often a very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in nuclear power plants. The basis for the identification of a change in the system is that different system faults and anomalies lead to different patterns of evolution of the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments, that represents a specific type of event. In this work, an approach for the identification of transients is presented, aiming at helping the operator to make a decision relative to the procedure to be followed in situations of accidents/transients at nuclear power plants. In this way, a diagnostic strategy based on hierarchical use artificial neural networks (ANN) for a first level transient diagnose. After the ANN has done a preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. In order to validate the method, a Nuclear Power Plant transient identification problem, comprising postulated accidents, is proposed. Noisy data was used to evaluate the method robustness. The results obtained reveal the ability of the method in dealing with dynamic identification of transients and its reliability degree. (author)

  1. Secondary cycle design considerations for reduction of reactor transients frequency

    International Nuclear Information System (INIS)

    Bevilacqua, L.; Leal, M.R.L.V.

    1980-01-01

    The secondary cycle systems should not be considered of secondary importance to the pressurized water reactor safety. The advanced design and analysis techniques used for components related to nuclear safety are suggested. (E.G.) [pt

  2. Fast reactor fuel pin behavior analyses in a LOF type transient event

    International Nuclear Information System (INIS)

    Mizuno, Tomoyasu; Koyama, Shin-ichi; Kaito, Takeji; Uwaba, Tomoyuki; Tanaka, Kenya

    2013-06-01

    In order to evaluate integrity limiting parameters of fuel pins during fast reactor core transient events, such as fuel center line temperature and cladding maximum temperature, fuel pin behavior calculations were made using the fast reactor fuel pin performance code CEDAR. The temperature histories of fuel pins during a loss of flow (LOF) type transient events was calculated based on Ross and Stoute type gap conductance model and constant gap conductance model, which is used in a core transient calculation code like HIPRAC. The calculated maximum temperatures of cladding and adjacent coolant channel were lower in the case with Ross and Stoute type model than in the case of constant gap conductance model due to the dynamic change of gap conductance of former case. It is indicated that core transient calculations with constant gap conductance give conservative cladding and coolant temperatures than that with Ross and Stoute type gap conductance model which is thought to be realistic. (author)

  3. An analysis of power transients observed in SPERT I reactors

    International Nuclear Information System (INIS)

    Clancy, B.E.; Connolly, J.W.; Harrington, B.V.

    1976-04-01

    The analytical method described in Part I of this series has been applied to the calculation of spert I transients performed with higher initial moderator temperatures and also to those performed in a highly undermoderated core. Reasonable agreement has been obtained between calculated and experimental burst data. (author)

  4. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  5. Neutronics methods for transient and safety analysis of fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marchetti, Marco

    2017-07-01

    Modeling the evolution of possible or postulated accidents in nuclear reactors is fundamental in designing safe systems. For the next generation of reactors, in particular fast reactors, fuel movement during an accident can, in principle, drive an energetic event. Such is the issue of recriticality. The thermal energy produced during these events will, possibly, be converted into mechanical energy by some mechanisms. For example, the nuclear heat deposited in the fuel could cause fuel vaporization and its subsequent expansion. This movement would accelerate the surrounding sodium: part of the initial energy in the fuel is thus converted into sodium kinetic energy. This mechanical energy will finally be absorbed, in some way or another, by the reactor vessel. Providing an accurate estimate for the maximum mechanical work that any accidental sequence can do onto the reactor vessel is an essential step in designing a reactor containment that would withstand any load generated by any accident. That would assure accident containment, without consequences for the general public. Fast reactor accident modeling is a complicated task. The outcome of an accident is determined by different physical phenomena, all acting at almost the same time. Safety analysts must track all these different phenomena. Multi-physics codes have been developed for this task. They must contain accurate models for fluid-dynamics, neutronics, and structures. This work has to do with neutronics modeling of such accidents. Past and recent analyses have been limited to the approximate description of the neutronic field, for example by using a rough description of the energy and/or of the angular dependence of the neutron flux. In this work, different neutronic solvers are selected and coupled into a general multi-physics code for fast reactor accident analysis. Performances of each of them is then assessed. Some emphasis has been put also in assessing the speed of these solvers for determining the

  6. 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)

  7. Fuel damage during off-normal transients in metal-fueled fast reactors

    International Nuclear Information System (INIS)

    Kramer, J.M.; Bauer, T.H.

    1990-01-01

    Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs

  8. Safety And Transient Analyses For Full Core Conversion Of The Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Luong Ba Vien; Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2011-01-01

    Preparing for full core conversion of Dalat Nuclear Research Reactor (DNRR), safety and transient analyses were carried out to confirm about ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled withdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core response were evaluated under these initial events based on RELAP/Mod3.2 computer code and other supported codes like ORIGEN, MCNP and MACCS2. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed. (author)

  9. Transient fission-product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.; Dickson, L.W.

    1997-12-01

    Sweep-gas experiments performed at AECL's Chalk River Laboratories from 1979 to 1985 have been further analysed to determine the fraction of the gaseous fission-product inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the stable xenon release from companion fuel elements and from a well-documented experimental fuel bundle irradiated in the NRU reactor. The calculated gas release could be matched to the measured values within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. There was also limited information on the fraction of the radioactive iodine that was exposed, but not released, on reactor shutdown. An empirical equation is proposed for calculating this fraction. (author)

  10. Transient fission product release during reactor shutdown and startup

    International Nuclear Information System (INIS)

    Hunt, C.E.L.; Lewis, B.J.

    1995-01-01

    Sweep gas experiments performed at CRL from 1979 to 1985 have been analysed to determine the fraction of the fission product gas inventory that is released on reactor shutdown and startup. Empirical equations were derived and applied to calculate the xenon release from companion fuel elements and from a well documented experimental fuel bundle irradiated in the NRU reactor. The measured gas release could be matched to within about a factor of two for an experimental irradiation with a burnup of 217 MWh/kgU. (author)

  11. NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK

    Directory of Open Access Journals (Sweden)

    Filip Osuský

    2016-12-01

    Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.

  12. Development of a computer code for transients simulation in PWR type reactors

    International Nuclear Information System (INIS)

    Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

    1981-01-01

    A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

  13. A three-dimensional operational transient simulation of the CANDU core with typical reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae; Park, Kyung Seok; Park, Jong Woon [Institute for Advanced Engineering, Taejon (Korea, Republic of)

    1995-07-01

    This paper describes the results of simulation of a CANDU operational transient problem (re-startup after short shutdown) using the Coupled Reactor Kinetics(CRKIN) code developed previously with CANDU Reactor Regulating System (RRS) logic. The performance in the simulation is focused on investigating the behaviours of neutron power and regulating devices in accordance with the changes of xenon concentration following the operation of the RRS.

  14. Molten fuel motion during a fast-reactor overpower transient

    International Nuclear Information System (INIS)

    Kolesar, D.C.; Padilla, A. Jr.; Lewis, C.H.; Waltar, A.E.

    1976-01-01

    Mechanistic models for postfailure fuel behavior during hypothetical transient overpower accidents are currently being developed for incorporation into the MELT accident analysis code. A new model for the fuel-coolant interaction and for the motion of fuel in the coolant channel has been developed and incorporated into the MELT-III code. A major limitation of the mechanistic fuel motion model is its dependence on the uniform interaction region of MELT-III. Consequently, a parallel effort is currently in progress to incorporate a non-uniform interaction region into the MELT code. Combination of the fuel motion and the nonuniform interaction region models will provide the framework for development of a mechanistic fuel plateout/blockage model for transient overpower accidents

  15. A faster reactor transient analysis methodology for PCs

    International Nuclear Information System (INIS)

    Ott, K.O.

    1991-10-01

    The simplified ANL model for LMR transient analysis, in which point kinetics as well as lumped descriptions of the heat transfer equations in all components are applied, is converted from a differential into an integral formulation. All differential balance equations are implicitly solved in terms of convolution integrals. The prompt jump approximation is applied as the strong negative feedback effectively keeps the net reactivity well below prompt critical. After implicit finite differencing of the convolution integrals, the kinetics equation assumes the form of a quadratic equation, the ''quadratic dynamics equation.'' This model forms the basis for GW-BASIC program, LTC, for LMR Transient Calculation program, which can effectively be run on a PC. The GW-BASIC version of the LTC program is described in detail in Volume 2 of this report

  16. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  17. Extensions of the MCNP5 and TRIPOLI4 Monte Carlo codes for transient reactor analysis

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    2013-01-01

    To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branch-less collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires the coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3*3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3*3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail. (authors)

  18. Methodology used to calculate moderator-system heat load at full power and during reactor transients in CANDU reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.

    1998-01-01

    Nine components determine the moderator-system heat load during full-power operation and during a reactor power transient in a CANDU reactor. The components that contribute to the total moderator-system heat load at any time consist of the heat generated in the calandria tubes, guide tubes and reactivity mechanisms, moderator and reflector; the heat transferred from calandria shell, the inner tubesheets and the fuel channels; and the heat gained from moderator pumps and heat lost from piping. The contributions from each of these components will vary with time during a reactor transient. The sources of heat that arise from the deposition of nuclear energy can be divided into two categories, viz., a) the neutronic component (which is directly proportional to neutronic power), which includes neutron energy absorption, prompt-fission gamma absorption and capture gamma absorption; and b) the fission-product decay-gamma component, which also varies with time after initiation of the transient. An equation was derived to calculate transient heat loads to the moderator. The equation includes two independent variables that are the neutronic power and fission-product decay-gamma power fractions during the transient and a constant term that represents the heat gained from moderator pumps and heat lost from piping. The calculated heat load in the moderator during steady-state full-power operation for a CANDU 6 reactor was compared with available measurements from the Point Lepreau, Wolsong 1 and Gentilly-2 nuclear generating stations. The calculated and measured values were in reasonably good agreement. (author)

  19. Digital computer study of nuclear reactor thermal transients during startup of 60-kWe Brayton power conversion system

    Science.gov (United States)

    Jefferies, K. S.; Tew, R. C.

    1974-01-01

    A digital computer study was made of reactor thermal transients during startup of the Brayton power conversion loop of a 60-kWe reactor Brayton power system. A startup procedure requiring the least Brayton system complication was tried first; this procedure caused violations of design limits on key reactor variables. Several modifications of this procedure were then found which caused no design limit violations. These modifications involved: (1) using a slower rate of increase in gas flow; (2) increasing the initial reactor power level to make the reactor respond faster; and (3) appropriate reactor control drum manipulation during the startup transient.

  20. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

    Crawford, D.C.; Swanson, R.W.

    1999-01-01

    The advent of high-burnup fuel implementation in LWRs has generated international interest in high-burnup LWR fuel performance. Recent testing under simulated RIA conditions has demonstrated that certain fuel designs fail at peak fuel enthalpy values that are below existing regulatory criteria. Because many of these tests were performed with non-prototypically aggressive test conditions (i.e., with power pulse widths less than 10 msec FWHM and with non-protoypic coolant configurations), the results (although very informative) do not indisputably identify failure thresholds and fuel behavior. The capability of the TREAT facility to perform simulated RIA tests with prototypic test conditions is currently being evaluated by ANL personnel. TREAT was designed to accommodate test loops and vehicles installed for in-pile transient testing. During 40 years of TREAT operation and fuel testing and evaluation, experimenters have been able to demonstrate and determine the transient behavior of several types of fuel under a variety of test conditions. This experience led to an evolution of test methodology and techniques which can be employed to assess RIA behavior of LWR fuel. A pressurized water loop that will accommodate RIA testing of LWR and CANDU-type fuel has completed conceptual design. Preliminary calculations of transient characteristics and energy deposition into test rods during hypothetical TREAT RIA tests indicate that with the installation of a pressurized water loop, the facility is quite capable of performing prototypic RIA testing. Typical test scenarios indicate that a simulated RIA with a 72 msec FWHM pulse width and energy deposition of 1200 kJ/kg (290 cal/gm) is possible. Further control system enhancements would expand the capability to pulse widths as narrow as 40 msec. (author)

  1. Status report for anticipated transients without scram for Combustion Engineering reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The NRC staff review of Combustion ATWS analyses included the anticipated transients expected to occur, the initial conditions and system parameters assumed in the analyses, the reliability of systems, the analytical techniques, the results of transient analysis of ATWS events and the design of the Reactor Protection System. Using the requirements of WASH-1270 as a guideline, the staff reviewed each relevant aspect of the Combustion model and analysis. The discussion of anticipated transients is presented, and the initial conditions, system parameters, and operating systems assumed in the analyses of these transients are discussed. The analytical techniques and computer programs are reviewed. An independent calculation conducted by the staff using the RELAP-3B code to determine the pressure within the reactor coolant pressure boundary during a complete loss of main feedwater ATWS event is described. A set of standard problems is defined for all pressurized water reactor vendors and the Regulatory staff to insure acceptability of computer codes used in all systems transient analyses. The model for calculating water discharge through primary valves is described. The comparison of the Combustion analyses to the requirements of WASH-1270 is presented. Certain outstanding issues are identified which require that Combustion or the applicant provide additional information or modify existing designs

  2. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

  3. First TREAT [Transient Reactor Test Facility] transient overpower tests on U-Pu-Zr fuel: M5 and M6

    International Nuclear Information System (INIS)

    Robinson, W.R.; Bauer, T.H.; Wright, A.E.; Rhodes, E.A.; Stanford, G.S.; Klickman, A.E.

    1987-01-01

    Transient Reactor Test Facility (TREAT) tests M5 and M6 were the first transient overpower (TOP) tests of the margin to cladding breach and prefailure elongation of metallic U-Pu-Zr ternary fuel, the reference fuel of the Integral Fast Reactor concept. Similar tests on U-Fs fueled EBR-II driver pins were previously performed and reported [1,2]. Results from these earlier tests indicated a margin to failure of about 4 times nominal power and significant axial elongation prior to failure, a feature that was very pronounced at low burnups. While these two fuel types are similar in many respects, the ternary alloy exhibits a much more complex physical structure and is typically irradiated at much higher temperatures. Thus, a prime motivation for performing M5 and M6 was to compare the safety related fuel performance characteristics of U-Fs and U-Pu-Zr. This report described conditions, results, and conclusions of testing of these fuel types

  4. Investigations of anticipated transients without scram (ATWS) for the high temperature reactor

    International Nuclear Information System (INIS)

    Heckhoff, H.D.

    1981-10-01

    In this study anticipated transients without scram (ATWS) are investigated for the high temperature reactor, especially for the thorium high temperature reactor (THTR) 300 MWe as an example. It is shown that the two ATWS 'feedwater flow reduction from full power' and 'positive reactivity insertion of 1 mNile/s from 40 per cent power' are the most important transients for the THTR. The additional load caused by the ATWS can be reduced sufficiently by some small modifications of the afterheat removal system. Supplementary precautions are not necessary. In the last part of this study some possibilities to improve the behaviour of the power plant are shown with regard to high temperature reactors of the future, the partial scram as well as some modifications of heating and cooling of the steam generator. (orig.) [de

  5. Transient thermal-hydraulic simulations of direct cycle gas cooled reactors

    International Nuclear Information System (INIS)

    Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe

    2005-01-01

    This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems

  6. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  7. Application of neural networks to connectional expert system for identification of transients in nuclear power plants

    International Nuclear Information System (INIS)

    Cheon, Se Woo; Kim, Wan Joo; Chang, Soon Heung; Roh, Myung Sub

    1991-01-01

    The Back-propagation Neural Network (BPN) algorithm is applied to connectionist expert system for the identification of BWR transients. Several powerful features of neural network-based expert systems over traditional rule-based expert systems are described. The general mapping capability of the neural networks enables to identify transients easily. A number of case studies were performed with emphasis on the applicability of the neural networks to the diagnostic domain. It is revealed that the BPN algorithm can identify transients properly, even when incomplete or untrained symptoms are given. It is also shown that multiple transients are easily identified

  8. 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

  9. BR2 reactor core steady state transient modeling

    International Nuclear Information System (INIS)

    Makarenko, A.; Petrova, T.

    2000-01-01

    A coupled neutronics/hydraulics/heat-conduction model of the BR2 reactor core is under development at SCK-CEN. The neutron transport phenomenon has been implemented as steady state and time dependent nodal diffusion. The non-linear heat conduction equation in-side fuel elements is solved with a time dependent finite element method. To allow coupling between functional modules and to simulate subcooled regimes, a simple single-phase hydraulics has been introduced, while the two-phase hydraulics is under development. Multiple tests, general benchmark cases as well as calculation/experiment comparisons demonstrated a good accuracy of both neutronic and thermal hydraulic models, numerical reliability and full code portability. A refinement methodology has been developed and tested for better neutronic representation in hexagonal geometry. Much effort is still needed to complete the development of an extended cross section library with kinetic data and two-phase flow representation. (author)

  10. Pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    1992-01-01

    The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude

  11. Transients in a reactor containment after a HCDA

    International Nuclear Information System (INIS)

    Ghosh, A.K.

    1984-01-01

    The consequences of a hypothetical core meltdown accident in a fast reactor is analysed. Shock waves are generated in the surrounding medium after the energy release, which is assumed to be instantaneous and at a point. After discussing the difference in the predicted and experimentally observed peak pressure a semi-empirical approach is taken to arrive at a better estimate. This defines the loading on the containment, which is idealised as a combination of a shell and a plate. To simplify the analysis the shell is assumed to be in plane strain and axial symmetry is assumed for both components. The shell response is evaluated by using elasticity theory. Numerical results are presented for peak overpressure and pressure-time history and dynamic stresses in the containment. (Author) [pt

  12. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  13. Identification of process dynamics. Stability monitoring in BWR type reactors

    International Nuclear Information System (INIS)

    Abrahamsson, P.; Hallgren, P.

    1991-06-01

    Identification of process dynamics is used for stability monitoring in nuclear reactors (Boiling Water Reactor). This report treats the problem of estimating a damping factor and a resonance frequency from the neutron flux as measured in the reactor. A new parametric online method for identification is derived and presented, and is shown to meet the requirements of stability monitoring. The technique for estimating the process parameters is based on a recursive lattice filter algorithm. The problem of time varying parameters and offset, as well as offline experiments and signal processing are treated. All parts are implemented in a realtime program, using the language C. In comparison with earlier identifications, the new way of estimating the damping factor is shown to work well. Estimates of both the damping factor and the resonance frequency show a stable and reliable behavior. Future development and improvements are also indicated. (au)

  14. The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.T.

    1982-01-01

    A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

  15. Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

    International Nuclear Information System (INIS)

    Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

    2013-01-01

    The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

  16. Identification of Chemical Reactor Plant’s Mathematical Model

    OpenAIRE

    Pyakullya, Boris Ivanovich; Kladiev, Sergey Nikolaevich

    2015-01-01

    This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  17. Identification of Chemical Reactor Plant’s Mathematical Model

    Directory of Open Access Journals (Sweden)

    Pyakillya Boris

    2015-01-01

    Full Text Available This work presents a solution of the identification problem of chemical reactor plant’s mathematical model. The main goal is to obtain a mathematical description of a chemical reactor plant from experimental data, which based on plant’s time response measurements. This data consists sequence of measurements for water jacket temperature and information about control input signal, which is used to govern plant’s behavior.

  18. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  19. Transient analysis of ABWR reactor using a best estimate code

    International Nuclear Information System (INIS)

    Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.

    2004-01-01

    Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)

  20. RELAP5 analyses of overcooling transients in a pressurized water reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.

    1983-01-01

    In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models

  1. Macroscopic behavior of fast reactor fuel subjected to simulated thermal transients

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-06-01

    High-speed cinematography has been used to characterize the macroscopic behavior of irradiated and unirradiated fuel subjected to thermal transients prototypical of fast reactor transients. The results demonstrate that as the cladding melts, the fuel can disperse via spallation if the fuel contains in excess of approx. 16 μmoles/gm of fission gas. Once the cladding has melted, the macroscopic behavior (time to failure and dispersive nature) was strongly influenced by the presence of volatile fission products and the heating rate

  2. Automated transient identification in the Dark Energy Survey

    Energy Technology Data Exchange (ETDEWEB)

    Goldstein, D. A. [Univ. of California, Berkeley, CA (United States); Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States). et al.

    2015-08-20

    We describe an algorithm for identifying point-source transients and moving objects on reference-subtracted optical images containing artifacts of processing and instrumentation. The algorithm makes use of the supervised machine learning technique known as Random Forest. We present results from its use in the Dark Energy Survey Supernova program (DES-SN), where it was trained using a sample of 898,963 signal and background events generated by the transient detection pipeline. After reprocessing the data collected during the first DES-SN observing season (2013 September through 2014 February) using the algorithm, the number of transient candidates eligible for human scanning decreased by a factor of 13.4, while only 1.0 percent of the artificial Type Ia supernovae (SNe) injected into search images to monitor survey efficiency were lost, most of which were very faint events. Here we characterize the algorithm's performance in detail, and we discuss how it can inform pipeline design decisions for future time-domain imaging surveys, such as the Large Synoptic Survey Telescope and the Zwicky Transient Facility.

  3. Possibilities of optimizing non-nuclear simulation of pressurized water reactor transients

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1985-01-01

    The GKSS-Forschungszentrum Geesthacht GmbH has instituted the concept of a scaled test facility (volume scale factor of 1/100) of a typical PWR of the 1 300 MWe class for the purpose of studying small breaks Loss-of-Coolant Accidents (LOCA) and transients. Having in mind the goal of an optimization of this concept has been choosen a station blackout with and without reactor shutdown and a small break LOCA in a primary loop piping to investigate the thermohydraulic behaviour of the test facility in comparison to the reactor plant. The computer code RELAP 5/MOD 1 has been utilized to compare the test facility behaviour with the reactor plant one. Recommendations are given for minimization of distortions between test facility and reactor plant. (orig./HP) [de

  4. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1993-01-01

    Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

  5. Thermal-structural response of EBR-II major components under reactor operational transients

    International Nuclear Information System (INIS)

    Chang, L.K.; Lee, M.J.

    1983-01-01

    Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  7. Molten fuel-coolant interactions resulting from power transients in aluminium plate/water moderated reactors

    International Nuclear Information System (INIS)

    Storr, G.J.

    1989-08-01

    The behaviour of two reactors SL1 and SPERT D12, which underwent fast nuclear power transients prior to core destruction by a molten fuel-coolant interaction (MFCI) has been analysed and the results compared with measured data. The calculated spatial melt distribution and the mechanical work done during the events leads to high (∼ 250 kJ/kg) conversion efficiencies for this type of interaction when compared with molten drop experiments. A simple model for the steam explosion, using static thermodynamic properties of high temperature and pressure steam is used to calculate the dynamics of the reactors following the MFCI. 26 refs., 5 figs., 5 tabs

  8. Verification of a neutronic code for transient analysis in reactors with Hex-z geometry

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Pintor, S.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain); Ginestar, D. [Departamento de Matematica Aplicada, Universitat Politecnica de Valencia, Cami de Vera, 14, 46022. Valencia (Spain)

    2012-07-01

    Due to the geometry of the fuel bundles, to simulate reactors such as VVER reactors it is necessary to develop methods that can deal with hexagonal prisms as basic elements of the spatial discretization. The main features of a code based on a high order finite element method for the spatial discretization of the neutron diffusion equation and an implicit difference method for the time discretization of this equation are presented and the performance of the code is tested solving the first exercise of the AER transient benchmark. The obtained results are compared with the reference results of the benchmark and with the results provided by PARCS code. (authors)

  9. Development of a digital card to simulate period transients in research reactors

    International Nuclear Information System (INIS)

    Masotti, Paulo Henrique Ferraz

    1999-01-01

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

  10. On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

    International Nuclear Information System (INIS)

    Freels, J.D.

    1993-01-01

    This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

  11. Enhanced backpropagation training algorithm for transient event identification

    International Nuclear Information System (INIS)

    Vitela, J.; Reifman, J.

    1993-01-01

    We present an enhanced backpropagation (BP) algorithm for training feedforward neural networks that avoids the undesirable premature saturation of the network output nodes and accelerates the training process even in cases where premature saturation is not present. When the standard BP algorithm is applied to train patterns of nuclear power plant (NPP) transients, the network output nodes often become prematurely saturated causing the already slow rate of convergence of the algorithm to become even slower. When premature saturation occurs, the gradient of the prediction error becomes very small, although the prediction error itself is still large, yielding negligible weight updates and hence no significant decrease in the prediction error until the eventual recovery of the output nodes from saturation. By defining the onset of premature saturation and systematically modifying the gradient of the prediction error at saturation, we developed an enhanced BP algorithm that is compared with the standard BP algorithm in training a network to identify NPP transients

  12. Best-estimate methodology for analysis of anticipated transients without scram in pressurized water reactors

    International Nuclear Information System (INIS)

    Rebollo, L.

    1993-01-01

    Union Fenosa, a utility company in Spain, has performed research on pressurized water reactor (PWR) safety with respect to the development of a best-estimate methodology for the analysis of anticipated transients without scram (ATWS), i.e., those anticipated transients for which failure of the reactor protection system is postulated. A scientific and technical approach is adopted with respect to the ATWS phenomenon as it affects a PWR, specifically the Zorita nuclear power plant, a single-loop Westinghouse-designed PWR in Spain. In this respect, an ATWS sequence analysis methodology based on published codes that is generically applicable to any PWR is proposed, which covers all the anticipated phenomena and defines the applicable acceptance criteria. The areas contemplated are cell neutron analysis, core thermal hydraulics, and plant dynamics, which are developed, qualified, and plant dynamics, which are developed, qualified, and validated by comparison with reference calculations and measurements obtained from integral or separate-effects tests

  13. SACI - O: A code for the analysis of transients in a pressurized water reactor core

    International Nuclear Information System (INIS)

    Resende Lobo, A.A. de; Soares, P.A.

    1979-03-01

    The SACI-O digital computer code consists basically of a pressurized water reactor core model. It is useful in the analysis of fast reactivity transients shorter than the loop transit time. The program can also be used for evaluating the core behaviour, during other transients, when the inlet coolant conditions are known. SACI-O uses point model neutron kinetics taking into account moderator and fuel reactivity effects, and fission products decay. The neutronic and thermal-hydraulic equations are solved for an average fuel pin described by a single axial node. To perform a more detailed calculation, the modeling of another cooling channel, which can be divided into axial segments, is included in the program. The reactor trip system is also partially simulated. (Author) [pt

  14. Patterns identification in supervisory systems of nuclear reactors installations and gas pipelines systems using self-organizing maps

    International Nuclear Information System (INIS)

    Doraskevicius Junior, Waldemar

    2005-01-01

    Self-Organizing Maps, SOM, of Kohonen were studied, implemented and tested with the aim of developing, for the energy branch, an effective tool especially for transient identification in nuclear reactors and for gas pipelines networks logistic supervision, by classifying operations and identifying transients or abnormalities. The digital system for the test was developed in Java platform, for the portability and scalability, and for belonging to free development platforms. The system, executed in personal computers, showed satisfactory results to aid in decision taking, by classifying IRIS (International Reactor Innovative and Secure) reactor operation conditions (data from simulator) and by classifying Southeast (owner: TRANSPETRO - Brazil) gas pipeline network. Various adaptations were needed for such business, as new topologies for the output layer of artificial neural network and particular preparation for the input data. (author)

  15. Power transients of Ghana research reactor-1 using PARET/ANL thermal hydraulic code

    International Nuclear Information System (INIS)

    Ampomah-Amoaka, E.; Akaho, E.H.K.; Anim-Sampong, S.; Nyarko, B.J.B.

    2010-01-01

    PARET/ANL(Version 7.3 of 2007) thermal-hydraulic code was used to perform transient analysis of the Ghana Research Reactor-1.The reactivities inserted were 2.1mk and 4mk.The peak power of 5.81kW was obtained for 2.1 mk insertion whereas the peak power for 4mk insertion of reactivity was 92.32kW.These results compare closely with experiments and theoretical studies conducted previously.

  16. Development of the computer code for transient analysis in experimental fast reactor

    International Nuclear Information System (INIS)

    Moreira, M.L.; Sato, E.F.

    1989-01-01

    A calculational model of heat transfer and fluid coolant dynamics, for thermal-hydraulic simulation of the primary system components of a pool type experimental fast breeder reactor, has developed. Programmed in FORTRAN, the SORES code was used to simulate transients as loss of pumping and loss of secondary sodium flow in the EBRII. The SORES results compared with measured data and NATDEMO code results was found to be good. (author) [pt

  17. Transient Behaviour of Superconducting Magnet Systems of Fusion Reactor ITER during Safety Discharge

    Directory of Open Access Journals (Sweden)

    A. M. Miri

    2008-01-01

    Full Text Available To investigate the transient behaviour of the toroidal and poloidal field coils magnet systems of the International Thermonuclear Experimental Reactor during safety discharge, network models with lumped elements are established. Frequency-dependant values of the network elements, that is, inductances and resistances are calculated with the finite element method. That way, overvoltages can be determined. According to these overvoltages, the insulation coordination of coils has to be selected.

  18. Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.

    1978-06-01

    Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results

  19. Method of transient identification based on a possibilistic approach, optimized by genetic algorithm

    International Nuclear Information System (INIS)

    Almeida, Jose Carlos Soares de

    2001-02-01

    This work develops a method for transient identification based on a possible approach, optimized by Genetic Algorithm to optimize the number of the centroids of the classes that represent the transients. The basic idea of the proposed method is to optimize the partition of the search space, generating subsets in the classes within a partition, defined as subclasses, whose centroids are able to distinguish the classes with the maximum correct classifications. The interpretation of the subclasses as fuzzy sets and the possible approach provided a heuristic to establish influence zones of the centroids, allowing to achieve the 'don't know' answer for unknown transients, that is, outside the training set. (author)

  20. Modeling transient thermal hydraulic behavior of a thermionic fuel element for nuclear space reactors

    International Nuclear Information System (INIS)

    Al-Kheliewi, A.S.; Klein, A.C.

    1994-01-01

    A transient code (TFETC) for determining the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) during changes in operating conditions has been successfully developed and tested. A fully implicit method is used to solve the system of equations for temperatures at each time step. Presently, TFETC has the ability to handle the following transients: startup, loss of flow accidents, and shutdown. The code has been applied to the startup of the ATI single cell configuration which appears to start up and shut down in an orderly and reasonable fashion. No unexpected transient features were observed. The TFE also appears to function robustly under loss of flow accident conditions. It appears hat sufficient time is available to shut the reactor down safely without melting point the fuel. The model shows that during a complete loss of flow accident (without shutdown) the coolant reaches its boiling point in approximately 35 seconds. The fuel may exceed its melting point after this time as the NaK coolant will boil if the reactor is not shut down. For 1/2, 1/3, and 1/4 pump failures, the fuel temperatures never exceed the fuel melting point even if the reactor is not shut down

  1. Severe transient analysis of the Penn State University Advanced Light Water Reactor

    International Nuclear Information System (INIS)

    Borkowski, J.A.

    1988-08-01

    The Penn State University Advanced Light Water Reactor (PSU ALWR) incorporates various passive and active ultra-safe features, such as continuous online injection and letdown for pressure control, a raised-loop primary system for enhanced natural circulation, a dedicated primary reservoir for enhanced thermal hydraulic control, and a secondary shutdown turbine. Because of the conceptual design basis of the project, the dynamic system modeling was to be performed using a code with a high degree of flexibility. For this reason the modeling has been performed with the Modular Modeling System (MMS). The basic design and normal transients have been performed successfully with MMS. However, the true test of an inherently safe concept lies in its response to more brutal transients. Therefore, such a demonstrative transient is chosen for the PSU ALWR: a turbine trip and reactor scram, concurrent with total loss of offsite ac power. Diesel generators are likewise unavailable. This transient demonstrates the utility of the pressure control system, the shutdown turbine generator, and the enhanced natural circulation of the PSU ALWR. The low flow rates, low pressure drops, and large derivative states encountered in such a transient pose special problems for the modeler and for MMS. The results of the transient analyses indicate excellent performance by the PSU ALWR in terms of inherently safe operation. The primary coolant enters full natural circulation, and removes all decay heat through the steam generators. Further, the steam generators continually supply sufficient steam to the shutdown power system, despite the abrupt changeover to the auxiliary feedwater system. Finally, even with coincident failures in the pressurization system, the primary repressurizes to near-normal values, without overpressurization. No core boiling or uncovery is predicted, and consequently fuel damage is avoided. 17 refs., 19 figs., 4 tabs

  2. Analysis of reactivity transient for the DIDO type research reactors using RELAP5

    International Nuclear Information System (INIS)

    Adorni, M.; Bousbia-Salah, A.; D'Auria, F.; Nabbi, R.

    2005-01-01

    Recent availability of high performance computers and computational methods together with the continuing increase in operational experience imposes revising some operational constrains and conservative safety margins. The application of Best-Estimate (BE) method constitutes a real necessity in the safety and design analysis and allows getting more realistic simulation of the processes taking place during the steady state operation and transients. In comparison to the conservative approaches, the application of Best-Estimate methods results in the mitigation of the constraining limits in design and operation. This paper presents the results of the application of the RELAP5/Mod3.3 system thermal-hydraulic code to the German FRJ-2 research reactor for a reactivity transient, which has been analyzed in the past using the verified system code CATHENA [1], [2], [3]. The work mainly aims checking the capability of RELAP5 [4] for research reactor transient analysis by the comparison of the results of the two codes and including modeling basis and analytical approaches. According to the existing references RELAP5 applications are concentrated on the transient analysis of nuclear power systems. The considered case consists of a simulation related to a hypothetical fast reactivity transient, which is assumed to be caused by the failure of one shutdown arm. The case has been chosen due to the importance of the models for the precise description of the complex phenomenon of subcooled boiling and two phase flow taking place during the transient. For this purpose, the fuel element assembly was modeled in detail according to design data. The primary circuit was included in the whole model in order to consider the interaction with individual fuel elements with core. In general the results of the two codes are in agreement and comparable during the initial phase of the transient. After reaching the flow regime with fully developed nucleate boiling and two phase flow RELAP5 exhibits

  3. Development of the containment transient analysis code for the passive reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Young Dong; Kim, Young In; Bae, Yoon Young; Chang, Moon Hi [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-05-01

    This study was performed to develop the analysis tools for the passively cooled steel containment and to construct the integrated code system which can analyze a thermal hydraulic behavior of the containment and reactor system during a loss of coolant accident. The computer code CONTEMPT4/MOD5/PCCS was developed by incorporating the passive containment cooling models to the containment pressure and temperature transient analysis computer code CONTEMPT4/MOD5. The integrated reactor thermal hydraulic analysis code system for passive reactor was constructed by coupling the best estimate thermal hydraulic system analysis code RELAP5/MOD3 and CONTEMPT4/MOD5/PCCS through the process control method. In addition, to evaluate the applicability of the code the CONTEMPT4/MOD5/PCCS was applied to the SMART(System-Integrated Modular Advanced Reactor). The pressure and temperature transient following the small break LOCA of SMART was analysed by modeling the safeguard vessel using both the newly added passive containment cooling model and existing pool model. (author). 16 refs., 22 figs., 7 tabs.

  4. Study of Fast Transient Pressure Drop in VVER-1000 Nuclear Reactor Using Acoustic Phenomenon

    Directory of Open Access Journals (Sweden)

    Soroush Heidari Sangestani

    2018-01-01

    Full Text Available This article aims to simulate the sudden and fast pressure drop of VVER-1000 reactor core coolant, regarding acoustic phenomenon. It is used to acquire a more accurate method in order to simulate the various accidents of reactor core. Neutronic equations should be solved concurrently by means of DRAGON 4 and DONJON 4 coupling codes. The results of the developed package are compared with WIMS/CITATION and final safety analysis report of Bushehr VVER-1000 reactor (FSAR. Afterwards, time dependent thermal-hydraulic equations are answered by employing Single Heated Channel by Sectionalized Compressible Fluid method. Then, the obtained results were validated by the same transient simulation in a pressurized water reactor core. Then, thermal-hydraulic and neutronic modules are coupled concurrently by use of producing group constants regarding the thermal feedback effect. Results were compared to the mentioned transient simulation in RELAP5 computer code, which show that mass flux drop is sensed at the end of channel in several milliseconds which causes heat flux drop too. The thermal feedback resulted in production of some perturbations in the changes of these parameters. The achieved results for this very fast pressure drop represent accurate calculations of thermoneutronic parameters fast changes.

  5. Comparison and analysis on transient characteristics of integral pressurized water reactors

    International Nuclear Information System (INIS)

    Zhang, Guoxu; Xie, Heng

    2017-01-01

    Highlights: • Two IPWR Relap5 models with different PSS design were developed. • Postulated SBO and SBLOCA were analyzed. • PRHRS in primary PSS design showed stable performance under different scenarios. • Secondary PRHRS design faced flow instability. - Abstract: In the present work, the similarities and differences of representative IPWRs (integral pressurized water reactor) are studied, and two typical reactor design schemes are summarized. To get a comprehensive understanding of their transient characteristics, SBO (station blackout) and SBLOCA (small break LOCA) are simulated and analyzed respectively by using Relap5/Mod3.2. The calculation results show that, both designs are effective in keeping reactor safe. However, the transient features of the two designs show significant differences. In the primary side passive safety system (PSS) connection design, PRHRS (passive residual heat removal system) shows a roughly congruent performance in removing residual heat under various accidents. While in secondary side PSS connection design, the capability of PRHRS is closely related to primary coolant circulation condition. In SBLOCA analysis, different design approach shows different primary coolant water inventory change trend. And primary PSS connection design could potentially keep reactor core well covered for a longer time.

  6. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1992-01-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

  7. Modeling the behavior of metallic fast reactor fuels during extended transients

    International Nuclear Information System (INIS)

    Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

    1992-01-01

    Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements

  8. Research of three-dimensional transient reactivity feedback in fast reactor

    International Nuclear Information System (INIS)

    Xu Li; Shi Gong; Ma Dayuan; Yu Hong

    2013-01-01

    To solve the three-dimensional time-spatial kinetics feedback problems in fast reactor, a mathematical model of the direct reactivity feedback was proposed. Based on the NAS code for fast reactor and the reactivity feedback mechanism, a feedback model which combined the direct reactivity feedback and feedback reflected by the cross section variation was provided for the transient calculation. Furthermore, the fast reactor group collapsing system was added to the code, thus the real time group collapsing calculation could be realized. The isothermal elevated temperature test of CEFR was simulated by using the code. By comparing the calculation result with the test result of the temperature reactivity coefficient, the validity of the model and the code is verified. (authors)

  9. Performance of fast reactor mixed-oxide fuels pins during extended overpower transients

    International Nuclear Information System (INIS)

    Tsai, H.; Neimark, L.A.; Asaga, T.; Shikakura, S.

    1991-02-01

    The Operational Reliability Testing (ORT) program, a collaborative effort between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corp. (PNC) of Japan, was initiated in 1982 to investigate the behavior of mixed-oxide fuel pin under various slow-ramp transient and duty-cycle conditions. In the first phase of the program, a series of four extended overpower transient tests, with severity sufficient to challenge the pin cladding integrity, was conducted. The objectives of the designated TOPI-1A through -1D tests were to establish the cladding breaching threshold and mechanisms, and investigate the thermal and mechanical effects of the transient on pin behavior. The tests were conducted in EBR-2, a normally steady-state reactor. The modes of transient operation in EBR-2 were described in a previous paper. Two ramp rates, 0.1%/s and 10%/s, were selected to provide a comparison of ramp-rate effects on fuel behavior. The test pins chosen for the series covered a range of design and pre-test irradiation parameters. In the first test (1A), all pins maintained their cladding integrity during the 0.1%/s ramp to 60% peak overpower. Fuel pins with aggressive designs, i.e., high fuel- smear density and/or thin cladding, were, therefore, included in the follow-up 1B and 1C tests to enhance the likelihood of achieving cladding breaching. In the meantime, a higher pin overpower capability, to greater than 100%, was established by increasing the reactor power limit from 62.5 to 75 MWt. In this paper, the significant results of the 1B and 1C tests are presented. 4 refs., 5 figs., 1 tab

  10. Transient plasma estimation: a noise cancelling/identification approach

    International Nuclear Information System (INIS)

    Candy, J.V.; Casper, T.; Kane, R.

    1985-03-01

    The application of a noise cancelling technique to extract energy storage information from sensors occurring during fusion reactor experiments on the Tandem Mirror Experiment-Upgrade (TMX-U) at the Lawrence Livermore National Laboratory (LLNL) is examined. We show how this technique can be used to decrease the uncertainty in the corresponding sensor measurements used for diagnostics in both real-time and post-experimental environments. We analyze the performance of algorithm on the sensor data and discuss the various tradeoffs. The algorithm suggested is designed using SIG, an interactive signal processing package developed at LLNL

  11. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  12. A pattern recognition approach based on DTW for automatic transient identification in nuclear power plants

    International Nuclear Information System (INIS)

    Galbally, Javier; Galbally, David

    2015-01-01

    Highlights: • Novel transient identification method for NPPs. • Low-complexity. • Low training data requirements. • High accuracy. • Fully reproducible protocol carried out on a real benchmark. - Abstract: Automatic identification of transients in nuclear power plants (NPPs) allows monitoring the fatigue damage accumulated by critical components during plant operation, and is therefore of great importance for ensuring that usage factors remain within the original design bases postulated by the plant designer. Although several schemes to address this important issue have been explored in the literature, there is still no definitive solution available. In the present work, a new method for automatic transient identification is proposed, based on the Dynamic Time Warping (DTW) algorithm, largely used in other related areas such as signature or speech recognition. The novel transient identification system is evaluated on real operational data following a rigorous pattern recognition protocol. Results show the high accuracy of the proposed approach, which is combined with other interesting features such as its low complexity and its very limited requirements of training data

  13. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  14. CEDNBR: a computer code for transient thermal margin analysis of a reactor core

    International Nuclear Information System (INIS)

    Shesler, A.T.; Lehmann, C.R.

    1976-09-01

    The report describes the CEDNBR computer code. This code was developed for the transient thermal analysis of a pressurized water reactor core or a critical heat flux test. Included are the code structure, conservation equations, and correlations utilized by CEDNBR. The methods of modelling a reactor core and hot channel and a CHF test are presented. Comparisons of CEDNBR calculations are made with both empirical pressure loss data and simulated loss of flow test data. The code solves the one-dimensional conservation of mass, energy, and momentum equations and the equation of state for the fluid for either steady-state or transient conditions. Tabular time dependent functions of inlet temperatures, pressure, mass velocity, axial heat flux distributions, normalized heat flux, radial peaking factors, and incremental mixing factors are required input to the code. Transient effects are included in the calculation of enthalpy rise and fluid properties. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by applying a Critical Heat Flux (CHF) correlation to the computed local fluid properties. A code user's guide is provided for preparing input to the code. In addition, descriptions of the sub-routines used by CEDNBR are given

  15. 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)

  16. Beam transient analyses of Accelerator Driven Subcritical Reactors based on neutron transport method

    Energy Technology Data Exchange (ETDEWEB)

    He, Mingtao; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Zheng, Youqi, E-mail: yqzheng@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China); Li, Xunzhao; Zhou, Shengcheng [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, Shaanxi (China)

    2015-12-15

    Highlights: • A transport-based kinetics code for Accelerator Driven Subcritical Reactors is developed. • The performance of different kinetics methods adapted to the ADSR is investigated. • The impacts of neutronic parameters deteriorating with fuel depletion are investigated. - Abstract: The Accelerator Driven Subcritical Reactor (ADSR) is almost external source dominated since there is no additional reactivity control mechanism in most designs. This paper focuses on beam-induced transients with an in-house developed dynamic analysis code. The performance of different kinetics methods adapted to the ADSR is investigated, including the point kinetics approximation and space–time kinetics methods. Then, the transient responds of beam trip and beam overpower are calculated and analyzed for an ADSR design dedicated for minor actinides transmutation. The impacts of some safety-related neutronics parameters deteriorating with fuel depletion are also investigated. The results show that the power distribution varying with burnup leads to large differences in temperature responds during transients, while the impacts of kinetic parameters and feedback coefficients are not very obvious. Classification: Core physic.

  17. Investigation of Natural Circulation Instability and Transients in Passively Safe Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, Mamoru [Purdue Univ., West Lafayette, IN (United State

    2016-11-30

    The NEUP funded project, NEUP-3496, aims to experimentally investigate two-phase natural circulation flow instability that could occur in Small Modular Reactors (SMRs), especially for natural circulation SMRs. The objective has been achieved by systematically performing tests to study the general natural circulation instability characteristics and the natural circulation behavior under start-up or design basis accident conditions. Experimental data sets highlighting the effect of void reactivity feedback as well as the effect of power ramp-up rate and system pressure have been used to develop a comprehensive stability map. The safety analysis code, RELAP5, has been used to evaluate experimental results and models. Improvements to the constitutive relations for flashing have been made in order to develop a reliable analysis tool. This research has been focusing on two generic SMR designs, i.e. a small modular Simplified Boiling Water Reactor (SBWR) like design and a small integral Pressurized Water Reactor (PWR) like design. A BWR-type natural circulation test facility was firstly built based on the three-level scaling analysis of the Purdue Novel Modular Reactor (NMR) with an electric output of 50 MWe, namely NMR-50, which represents a BWR-type SMR with a significantly reduced reactor pressure vessel (RPV) height. The experimental facility was installed with various equipment to measure thermalhydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests were performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The control system and data acquisition system were programmed with LabVIEW to realize the realtime control and data storage. The thermal-hydraulic and nuclear coupled startup transients were performed to investigate the flow instabilities at low pressure and low power conditions for NMR-50. Two different power ramps were chosen to study the effect of startup

  18. Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

    International Nuclear Information System (INIS)

    Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

    2010-01-01

    In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

  19. Transient bowing of core assemblies in advanced liquid metal fast reactors

    International Nuclear Information System (INIS)

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety

  20. MASTER-2.0: Multi-purpose analyzer for static and transient effects of reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Byung Oh; Song, Jae Seung; Joo, Han Gyu [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-01-01

    MASTER-2.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the two group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM(Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with AFEN/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. Master-2.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P model can be used selectively. In addition, MASTER-2.0 is designed to cover various PWRs including SMART as well as WH-and CE-type reactors, providing all data required in their design procedures. (author). 39 refs., 12 figs., 4 tabs.

  1. Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

    International Nuclear Information System (INIS)

    Huda, M.Q.; Bhuiyan, S.I.; Mondal, M.A.W.

    1996-12-01

    Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76 o C which is much less than the limiting maximum value of fuel temperature of 1150 0 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000 o C, the calculated outer cladding wall temperature was observed to be 702.39 0 C compared to a value of 760 o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

  2. Hybrid parallel strategy for the simulation of fast transient accidental situations at reactor scale

    International Nuclear Information System (INIS)

    Faucher, V.; Galon, P.; Beccantini, A.; Crouzet, F.; Debaud, F.; Gautier, T.

    2015-01-01

    Highlights: • Reference accidental situations for current and future reactors are considered. • They require the modeling of complex fluid–structure systems at full reactor scale. • EPX software computes the non-linear transient solution with explicit time stepping. • Focus on the parallel hybrid solver specific to the proposed coupled equations. - Abstract: This contribution is dedicated to the latest methodological developments implemented in the fast transient dynamics software EUROPLEXUS (EPX) to simulate the mechanical response of fully coupled fluid–structure systems to accidental situations to be considered at reactor scale, among which the Loss of Coolant Accident, the Core Disruptive Accident and the Hydrogen Explosion. Time integration is explicit and the search for reference solutions within the safety framework prevents any simplification and approximations in the coupled algorithm: for instance, all kinematic constraints are dealt with using Lagrange Multipliers, yielding a complex flow chart when non-permanent constraints such as unilateral contact or immersed fluid–structure boundaries are considered. The parallel acceleration of the solution process is then achieved through a hybrid approach, based on a weighted domain decomposition for distributed memory computing and the use of the KAAPI library for self-balanced shared memory processing inside subdomains

  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. Master-3.0: multi-purpose analyzer for static and transient effects of reactors

    International Nuclear Information System (INIS)

    Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

    2002-03-01

    MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

  5. Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

    International Nuclear Information System (INIS)

    Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

    1999-02-01

    Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

  6. Thermal-Hydraulic Analyses of Transients in an Actinide-Burner Reactor Cooled by Forced Convection of Lead Bismuth

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee

    2003-09-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) and the Massachusetts Institute of Technology (MIT) are investigating the suitability of lead or lead–bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The current analysis evaluated a pool type design that relies on forced circulation of the primary coolant, a conventional steam power conversion system, and a passive decay heat removal system. The ATHENA computer code was used to simulate various transients without reactor scram, including a primary coolant pump trip, a station blackout, and a step reactivity insertion. The reactor design successfully met identified temperature limits for each of the transients analyzed.

  7. The effect of retarding torque during a flow transient for Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Farhadi, Kazem, E-mail: kfarhadi@aeoi.org.ir [Engineering Science Research Group, Nuclear Science Research School, Nuclear Science and Technology Research Institute, AEOI, P.O. Box 11365-3486, Tehran (Iran, Islamic Republic of)

    2011-02-15

    The primary cooling system of the Tehran Research Reactor (TRR) has been analysed for a possible flow transient phenomenon caused by power cut-off. All the components of the TRR primary cooling loop that offer resistance to the coolant flow are physically modelled. Differential equations of motion for the coolant in the primary piping of the TRR and for the rotating parts of the centrifugal pump are then derived. The equation of flow motion is solved simultaneously with momentum conservation equation of the rotating parts of the pump which predicts the TRR pump speed during the flow transient. Electrical and mechanical losses are measured for the TRR three-phase induction motor in order to calculate the motor retarding torque during the event. The results of the present study are compared with the other similar primary loop results. The present model shows good agreement with the existing experimental and theoretical studies.

  8. RAP-2A Computer code for transients analysis in fast reactors

    International Nuclear Information System (INIS)

    Iftode, I.; Popescu, C.; Turcu, I.; Biro, L.

    1975-10-01

    The RAP-2A computer code is designed for analyzing thermohydraulic transients and/or steady state problems for large LMFBR cores. Physical and mathematical models, main input-output data, the flow chart of the code and a sample problem are given. RAP-2A calculates the power and the thermoydraulic transients initiated by a flow or reactivity changes, from a normal operating state of the reactor up to core disassembly. In this analysis a representative fuel pin is considered: a one-group space-independent (point) kinetics model to describe the neutron kinetics and a one-dimensional model describing the heat transfer (radial in the fuel and axial in the coolant) are used. Mechanical deformations due to temperature gradient, pressure losses, fuel melting, etc., are also calculated. The code is written in FORTRAN-4 language and is running on a IBM-370/135 computer

  9. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  10. Statistical techniques for the identification of reactor component structural vibrations

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    1975-01-01

    The identification, on-line and in near real-time, of the vibration frequencies, modes and amplitudes of selected key reactor structural components and the visual monitoring of these phenomena by nuclear power plant operating staff will serve to further the safety and control philosophy of nuclear systems and lead to design optimisation. The School of Nuclear Engineering has developed a data acquisition system for vibration detection and identification. The system is interfaced with the HIFAR research reactor of the Australian Atomic Energy Commission. The reactor serves to simulate noise and vibrational phenomena which might be pertinent in power reactor situations. The data acquisition system consists of a small computer interfaced with a digital correlator and a Fourier transform unit. An incremental tape recorder is utilised as a backing store and as a means of communication with other computers. A small analogue computer and an analogue statistical analyzer can be used in the pre and post computational analysis of signals which are received from neutron and gamma detectors, thermocouples, accelerometers, hydrophones and strain gauges. Investigations carried out to date include a study of the role of local and global pressure fields due to turbulence in coolant flow and pump impeller induced perturbations on (a) control absorbers, (B) fuel element and (c) coolant external circuit and core tank structure component vibrations. (Auth.)

  11. Location identification of closed crack based on Duffing oscillator transient transition

    Science.gov (United States)

    Liu, Xiaofeng; Bo, Lin; Liu, Yaolu; Zhao, Youxuan; Zhang, Jun; Deng, Mingxi; Hu, Ning

    2018-02-01

    The existence of a closed micro-crack in plates can be detected by using the nonlinear harmonic characteristics of the Lamb wave. However, its location identification is difficult. By considering the transient nonlinear Lamb under the noise interference, we proposed a location identification method for the closed crack based on the quantitative measurement of Duffing oscillator transient transfer in the phase space. The sliding short-time window was used to create a window truncation of to-be-detected signal. And then, the periodic extension processing for transient nonlinear Lamb wave was performed to ensure that the Duffing oscillator has adequate response time to reach a steady state. The transient autocorrelation method was used to reduce the occurrence of missed harmonic detection due to the random variable phase of nonlinear Lamb wave. Moreover, to overcome the deficiency in the quantitative analysis of Duffing system state by phase trajectory diagram and eliminate the misjudgment caused by harmonic frequency component contained in broadband noise, logic operation method of oscillator state transition function based on circular zone partition was adopted to establish the mapping relation between the oscillator transition state and the nonlinear harmonic time domain information. Final state transition discriminant function of Duffing oscillator was used as basis for identifying the reflected and transmitted harmonics from the crack. Chirplet time-frequency analysis was conducted to identify the mode of generated harmonics and determine the propagation speed. Through these steps, accurate position identification of the closed crack was achieved.

  12. Reliability of fast reactor mixed-oxide fuel during operational transients

    International Nuclear Information System (INIS)

    Boltax, A.; Neimark, L.A.; Tsai, Hanchung; Katsuragawa, M.; Shikakura, S.

    1991-07-01

    Results are presented from the cooperative DOE and PNC Phase 1 and 2 operational transient testing programs conducted in the EBR-2 reactor. The program includes second (D9 and PNC 316 cladding) and third (FSM, AST and ODS cladding) generation mixed-oxide fuel pins. The irradiation tests include duty cycle operation and extended overpower tests. the results demonstrate the capability of second generation fuel pins to survive a wide range of duty cycle and extended overpower events. 15 refs., 9 figs., 4 tabs

  13. Modeling of boron control during power transients in a pressurized water reactor

    International Nuclear Information System (INIS)

    Mathieu, P.; Distexhe, E.

    1986-01-01

    Accurate control instructions in a reactor control aid computer are included in order to realize the boron makeup throughput, which is required to obtain the boron concentration in the primary coolant loop, predicted by a neutronic code. A modeling of the transfer function between the makeup and the primary loop is proposed. The chemical and volumetric control system, the pressurizer, and the primary loop are modeled as instantaneous diffusion cells. The pipes are modeled as time lag lines. The model provides the unstationary boron distributions in the different elements of the setup. A numerical code is developed to calculate the time evolutions of the makeup throughput during power transients

  14. Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients

    International Nuclear Information System (INIS)

    Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

    1980-01-01

    The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations

  15. Transient Analyses for a Molten Salt Transmutation Reactor Using the Extended SIMMER-III Code

    International Nuclear Information System (INIS)

    Wang, Shisheng; Rineiski, Andrei; Maschek, Werner; Ignatiev, Victor

    2006-01-01

    Recent developments extending the capabilities of the SIMMER-III code for the dealing with transient and accidents in Molten Salt Reactors (MSRs) are presented. These extensions refer to the movable precursor modeling within the space-time dependent neutronics framework of SIMMER-III, to the molten salt flow modeling, and to new equations of state for various salts. An important new SIMMER-III feature is that the space-time distribution of the various precursor families with different decay constants can be computed and took into account in neutron/reactivity balance calculations and, if necessary, visualized. The system is coded and tested for a molten salt transmuter. This new feature is also of interest in core disruptive accidents of fast reactors when the core melts and the molten fuel is redistributed. (authors)

  16. 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

  17. Thermal-hydraulics of the Loviisa reactor pressure vessel overcooling transients

    International Nuclear Information System (INIS)

    Tuomisto, Harri.

    1987-06-01

    In the Loviisa reactor pressure vessel safety analyses, the thermal-hydraulics of various overcooling transients has been evaluated to give pertinent initial data for fracture-mechanics calculations. The thermal-hydraulic simulations of the developed overcooling scenarios have been performed using best-estimate thermal-hydraulic computer codes. Experimental programs have been carried out to study phenomena related to natural circulation interruptions in the reactor coolant system. These experiments include buoyancy-induced phenomena such as thermal mixing and stratification of cold high-pressure safety injection water in the cold legs and the downcomer, and oscillations of the single-phase natural circulation. In the probabilistic pressurized thermal shock study, the Loviisa training simulator and the advanced system code RELAP5/MOD2 were utilized to simulate selected sequences. Flow stagnation cases were separately calculated with the REMIX computer program. The methods employed were assessed for these calculations against the plant data and own experiments

  18. VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

    Directory of Open Access Journals (Sweden)

    Muhammad Darwis Isnaini

    2016-03-01

    Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

  19. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

  20. Investigation on in-vessel thermal transients in a fast breeder reactor

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Kasahara, Naoto

    1999-01-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

  1. 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

  2. Investigation of natural circulation instability and transients in passively safe novel modular reactor

    Science.gov (United States)

    Shi, Shanbin

    The Purdue Novel Modular Reactor (NMR) is a new type small modular reactor (SMR) that belongs to the design of boiling water reactor (BWR). Specifically, the NMR is one third the height and area of a conventional BWR reactor pressure vessel (RPV) with an electric output of 50 MWe. The fuel cycle length of the NMR-50 is extended up to 10 years due to optimized neutronics design. The NMR-50 is designed with double passive engineering safety system. However, natural circulation BWRs (NCBWR) could experience certain operational difficulties due to flow instabilities that occur at low pressure and low power conditions. Static instabilities (i.e. flow excursion (Ledinegg) instability and flow pattern transition instability) and dynamic instabilities (i.e. density wave instability and flashing/condensation instability) pose a significant challenge in two-phase natural circulation systems. In order to experimentally study the natural circulation flow instability, a proper scaling methodology is needed to build a reduced-size test facility. The scaling analysis of the NMR uses a three-level scaling method, which was developed and applied for the design of the Purdue Multi-dimensional Integral Test Assembly (PUMA). Scaling criteria is derived from dimensionless field equations and constitutive equations. The scaling process is validated by the RELAP5 analysis for both steady state and startup transients. A new well-scaled natural circulation test facility is designed and constructed based on the scaling analysis of the NMR-50. The experimental facility is installed with different equipment to measure various thermal-hydraulic parameters such as pressure, temperature, mass flow rate and void fraction. Characterization tests are performed before the startup transient tests and quasi-steady tests to determine the loop flow resistance. The controlling system and data acquisition system are programmed with LabVIEW to realize the real-time control and data storage. The thermal

  3. 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

  4. Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

    International Nuclear Information System (INIS)

    Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

    2013-01-01

    Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

  5. Design base transient analysis using the real-time nuclear reactor simulator model

    International Nuclear Information System (INIS)

    Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.

    1987-01-01

    A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and moderator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes

  6. Design base transient analysis using the real-time nuclear reactor simulator model

    International Nuclear Information System (INIS)

    Tien, K.K.; Yakura, S.J.; Morin, J.P.; Gregory, M.V.

    1987-01-01

    A real-time simulation model has been developed to describe the dynamic response of all major systems in a nuclear process reactor. The model consists of a detailed representation of all hydraulic components in the external coolant circulating loops consisting of piping, valves, pumps and heat exchangers. The reactor core is described by a three-dimensional neutron kinetics model with detailed representation of assembly coolant and mode-rator thermal hydraulics. The models have been developed to support a real-time training simulator, therefore, they reproduce system parameters characteristic of steady state normal operation with high precision. The system responses for postulated severe transients such as large pipe breaks, loss of pumping power, piping leaks, malfunctions in control rod insertion, and emergency injection of neutron absorber are calculated to be in good agreement with reference safety analyses. Restrictions were imposed by the requirement that the resulting code be able to run in real-time with sufficient spare time to allow interfacing with secondary systems and simulator hardware. Due to hardware set-up and real plant instrumentation, simplifications due to symmetry were not allowed. The resulting code represents a coarse-node engineering model in which the level of detail has been tailored to the available computing power of a present generation super-minicomputer. Results for several significant transients, as calculated by the real-time model, are compared both to actual plant data and to results generated by fine-mesh analysis codes

  7. CATHARE simulation of transients for the 2400 MW gas fast reactor concept

    International Nuclear Information System (INIS)

    Bentivoglio, Fabrice; Messie, Anne; Geffraye, Genevieve; Malo, Jean-Yves; Bertrand, Frederic; Plancq, David

    2009-01-01

    The Gas cooled Fast Reactor (GFR) is one of the six reactor concepts selected in the framework of the Generation IV forum and a high priority in the French Commissariat a l'Energie Atomique (CEA) R and D program on the Future Nuclear Energy Systems. A first design of this GFR2400 reactor has been completed by the CEA at the end of year 2005. The main characteristics of the concept are a 2400MW core based on plate type fuel elements, with an inlet temperature of 400degC and an outlet temperature of 850degC. The power conversion system is based on an indirect combined cycle with helium on the primary circuit, a Brayton cycle with a mixture of nitrogen and helium on the secondary circuit and a steam cycle on the tertiary circuit. In accidental situations, the use of the gas coolant circulation as the main way to remove the decay heat has been selected. A specific system (DHR system) has been designed: it consists of three loops (3 * 100% redundancy) in extension of the pressure vessel, equipped with heat exchangers and blowers. Between 2006 and 2007 a pre-conceptual study has been achieve, leading to the CEA milestone project of the 'GFR viability' at the end of year 2007. In the frame of this milestone, a wide range of CATHARE2 transients has been achieved to consolidate and improve the decay heat removal strategy; in particular the DHR blowers working on a large pressure range and the use of natural convection as a second way to remove decay heat. The paper first presents the CATHARE2 code applied to gas cooled reactor, focusing on the dedicated features included in the standard option of the code in order to obtain a multi-fluid reliable and performing tool. Then the modeling of the GFR2400 is presented, including the core, the vessel, the primary and secondary circuit with the turbo-machine, and a simplified tertiary circuit with boundary conditions. The decay heat removal loops (DHR loop) are also modeled, with a first circuit in helium and a secondary circuit in

  8. Web-based, Interactive, Nuclear Reactor Transient Analyzer using LabVIEW and RELAP5 (ATHENA)

    International Nuclear Information System (INIS)

    Kim, K. D.; Chung, B. D.; Rizwan-uddin

    2006-01-01

    In nuclear engineering, large system analysis codes such as RELAP5, TRAC-M, etc. play an important role in evaluating a reactor system behavior during a wide range of transient conditions. One limitation that restricts their use on a wider scale is that these codes often have a complicated I/O structure. This has motivated the development of GUI tools for best estimate codes, such as SNAP and ViSA, etc. In addition to a user interface, a greater degree of freedom in simulation and analyses of nuclear transient phenomena can be achieved if computer codes and their outputs are accessible from anywhere through the web. Such a web-based interactive interface can be very useful for geographically distributed groups when there is a need to share real-time data. Using mostly off-the-shelf technology, such a capability - a web-based transient analyzer based on a best-estimate code - has been developed. Specifically, the widely used best-estimate code RELAP5 is linked with a graphical interface. Moreover, a capability to web-cast is also available. This has been achieved by using the LabVIEW virtual instruments (VIs). In addition to the graphical display of the results, interactive control functions have also been added that allow operator's actions as well as, if permitted, by a distant user through the web

  9. The analysis with the code TANK of a postulated reactivity-insertion transient in a 10-MW MAPLE research reactor

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-10-01

    This report discusses the analysis of a postulated loss-of-regulation (LOR) accident in a metal-fuelled MAPLE Research Reactor. The selected transient scenario involves a slow LOR from low reactor power; the control rods are assumed to withdraw slowly until a trip at 12 MW halts the withdrawal. The simulation was performed using the space-time reactor kinetics computer code TANK, and modelling the reactor in detail in two dimensions and in two neutron-energy groups. Emphasis in this report is placed on the modelling techniques used in TANK and the physics considerations of the analysis

  10. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  11. Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

    International Nuclear Information System (INIS)

    Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

    2014-01-01

    The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

  12. Identification of the reduced order models of a BWR reactor

    International Nuclear Information System (INIS)

    Hernandez S, A.

    2004-01-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  13. Transient response of small molten salt reactor at duct blockage accident

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi; Ikeuchi, Koji; Suzuki, Takashi

    2005-01-01

    This paper performed transient core analysis of a small Molten Salt Reactor (MSR) at the time of a duct blockage accident. The numerical model employed in this study consists of continuity and momentum conservation equations for fuel salt flow, two group diffusion equations for fast and thermal neutron fluxes, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The analysis shows that (1) the effective multiplication factor and reactor power after the blockage accident hardly change because of the self-control performance of the MSR, (2) fuel salt and graphite moderator temperatures rise at the blockage point and its vicinity, drastically but locally, (3) the highest temperature after the blockage accident is 1 363 K, very lower than the boiling point of fuel salt and melt point of reactor vessel, (4) fast and thermal neutron fluxes distributions after the blockage accident hardly change, and (5) delayed neutron precursors accumulate at the blockage point, especially 1st delayed neutron precursor due to is large decay constant. These results lead that the safety of MSR is assured in the blockage accident. (author)

  14. Simulation of corrosion product activity in pressurized water reactors under flow rate transients

    International Nuclear Information System (INIS)

    Mirza, Anwar M.; Mirza, Nasir M.; Mir, Imran

    1998-01-01

    Simulation of coolant activation due to corrosion products and impurities in a typical pressurized water reactor has been done under flow rate transients. Employing time dependent production and losses of corrosion products in the primary coolant path an approach has been developed to calculate the coolant specific activity. Results for 24 Na, 56 Mn, 59 Fe, 60 Co and 99Mo show that the specific activity in primary loop approaches equilibrium value under normal operating conditions fairly rapidly. Predominant corrosion product activity is due to Mn-56. Parametric studies at full power for various ramp decreases in flow rate show initial decline in the activity and then a gradual rise to relatively higher saturation values. The minimum value and the time taken to reach the minima are strong functions of the slope of linear decrease in flow rate. In the second part flow rate coastdown was allowed to occur at different flow half-times. The reactor scram was initiated at 90% of the normal flow rate. The results show that the specific activity decreases and the rate of decrease depends on pump half time and the reactor scram conditions

  15. Analysis of Precooling Injection Transient of Steam Generator for High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Yan Wang

    2017-01-01

    Full Text Available After a postulated design basis accident leads high temperature gas cooled reactor to emergency shutdown, steam generator still remains with high temperature level and needs to be cooled down by a precooling before reactor restarts with clearing of fault. For the large difference of coolant temperature between inlet and outlet of steam generator in normal operation, the temperature distribution on the components of steam generator is very complicated. Therefore, the temperature descending rate of the components in steam generator needs to be limited to avoid the potential damage during the precooling stage. In this paper, a pebble-bed high temperature gas cooled reactor is modeled by thermal-hydraulic system analysis code and several postulated precooling injection transients are simulated and compared to evaluate their effects, which will provide support for the precooling design. The analysis results show that enough precooling injection is necessary to satisfy the precooling requirements, and larger mass flow rate of precooling water injection will accelerate the precooling process. The temperature decrease of steam generator is related to the precooling injection scenarios, and the maximal mass flow rate of the precooling injection should be limited to avoid the excessively quick temperature change of the structures in steam generator.

  16. Numerical solution for identification of feedback coefficients in nuclear reactors

    International Nuclear Information System (INIS)

    Ebizuka, Yoshie; Sakai, Hideo

    1975-01-01

    Quasilinearization technique was studied to determine the Kinetic parameters of nuclear reactors. The method of solution was generalized to the determination of the parameters contained in a nonlinear system with nonlinear boundary conditions. A computer program, SNR-3, was developed to solve the resulting nonlinear two-point boundary value equations with generalized boundary conditions. In this paper, the problem formulation and the method of solution are explained for a general type of time dependent problem. A flow chart shows the procedure of numerical solution. The method was then applied to the determination of the critical factor and the reactivity feedback coefficients of reactors to investigate the accuracy and the applicability of the present method. The results showed that the present method was considerably successful, but that the random observation error effected the results of the identification. (Aoki, K.)

  17. Vibration analysis of primary inlet pipe line during steady state and transient conditions of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Ayazuddin, S.K.; Qureshi, A.A.; Hayat, T.

    1997-11-01

    The Primary Water Inlet Pipeline (PW-IPL) is of stainless steel conveying demineralized water from hold-up tank to the reactor pool of Pakistan Research Reactor-1 (PARR-1). The section of the pipeline from heat exchangers to the valve pit is hanger supported in the pump room and the rest of the section from valve pit to the reactor pool is embedded. The PW-IPL is subjected to steady state and transient vibrations. The reactor pumps, which drive the coolant through various circuits mainly contribute the steady state vibrations, while transient vibrations arise due to instant closure of the check valve (water hammer). The ASME Boiler and Pressure Vessel code provides data about the acceptable limits of stresses related to the primary static stress due to steady state vibrations. However, due to complexity in the pipe structure, stresses related to the transient vibrations are neglected in the code. In this report attempt has been made to analyzed both steady state and transient vibrations of PW-IPL of PARR-1. Since, both the steady state and transient vibrations affect the hanger-supported section of the PW-IPL, therefore, it was selected for vibration test measurements. In the analysis vibration data was compared with the allowable limits and estimations of maximum pressure build-up, eflection, natural frequency, tensile and shear load on hanger support, and the ratio of maximum combine stress to the allowable load were made. (author)

  18. An investigation of decreasing reactor coolant inventory as a mechanism to reduce power during a boiling water reactor anticipated transient without scram

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Gose, G.C.; Hentzen, R.D.; Layman, W.H.

    1985-01-01

    Under certain anticipated transient without scram (ATWS) sequences for a boiling water reactor, it would be desirable to reduce system power, particularly where the primary system has been isolated by closure of all main steam isolation valves and is discharging steam through its safety/relief valve system to the suppression pool. Reducing reactor power increases the time available to shut down the reactor by minimizing the heat dumped to the suppression pool and by helping to keep the suppression pool temperature within limits. Under proposed emergency procedure guidelines for the ATWS event, the reactor water level would be lowered to reduce reactor power. The analyses provide an assessment of the power level that would be attained, assuming the reactor operators were to reduce the the downcomer level down to the top of the active fuel

  19. Multimodal Person Re-identification Using RGB-D Sensors and a Transient Identification Database

    DEFF Research Database (Denmark)

    Møgelmose, Andreas; Moeslund, Thomas B.; Nasrollahi, Kamal

    2013-01-01

    This paper describes a system for person re-identification using RGB-D sensors. The system covers the full flow, from detection of subjects, over contour extraction, to re-identification using soft biometrics. The biometrics in question are part-based color histograms and the subjects height...

  20. Failed fuel identification techniques for liquid-metal cooled reactors

    International Nuclear Information System (INIS)

    Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

    1995-01-01

    The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures

  1. The effect of boron dilution transient on the VVER-1000 reactor core using MCNP and COBRA-EN codes

    Energy Technology Data Exchange (ETDEWEB)

    Jafari, Naser; Talebi, Saeed [Amirkabir Univ. of Technology, Tehran Polytechnic (Iran, Islamic Republic of). Dept. of Energy Engineering and Physics

    2017-07-15

    In this paper, the effect of boron dilution transient, as a consequence of the malfunction of the boron control system, was investigated in a VVER-1000 reactor, and then an appropriate setpoint was determined for the actuation of the emergency protection system to the reactor shutdown. In order to simulate the boron dilution, first, the whole reactor core was simulated by MCNPX code to compute the radial and axial power distribution. Then, the COBRA-EN code was employed using calculated power distribution for analyzing the thermal-hydraulic of hot fuel assembly and for extracting the safety parameters. For the safe operation of the reactor, certain parameters must be in defined specified ranges. Comparison between our results and FSARs data shows that the present modeling provides a good prediction of boron dilution transient with the maximum relative difference about 4%.

  2. Analysis of a main steam isolation valve closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main stream isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4

  3. Advanced Neutron Source Reactor (ANSR) phenomena identification and ranking (PIR) for large break loss of coolant accidents (LBLOCA)

    International Nuclear Information System (INIS)

    Ruggles, A.E.; Cheng, L.Y.; Dimenna, R.A.; Griffith, P.; Wilson, G.E.

    1994-06-01

    A team of experts in reactor analysis conducted a phenomena identification and ranking (PIR) exercise for a large break loss-of-coolant accident (LBLOCA) in the Advanced Neutron source Reactor (ANSR). The LBLOCA transient is broken into two separate parts for the PIR exercise. The first part considers the initial depressurization of the system that follows the opening of the break. The second part of the transient includes long-term decay heat removal after the reactor is shut down and the system is depressurized. A PIR is developed for each part of the LBLOCA. The ranking results are reviewed to establish if models in the RELAP5-MOD3 thermalhydraulic code are adequate for use in ANSR LBLOCA simulations. Deficiencies in the RELAP5-MOD3 code are identified and existing data or models are recommended to improve the code for this application. Experiments were also suggested to establish models for situations judged to be beyond current knowledge. The applicability of the ANSR PIR results is reviewed for the entire set of transients important to the ANSR safety analysis

  4. Development of Transient-Reactor Analysis Code (TRAC) for real-time applications

    International Nuclear Information System (INIS)

    Niederauer, G.F.; Giguere, P.T.; Lime, J.F.; Knight, T.D.; Ashy, O.; Fakory, R.

    1997-01-01

    This is the final report of a six-month, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). Nuclear-plant training simulators employ simplified one-dimensional thermal-hydraulics codes because of the demands to run in real time and with limited computing power. The objective of this project was to investigate the feasibility of using the advanced Transient-Reactor Analysis Code (TRAC) in a simulator to increase the fidelity of a simulator. Many issues need to be addressed to take such a complex code from a batch engineering environment to a real-time environment. Working with simulator vendor, GSE, the authors investigated the technical issues relating to integrating TRAC into a real-time environment. They also modified a nuclear power plant model for simulator purposes and investigated its performance in real time

  5. Hybrid parallel strategy for the simulation of fast transient accidental situations at reactor scale

    International Nuclear Information System (INIS)

    Faucher, V.; Galon, P.; Beccantini, A.; Crouzet, F.; Debaud, F.; Gautier, T.

    2013-01-01

    This contribution is dedicated to the latest methodological developments implemented in the fast transient dynamics software EUROPLEXUS (EPX) to simulate the mechanical response of fully coupled fluid-structure systems to accidental situations to be considered at reactor scale, among which the Loss of Coolant Accident, the Core Disruptive Accident and the Hydrogen Explosion. Time integration is explicit and the search for reference solutions within the safety framework prevents any simplification and approximations in the coupled algorithm: for instance, all kinematic constraints are dealt with using Lagrange Multipliers, yielding a complex flow chart when non-permanent constraints such as unilateral contact or immersed fluid-structure boundaries are considered. The parallel acceleration of the solution process is then achieved through a hybrid approach, based on a weighted domain decomposition for distributed memory computing and the use of the KAAPI library for self-balanced shared memory processing inside sub-domains. (authors)

  6. Nodal methods for calculating nuclear reactor transients, control rod patterns, and fuel pin powers

    International Nuclear Information System (INIS)

    Cho, Byungoh.

    1990-01-01

    Nodal methods which are used to calculate reactor transients, control rod patterns, and fuel pin powers are investigated. The 3-D nodal code, STORM, has been modified to perform these calculations. Several numerical examples lead to the following conclusions: (1) By employing a thermal leakage-to-absorption ratio (TLAR) approximation for the spatial shape of the thermal fluxes for the 3-D Langenbuch-Maurer-Werner (LMW) and the superprompt critical transient problems, the convergence of the conventional two-group scheme is accelerated. (2) By employing the steepest-ascent hill climbing search with heuristic strategies, Optimum Control Rod Pattern Searcher (OCRPS) is developed for solving control rod positioning problem in BWRs. Using the method of approximation programming the objective function and the nuclear and thermal-hydraulic constraints are modified as heuristic functions that guide the search. The test calculations have demonstrated that, for the first cycle of the Edwin Hatch Unit number-sign 2 reactor, OCRPS shows excellent performance for finding a series of optimum control rod patterns for six burnup steps during the operating cycle. (3) For the modified two-dimensional EPRI-9R problem, the least square second-order polynomial flux expansion method was demonstrated to be computationally about 30 times faster than a fine-mesh finite difference calculation in order to achieve comparable accuracy for pin powers. The basic assumption of this method is that the reconstructed flux can be expressed as a product of an assembly form function and a second-order polynomial function

  7. Transient Analysis of a Gas-cooled Fast Reactor for Single Control Assembly Withdrawal

    International Nuclear Information System (INIS)

    Choi, Hangbok

    2014-01-01

    The Energy Multiplier Module (EMZ) system response has been evaluated for control assembly withdrawal transients. Currently the EM2 core is equipped with six cylindrical drum-type control assemblies in the reflector zone for excess reactivity control and power maneuvering during the operating core life. This study investigates the system response to the control assembly withdrawal accident with various rotational speeds and reactivity worth to determine feasible control assembly design requirements from the physics viewpoint. The simulations have been conducted for single control assembly withdrawal transients without scram by a gas-cooled reactor plant simulator, which is based on a simplified plant nodal model, including the point reactor kinetics, single channel core thermal-fluid model, and a turbo-machinery performance model. Simulations were conducted for the middle-of- cycle core, when the excess reactivity of the core is the highest. Control assembly withdrawal times were varied from 1 (runaway) to 180 sec and reactivity worth was varied from 100 to 400 pcm. For a single control assembly withdrawal, the simulation has shown that the peak fuel temperature is expected to be ~1820°C when the assembly worth is 200 pcm and the runaway time is 1 sec per 180 degree rotation. The peak temperature could be reduced to ~1780°C if the assembly is rotated out in a moderate speed such as 1 degree/sec. These peak temperatures give a thermal margin of 22 to 24% to the melting point of uranium carbide fuel. The results also indicate that the current design with a single control assembly worth of 314 pcm may need adjustments in the future design. (author)

  8. CFD investigation of flow inversion in typical MTR research reactor undergoing thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Salama, Amgad

    2011-01-01

    Highlights: → The 3D, CFD simulation of FLOFA accident in the generic IAEA 10 MW research reactor is carried out. → The different flow and heat transfer mechanisms involved in this process were elucidated. → The transition between these mechanisms during the course of FLOFA is discussed and investigated. → The interesting inversion process upon the transition from downward flow to upward flow is shown. → The temperature field and the friction coefficient during the whole transient process were shown. - Abstract: Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.

  9. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

  10. Diagnosis of class using swarm intelligence applied to problem of identification of nuclear transient

    Energy Technology Data Exchange (ETDEWEB)

    Villas Boas Junior, Manoel; Strauss, Edilberto, E-mail: junior@lmp.ufrj.b [Instituto Federal de Educacao, Ciencia e Tecnologia do Ceara/ Universidade do Estado do Ceara, Itaperi, CE (Brazil). Mestrado Integrado em Computacao Aplicada; Nicolau, Andressa dos Santos; Schirru, Roberto, E-mail: andressa@lmp.ufrj.b [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Mello, Flavio Luis de [Universidade Federal do Rio de Janeiro (POLI/UFRJ), RJ (Brazil). Escola Politecnica. Dept. de Engenharia Eletronica e Computacao

    2011-07-01

    This article presents a computational model of the diagnostic system of transient. The model makes use of segmentation techniques applied to support decision making, based on identification of classes and optimized by Particle Swarm Optimization algorithm (PSO). The method proposed aims to classify an anomalous event in the signatures of three classes of the design basis transients postulated for the Angra 2 nuclear plant, where the PSO algorithm is used as a method of separation of classes, being responsible for finding the best centroid prototype vector of each accident/transient, ie equivalent to Voronoi vector that maximizes the number of correct classifications. To make the calculation of similarity between the set of the variables anomalous event in a given time t, and the prototype vector of variables of accident/transients, the metrics of Manhattan, Euclidean and Minkowski were used. The results obtained by the method proposed were compatible with others methods reported in the literature, allowing a solution that approximates the ideal solution, ie the Voronoi vectors. (author)

  11. Diagnosis of class using swarm intelligence applied to problem of identification of nuclear transient

    International Nuclear Information System (INIS)

    Villas Boas Junior, Manoel; Strauss, Edilberto; Nicolau, Andressa dos Santos; Schirru, Roberto; Mello, Flavio Luis de

    2011-01-01

    This article presents a computational model of the diagnostic system of transient. The model makes use of segmentation techniques applied to support decision making, based on identification of classes and optimized by Particle Swarm Optimization algorithm (PSO). The method proposed aims to classify an anomalous event in the signatures of three classes of the design basis transients postulated for the Angra 2 nuclear plant, where the PSO algorithm is used as a method of separation of classes, being responsible for finding the best centroid prototype vector of each accident/transient, ie equivalent to Voronoi vector that maximizes the number of correct classifications. To make the calculation of similarity between the set of the variables anomalous event in a given time t, and the prototype vector of variables of accident/transients, the metrics of Manhattan, Euclidean and Minkowski were used. The results obtained by the method proposed were compatible with others methods reported in the literature, allowing a solution that approximates the ideal solution, ie the Voronoi vectors. (author)

  12. Anticipated transient without scram analysis of the simplified boiling water reactor following main steam isolation valve closure with boron injection

    International Nuclear Information System (INIS)

    Khan, H.J.; Cheng, H.S.; Rohatgi, U.S.

    1996-01-01

    The simplified boiling water reactor (SBWR) operating in natural circulation is designed with many passive safety features. An anticipated transient without scram (ATWS) initiated by inadvertent closure of the main steam isolation valve (MSIV) in an SBWR has been analyzed using the RAMONA-4B code of Brookhaven National Laboratory. This analysis demonstrates the predicted performance of the SBWR during an MSIV closure ATWS, followed by shutdown of the reactor through injection of boron into the reactor core from the standby liquid control system

  13. Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

    International Nuclear Information System (INIS)

    Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

    2003-01-01

    The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

  14. Abstracts of papers from the literature on anticipated transients without scram for light water reactors 1. 1975-1979

    International Nuclear Information System (INIS)

    Kinnersley, S.R.

    1981-05-01

    INIS ATOMINDEX abstracts relating to ATWS for light water reactors for the years 1975-1979 are presented under the subject headings of; general, licensing and standards, models and computer codes, frequency of occurrence of ATWS, transient calculations of results including probabilistic analysis, radiological consequences of ATWS, fuel behaviour, and studies of plant components. (U.K.)

  15. TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

    International Nuclear Information System (INIS)

    Chalhoub, E.S.

    1980-09-01

    A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

  16. Pressure transients analysis of a high-temperature gas-cooled reactor with direct helium turbine cycle

    Energy Technology Data Exchange (ETDEWEB)

    Dang, M.; Dupont, J. F.; Jacquemoud, P.; Mylonas, R. [Eidgenoessisches Inst. fuer Reaktorforschung, Wuerenlingen (Switzerland)

    1981-01-15

    The direct coupling of a gas cooled reactor with a closed gas turbine cycle leads to a specific dynamic plant behaviour, which may be summarized as follows: a) any operational transient involving a variation of the core mass flow rate causes a variation of the pressure ratio of the turbomachines and leads unavoidably to pressure and temperature transients in the gas turbine cycle; and b) very severe pressure equalization transients initiated by unlikely events such as the deblading of one or more turbomachines must be taken into account. This behaviour is described and illustrated through results gained from computer analyses performed at the Swiss Federal Institute for Reactor Research (EIR) in Wurenlingen within the scope of the Swiss-German HHT project.

  17. Environmental impacts of radiological consequences during the anticipated transients without scram (ATWS) events in nuclear power reactors

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    2011-01-01

    Anticipated transients without scram (ATWS), is one of the (worst case) accidents could happen if the system that provides a highly reliable means of shutting down the reactor (scram system )fails to work during a reactor event (anticipated transient).It has two general characteristics: (1) Initiation by a transient anticipated to occur one or more times in the life of reactor and ,(2) Assumed to proceed without scram.The types of events considered are those used for designing the plant .The evaluation of the radiological consequences during the assessment of the nuclear events,especially ATWS in nuclear power reactors, is very essential for environmental studies and public safety. In this paper, the root cases for nuclear events and dose calculation are presented. Scenario of accident sequences together with radiological impacts is illustrated for loss of coolant accident (LOCA) for a typical pressurized water reactor nuclear power plant. Recommendations for mitigating or preventing the release of radiation and high radioactive materials to environment are presented.

  18. Transient heat transfer in a directly-irradiated solar chemical reactor for the thermal dissociation of ZnO

    International Nuclear Information System (INIS)

    Mueller, R.; Lipinski, W.; Steinfeld, A.

    2008-01-01

    A numerical and experimental investigation is carried out in a solar thermochemical reactor for the thermal dissociation of ZnO at 2000 K using concentrated solar energy. The reactor consists of a cavity-receiver lined with ZnO particles and directly exposed to high-flux irradiation. A transient heat transfer model is formulated to link the rate of radiation, convection, and conduction heat transfer to the reaction kinetics. The radiosity and Monte Carlo methods are applied to obtain the distribution of net radiative fluxes at the internal surfaces of the reactor cavity and at the surface of the ZnO bed. Validation is accomplished in terms of the calculated and measured transient temperature profiles and chemical reaction rates

  19. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  20. Transient recovery voltage analysis for various current breaking mathematical models: shunt reactor and capacitor bank de-energization study

    Directory of Open Access Journals (Sweden)

    Oramus Piotr

    2015-09-01

    Full Text Available Electric arc is a complex phenomenon occurring during the current interruption process in the power system. Therefore performing digital simulations is often necessary to analyse transient conditions in power system during switching operations. This paper deals with the electric arc modelling and its implementation in simulation software for transient analyses during switching conditions in power system. Cassie, Cassie-Mayr as well as Schwarz-Avdonin equations describing the behaviour of the electric arc during the current interruption process have been implemented in EMTP-ATP simulation software and presented in this paper. The models developed have been used for transient simulations to analyse impact of the particular model and its parameters on Transient Recovery Voltage in different switching scenarios: during shunt reactor switching-off as well as during capacitor bank current switching-off. The selected simulation cases represent typical practical scenarios for inductive and capacitive currents breaking, respectively.

  1. Neural-net based unstable machine identification using individual energy functions. [Transient disturbances in power systems

    Energy Technology Data Exchange (ETDEWEB)

    Djukanovic, M [Institut Nikola Tesla, Belgrade (Yugoslavia); Sobajic, D J; Pao, Yohhan [Case Western Reserve Univ., Cleveland, OH (United States)

    1991-10-01

    The identification of the mode of instability plays an essential role in generating principal energy boundary hypersurfaces. We present a new method for unstable machine identification based on the use of supervised learning neural-net technology, and the adaptive pattern recognition concept. It is shown that using individual energy functions as pattern features, appropriately trained neural-nets can retrieve the reliable characterization of the transient process including critical clearing time parameter, mode of instability and energy margins. Generalization capabilities of the neural-net processing allow for these assessments to be made independently of load levels. The results obtained from computer simulations are presented using the New England power system, as an example. (author).

  2. Analysis of a main steam isolation value closure anticipated transient without scram in a boiling water reactor

    International Nuclear Information System (INIS)

    Liaw, T.J.; Pan, C.; Chen, G.S.

    1989-01-01

    Anticipated transient without scram (ATWS) could be a major accident sequence with possible core melt and containment damage in a boiling water reactor (BWR). The behavior of a BWR/6 during a main steam isolation valve closure ATWS is investigated using the best-estimate computer program, RETRAN-02. The effects of both makeup coolant and boron injection on the reactor behavior are studied. It is found that the BWR/6 behaves similarly to the BWR/2 and BWR/4. Without boron injection and makeup coolant, the reactor loses its coolant inventory very quickly and the reactor power drops rapidly to ∼ 16% of rated power due to negative void reactivity. With coolant makeup from the high-pressure core spray and the reactor core isolation cooling systems, the rector reaches a quasi-steady-state condition after an initially rapidly changing transient. The dome pressure, downcomer water level, and core power oscillate around a mean value; the average core power is ∼ 15%, which is approximately equal to the power needed to heat and evaporate the subcooled makeup coolant. Lower boron concentrations in the core tend to complicate reactor behavior due to the combination of two competing phenomena: the negative boron reactivity and the positive reactivity caused by a void collapse

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  4. The improved quasi-static method vs the direct method: a case study for CANDU reactor transients

    International Nuclear Information System (INIS)

    Kaveh, S.; Koclas, J.; Roy, R.

    1999-01-01

    Among the large number of methods for the transient analysis of nuclear reactors, the improved quasi-static procedure is one of the most widely used. In recent years, substantial increase in both computer speed and memory has motivated a rethinking of the limitations of this method. The overall goal of the present work is a systematic comparison between the improved quasi-static and the direct method (mesh-centered finite difference) for realistic CANDU transient simulations. The emphasis is on the accuracy of the solutions as opposed to the computational speed. Using the computer code NDF, a typical realistic transient of CANDU reactor has been analyzed. In this transient the response of the reactor regulating system to a substantial local perturbation (sudden extraction of the five adjuster rods) has been simulated. It is shown that when updating the detector responses is of major importance, it is better to use a well-optimized direct method rather than the improved quasi-static method. (author)

  5. Time-varying singular value decomposition for periodic transient identification in bearing fault diagnosis

    Science.gov (United States)

    Zhang, Shangbin; Lu, Siliang; He, Qingbo; Kong, Fanrang

    2016-09-01

    For rotating machines, the defective faults of bearings generally are represented as periodic transient impulses in acquired signals. The extraction of transient features from signals has been a key issue for fault diagnosis. However, the background noise reduces identification performance of periodic faults in practice. This paper proposes a time-varying singular value decomposition (TSVD) method to enhance the identification of periodic faults. The proposed method is inspired by the sliding window method. By applying singular value decomposition (SVD) to the signal under a sliding window, we can obtain a time-varying singular value matrix (TSVM). Each column in the TSVM is occupied by the singular values of the corresponding sliding window, and each row represents the intrinsic structure of the raw signal, namely time-singular-value-sequence (TSVS). Theoretical and experimental analyses show that the frequency of TSVS is exactly twice that of the corresponding intrinsic structure. Moreover, the signal-to-noise ratio (SNR) of TSVS is improved significantly in comparison with the raw signal. The proposed method takes advantages of the TSVS in noise suppression and feature extraction to enhance fault frequency for diagnosis. The effectiveness of the TSVD is verified by means of simulation studies and applications to diagnosis of bearing faults. Results indicate that the proposed method is superior to traditional methods for bearing fault diagnosis.

  6. Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors

    International Nuclear Information System (INIS)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2014-01-01

    The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

  7. Comparison of in-plant performance test data with analytic prediction of reactor safety system injection transient (U)

    International Nuclear Information System (INIS)

    Roy, B.N.; Neill, C.H. Jr.

    1993-01-01

    This paper compares the performance test data from injection transients for both of the subsystems of the Supplementary Safety System of the Savannah River Site production reactor with analytical predictions from an in-house thermal hydraulic computer code. The code was initially developed for design validation of the new Supplementary Safety System subsystem, but is shown to be equally capable of predicting the performance of the Supplementary Safety System existing subsystem even though the two subsystem transient injections have marked differences. The code itself was discussed and its validation using prototypic tests with simulated fluids was reported in an earlier paper (Roy and Nomm 1991)

  8. The effect of code user and boundary conditions on RELAP calculations of MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The safety evaluation of nuclear power and re search reactors is a very important step before their construction and during their operation. This evaluation based on the best estimate calculations requires qualified codes qualified users, and qualified nodalizations. The effect of code users on the RELAP5 results during the analysis of loss of flow transient in MTR research reactors is presented in this pa per. To clarify this effect, two nodalizations for research reactor different in the simulation of the open water surface boundary conditions of the reactor pool have been used. Very different results are obtained with few choices for code users. The core natural circulation flow with the be ginning of core boiling doesn't stop but in creases. The in creasing in the natural circulation flow shifts out the boiling from the core and the clad temperature decreases be low the local saturation temperature.

  9. Analyses and computer code developments for accident-induced thermohydraulic transients in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    Wulff, W.

    1977-01-01

    A review is presented on the development of analyses and computer codes for the prediction of thermohydraulic transients in nuclear reactor systems. Models for the dynamics of two-phase mixtures are summarized. Principles of process, reactor component and reactor system modeling are presented, as well as the verification of these models by comparing predicted results with experimental data. Codes of major importance are described, which have recently been developed or are presently under development. The characteristics of these codes are presented in terms of governing equations, solution techniques and code structure. Current efforts and problems of code verification are discussed. A summary is presented of advances which are necessary for reducing the conservatism currently implied in reactor hydraulics codes for safety assessment

  10. Transient Temperature Distribution in a Reactor Core with Cylindrical Fuel Rods and Compressible Coolant

    Energy Technology Data Exchange (ETDEWEB)

    Vollmer, H

    1968-04-15

    Applying linearization and Laplace transformation the transient temperature distribution and weighted temperatures in fuel, canning and coolant are calculated analytically in two-dimensional cylindrical geometry for constant material properties in fuel and canning. The model to be presented includes previous models as special cases and has the following novel features: compressibility of the coolant is accounted for. The material properties of the coolant are variable. All quantities determining the temperature field are taken into account. It is shown that the solution for fuel and canning temperature may be given by the aid of 4 basic transfer functions depending on only two variables. These functions are calculated for all relevant rod geometries and material constants. The integrals involved in transfer functions determining coolant temperatures are solved for the most part generally by application of coordinate and Laplace transformation. The model was originally developed for use in steam cooled fast reactor analysis where the coolant temperature rise and compressibility are considerable. It may be applied to other fast or thermal systems after suitable simplifications.

  11. Accidents and transients analyses of a super fast reactor with single flow pass core

    International Nuclear Information System (INIS)

    Sutanto,; Oka, Yoshiaki

    2014-01-01

    Highlights: • Safety analysis of a Super FR with single flow pass core is conducted. • Loss of feed water flow leads to a direct effect on the loss of fuel channel flow. • The core pressure is sensitive to LOCA accidents due to the direct effect. • Small LOCA introduces a critical break. • The safety criteria for all selected events are satisfied. - Abstract: The supercritical water cooled fast reactor with single flow pass core has been designed to simplify refueling and the structures of upper and lower mixing plenums. To evaluate the safety performance, safety analysis has been conducted with regard to LOCA and non-LOCA accidents including transient events. Safety analysis results show that the safety criteria are satisfied for all selected events. The total loss of feed water flow is the most important accident which the maximum cladding surface temperature (MCST) is high due to a direct effect of the accident on the total loss of flow in all fuel assemblies. However, actuation of the ADS can mitigate the accident. Small LOCA also introduces a critical break at 7.8% break which results high MCST at BOC because the scram and ADS are not actuated. Early ADS actuation is effective to mitigate the accident. In large LOCA, 100% break LOCA results a high MCST of flooding phase at BOC due to high power peaking at the bottom part. Use of high injection flow rate by 2 LPCI units is effective to decrease the MCST

  12. Application programming interface document for the modernized Transient Reactor Analysis Code (TRAC-M)

    International Nuclear Information System (INIS)

    Mahaffy, J.; Boyack, B.E.; Steinke, R.G.

    1998-05-01

    The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization

  13. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  14. High efficiency algorithm for 3D transient thermo-elasto-plastic contact problem in reactor pressure vessel sealing system

    International Nuclear Information System (INIS)

    Xu Mingyu; Lin Tengjiao; Li Runfang; Du Xuesong; Li Shuian; Yang Yu

    2005-01-01

    There are some complex operating cases such as high temperature and high pressure during the operating process of nuclear reactor pressure vessel. It is necessary to carry out mechanical analysis and experimental investigation for its sealing ability. On the basis of the self-developed program for 3-D transient sealing analysis for nuclear reactor pressure vessel, some specific measures are presented to enhance the calculation efficiency in several aspects such as the non-linear solution of elasto-plastic problem, the mixed solution algorithm for contact problem as well as contract heat transfer problem and linear equation set solver. The 3-D transient sealing analysis program is amended and complemented, with which the sealing analysis result of the pressure vessel model can be obtained. The calculation results have good regularity and the calculation efficiency is twice more than before. (authors)

  15. 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

  16. 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

  17. A highly efficient autothermal microchannel reactor for ammonia decomposition: Analysis of hydrogen production in transient and steady-state regimes

    Science.gov (United States)

    Engelbrecht, Nicolaas; Chiuta, Steven; Bessarabov, Dmitri G.

    2018-05-01

    The experimental evaluation of an autothermal microchannel reactor for H2 production from NH3 decomposition is described. The reactor design incorporates an autothermal approach, with added NH3 oxidation, for coupled heat supply to the endothermic decomposition reaction. An alternating catalytic plate arrangement is used to accomplish this thermal coupling in a cocurrent flow strategy. Detailed analysis of the transient operating regime associated with reactor start-up and steady-state results is presented. The effects of operating parameters on reactor performance are investigated, specifically, the NH3 decomposition flow rate, NH3 oxidation flow rate, and fuel-oxygen equivalence ratio. Overall, the reactor exhibits rapid response time during start-up; within 60 min, H2 production is approximately 95% of steady-state values. The recommended operating point for steady-state H2 production corresponds to an NH3 decomposition flow rate of 6 NL min-1, NH3 oxidation flow rate of 4 NL min-1, and fuel-oxygen equivalence ratio of 1.4. Under these flows, NH3 conversion of 99.8% and H2 equivalent fuel cell power output of 0.71 kWe is achieved. The reactor shows good heat utilization with a thermal efficiency of 75.9%. An efficient autothermal reactor design is therefore demonstrated, which may be upscaled to a multi-kW H2 production system for commercial implementation.

  18. Finite strain transient creep of D16T alloy: identification and validation employing heterogeneous tests

    Science.gov (United States)

    Shutov, A. V.; Larichkin, A. Yu

    2017-10-01

    A cyclic creep damage model, previously proposed by the authors, is modified for a better description of the transient creep of D16T alloy observed in the finite strain range under rapidly changing stresses. The new model encompasses the concept of kinematic hardening, which allows us to account for the creep-induced anisotropy. The model kinematics is based on the nested multiplicative split of the deformation gradient, proposed by Lion. The damage evolution is accounted for by the classical Kachanov-Rabotnov approach. The material parameters are identified using experimental data on cyclic torsion of thick-walled samples with different holding times between load reversals. For the validation of the proposed material model, an additional experiment is analyzed. Although this additional test is not involved in the identification procedure, the proposed cyclic creep damage model describes it accurately.

  19. Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

    International Nuclear Information System (INIS)

    Shin, Hyeong-Ki

    1999-01-01

    The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

  20. Diagnostic techniques for measuring temperature transients and stress transients in the first wall of an ICF reactor

    International Nuclear Information System (INIS)

    Melamed, N.T.; Taylor, L.H.

    1983-01-01

    The primary challenge in the design of an Inertial Confinement Fusion (ICF) power reactor is to make the first wall survive the frequent explosions of the pellets. Westinghouse has proposed a dry wall design consisting of steel tubes coated with tantalum. This report describes the design of a test chamber and two diagnostic procedures for experimentally determining the reliability of the Westinghouse design. The test chamber simulates the x-ray and ion pulse irradiation of the wall due to a pellet explosion. The diagnostics consist of remote temperature sensing and surface deformation measurements. The chamber and diagnostics can also be used to test other first-wall designs

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

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Sunagawa, Takeyoshi

    2014-01-01

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

  2. A three-dimensional transient calculation for the reactor model RAMONA using the COMMIX-2(V) code

    International Nuclear Information System (INIS)

    Weinberg, D.; Frey, H.H.; Tschoeke, H.

    1993-01-01

    The safety graded decay heat removal system of the European Fast Reactor needs a high availability. This system operates entirely under natural convection. To guarantee a proper design, experiments are carried out to verify thermal-hydraulic computer codes able to predict precisely local temperature loadings of the components and the reactor tank in the transition region from nominal operation under forced convection to the decay heat removal operation. - With the COMMIX-2 (V) code three-dimensional transient calculations have been performed to simulate experiments in the 360 deg. reactor model RAMONA, scaled 1:20 to the reality with water as simulant fluid for sodium. The computed average and local temperatures as well as the velocity distributions show a good agreement with the experimental results. Further efforts are necessary to reduce the computation time. (orig.)

  3. Probabilistic fracture mechanics analysis of boiling water reactor vessel for cool-down and low temperature over-pressurization transients

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jeong Soon; Choi, Young Hwan; Jhung, Myung Jo [Safety Research Division, Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-04-15

    The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  4. Implementation into a CFD code of neutron kinetics and fuel pin models for nuclear reactor transient analyses

    International Nuclear Information System (INIS)

    Chen Zhao; Chen, Xue-Nong; Rineiski, Andrei; Zhao Pengcheng; Chen Hongli

    2014-01-01

    Safety analysis is an important tool for justifying the safety of nuclear reactors. The traditional method for nuclear reactor safety analysis is performed by means of system codes, which use one-dimensional lumped-parameter method to model real reactor systems. However, there are many multi-dimensional thermal-hydraulic phenomena cannot be predicated using traditional one-dimensional system codes. This problem is extremely important for pool-type nuclear systems. Computational fluid dynamics (CFD) codes are powerful numerical simulation tools to solve multi-dimensional thermal-hydraulics problems, which are widely used in industrial applications for single phase flows. In order to use general CFD codes to solve nuclear reactor transient problems, some additional models beyond general ones are required. Neutron kinetics model for power calculation and fuel pin model for fuel pin temperature calculation are two important models of these additional models. The motivation of this work is to develop an advance numerical simulation method for nuclear reactor safety analysis by implementing neutron kinetics model and fuel pin model into general CFD codes. In this paper, the Point Kinetics Model (PKM) and Fuel Pin Model (FPM) are implemented into a general CFD code FLUENT. The improved FLUENT was called as FLUENT/PK. The mathematical models and implementary method of FLUENT/PK are descripted and two demonstration application cases, e.g. the unprotected transient overpower (UTOP) accident of a Liquid Metal cooled Fast Reactor (LMFR) and the unprotected beam overpower (UBOP) accident of an Accelerator Driven System (ADS), are presented. (author)

  5. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report

    International Nuclear Information System (INIS)

    Rohde, Ulrich; Pivovarov, Valeri; Matveev, Yurij

    2010-12-01

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  6. Numerical scheme for optimization of xenon transient processes in a reactor. Problem on fast response without a limitation for phase variables

    International Nuclear Information System (INIS)

    Gerasimov, A.S.

    1975-01-01

    A numerical diagram is suggested of minimizing a period of xenon transient process in the reactor without any limitation of xenon-135 concentration. The problem is solved with a computer in a point model. Pontryagin's maximum principle is used so as to check optimization of the transient process

  7. Analysis on blow-down transient in water ingress accident of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang, Yan; Zheng, Yanhua; Li, Fu; Shi, Lei

    2014-01-01

    Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor, which will cause a positive reactivity introduction with the increase of steam density in reactor core to enhance neutron slowing-down, also the chemical corrosion of graphite fuel elements and the damage of reflector structure material. The increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The research on water ingress transient is significant for the verification of inherent safety characteristics of high temperature gas-cooled reactor. The 200 MWe high temperature gas-cooled reactor (HTR-PM), designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is exampled to be analyzed in this paper. The design basis accident (DBA) scenarios of double-ended guillotine break of single heat-exchange tube (steam generator heat-exchange tube rupture) are simulated by the thermal-hydraulic analysis code, and some key concerns which are relative to the amount of water into the reactor core during the blow-down transient are analyzed in detail. The results show that both of water mass and steam ratio of the fluid spouting from the broken heat-exchange tube are affected by break location, which will increase obviously with the broken location closing to the outlet of the heat-exchange tube. The double-ended guillotine rupture at the outlet of the heat-exchange will result more steam penetrates into the reactor core in the design basis accident of water ingress. The mass of water ingress will also be affected by the draining system. It is concluded that, with reasonable optimization on design to balance safety and economy, the total mass of water ingress into the primary circuit of reactor could be limited effectively to meet the safety requirements, and the pollution of

  8. Transient temperature and stress distributions in the pressure vessel's wall of a nuclear reactor

    International Nuclear Information System (INIS)

    Silva, G.A. da

    1979-01-01

    In order to calculate the temperature distribution in a reactor vessel wall which is under the effect of gamma radiation originated in the reactor core, a numerical solution is proposed. This problem may arise from a reactor cooling pump failure .The thermal stresses are also calculated. (Author) [pt

  9. Assessments of the kinetic and dynamic transient behavior of sub-critical systems (ADS) in comparison to critical reactor systems

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    2001-01-01

    The neutron kinetic and the reactor dynamic behavior of Accelerator Driven Systems (ADS) is significantly different from those of conventional power reactor systems currently in use for the production of power. It is the objective of this study to examine and to demonstrate the intrinsic differences of the kinetic and dynamic behavior of accelerator driven systems to typical plant transient initiators in comparison to the known, kinetic and dynamic behavior of critical thermal and fast reactor systems. It will be shown that in sub-critical assemblies, changes in reactivity or in the external neutron source strength lead to an asymptotic power level essentially described by the instantaneous power change (i.e. prompt jump). Shutdown of ADS operating at high levels of sub-criticality, (i.e. k eff ∼0.99), without the support of reactivity control systems (such as control or safety rods), may be problematic in case the ability of cooling of the core should be impaired (i.e. loss of coolant flow). In addition, the dynamic behavior of sub-critical systems to typical plant transients such as protected or unprotected loss of flow (LOF) or heat sink (LOH) transients are not necessarily substantially different from the plant dynamic behavior of critical systems if the reactivity feedback coefficients of the ADS design are unfavorable. As expected, the state of sub-criticality and the temperature feedback coefficients, such as Doppler and coolant temperature coefficient, play dominant roles in determining the course and direction of plant transients. Should the combination of these safety coefficients be very unfavorable, not much additional margin in safety may be gained by making a critical system only sub-critical (i.e. k eff ∼0.95). A careful optimization procedure between the selected operating level of sub-criticality, the safety reactivity coefficients and the possible need for additional reactivity control systems seems, therefore, advisable during the early

  10. CATALisp - an expert system for real-time diagnosis of multiple-failure transients in pressurized water reactors

    International Nuclear Information System (INIS)

    Martin, R.P.; Nassersharif, B.

    1988-01-01

    The state of the art in artificial intelligence (AI) and expert system (ES) technology has matured to a degree that the potential development of a computer-aided/automated diagnostic and transient mitigation system in the area of nuclear reactor operation can be considered. Since traditional methods cannot handle complex systems efficiently, AI techniques provide a means to emulate an expert reactor operator rather than follow mechanistic methods. Computer-aided transient analysis coded in LISP (CATA-Lisp) is a confidence level based expert system written in Common LISP on the SYMBOLICS 3640 computer system. New versions are being developed in Common LISP for the Texas Instruments (TI) Explorer and the Sun microsystems machines, CATALisp manipulates both a knowledge base of transient identifier patterns (tree structured to allow for zooming in diagnostics) and a knowledge base containing a qualitative model of a nuclear power plant. The interference engine used by CATALisp uses the information stored in both knowledge bases to arrive at confidence level values that are used to infer particular plant states

  11. 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

  12. Validation of SCALE4.4a for Calculation of Xe-Sm Transients After a Scram of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2007-01-01

    The aim of this report is to validate the computational modules system SCALE4.4a for evaluation of reactivity changes, macroscopic absorption cross sections and calculations of the positions of the Control Rods during their motion in Xe-Sm transient after a scram of the BR-2 reactor. The rapid shutting down of the reactor by inserting of negative reactivity by the Control Rods is known as a reactor scram. Following reactor scram, a large xenon and samarium buildup occur in the reactor, which may appreciably affect the multiplication factor of the core due to enormous neutron absorption. The validation of the calculations of Xe-Sm transients by SCALE4.4a has been performed on the measurements of the positions of the Control Rods during their motion in Xe-Sm transients of the BR-2 reactor and on comparison with the calculations by the standard procedure XESM, developed at the BR-2 reactor. A final conclusion is made that the SCALE4.4a modules system can be used for evaluation of Xe-Sm transients of the BR-2 reactor. The utilization of the code is simple, the computational time takes from few seconds.

  13. Fatigue status assessment for reactor pressure vessel based on actual operational transient

    International Nuclear Information System (INIS)

    Zhu Guangqiang; Liao Changbin; Dai Bing; Gui Chun

    2013-01-01

    Background: Fatigue is an important aging mechanism in RPV and it must be contained to aging management working range. Purpose: In order to ensure the safety operation of nuclear power plants, as extension of RPV service time, it is necessary to assess the fatigue damage caused by actual operation transient. Methods: Based on monitoring data of actual operation during the past eleven years, refer to design transient, the statistic analysis for types and occurrence times of actual transient is carried out, at the same time, every transients are combined as different operation cycles and the temperature field and stress field of typical components are analyzed by FEM. Results: Based on these information, fatigue analysis and assessment are finished, if later-actual transients are similar with the previous transients, the calculation result shows that the ratio between maximum of cumulative usage factors and design calculation value is 0.4967 the design transients is conservative. Conclusions: Fatigue status of RPV could be assessed and traced quickly through fatigue status assessment method in this paper based on actual operational transient and assessment result would be a good reference for RPV aging management. (authors)

  14. Development of THYDE-HTGR: computer code for transient thermal-hydraulics of high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Hirano, Masashi; Hada, Kazuhiko

    1990-04-01

    The THYDE-HTGR code has been developed for transient thermal-hydraulic analyses of high-temperature gas-cooled reactors, based on the THYDE-W code. THYDE-W is a code developed at JAERI for the simulation of Light Water Reactor plant dynamics during various types of transients including loss-of-coolant accidents. THYDE-HTGR solves the conservation equations of mass, momentum and energy for compressible gas, or single-phase or two-phase flow. The major code modification from THYDE-W is to treat helium loops as well as water loops. In parallel to this, modification has been made for the neutron kinetics to be applicable to helium-cooled graphite-moderated reactors, for the heat transfer models to be applicable to various types of heat exchangers, and so forth. In order to assess the validity of the modifications, analyses of some of the experiments conducted at the High Temperature Test Loop of ERANS have been performed. In this report, the models applied in THYDE-HTGR are described focusing on the present modifications and the results from the assessment calculations are presented. (author)

  15. Transient freezing of molten salts in pipe-flow systems: Application to the direct reactor auxiliary cooling system (DRACS)

    International Nuclear Information System (INIS)

    Le Brun, N.; Hewitt, G.F.; Markides, C.N.

    2017-01-01

    Highlights: • A thermo-hydraulic model has been proposed to simulate the transient freezing of molten salts in complex piping systems. • The passive safety system DRACS in Generation-IV, molten salt reactor is susceptible to failure due to salt freezing. • For the prototypical 0.2 MW reactor considered in this study considerable freezing occurs after 20 minutes leading to reactor temperatures above 900 °C within 4 hours. • Conservative criteria for the most important/least known variables in the design of DRACS have been discussed. • Over-conservative approaches in designing the NDHX should be used with caution as they can promote pipe clogging due to freezing. - Abstract: The possibility of molten-salt freezing in pipe-flow systems is a key concern for the solar-energy industry and a safety issue in the new generation of molten-salt reactors, worthy of careful consideration. This paper tackles the problem of coolant solidification in complex pipe networks by developing a transient thermohydraulic model and applying it to the ‘Direct Reactor Auxiliary Cooling System’ (DRACS), the passive-safety system proposed for the Generation-IV molten-salt reactors. The results indicate that DRACS, as currently envisioned, is prone to failure due to freezing in the air/molten-salt heat exchanger, which can occur after approximately 20 minutes, leading to reactor temperatures above 900 °C within 4 hours. The occurrence of this scenario is related to an unstable behaviour mode of DRACS in which newly formed solid-salt deposit on the pipe walls acts to decrease the flow-rate in the secondary loop, facilitating additional solid-salt deposition. Conservative criteria are suggested to facilitate preliminary assessments of early-stage DRACS designs. The present study is, to the knowledge of the authors, the first of its kind in serving to illustrate possible safety concerns in molten-salt reactors, which are otherwise considered very safe in the literature. Furthermore

  16. Real-Time Gas Identification by Analyzing the Transient Response of Capillary-Attached Conductive Gas Sensor

    Directory of Open Access Journals (Sweden)

    Behzad Bahraminejad

    2010-05-01

    Full Text Available In this study, the ability of the Capillary-attached conductive gas sensor (CGS in real-time gas identification was investigated. The structure of the prototype fabricated CGS is presented. Portions were selected from the beginning of the CGS transient response including the first 11 samples to the first 100 samples. Different feature extraction and classification methods were applied on the selected portions. Validation of methods was evaluated to study the ability of an early portion of the CGS transient response in target gas (TG identification. Experimental results proved that applying extracted features from an early part of the CGS transient response along with a classifier can distinguish short-chain alcohols from each other perfectly. Decreasing time of exposition in the interaction between target gas and sensing element improved the reliability of the sensor. Classification rate was also improved and time of identification was decreased. Moreover, the results indicated the optimum interval of the early transient response of the CGS for selecting portions to achieve the best classification rates.

  17. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  18. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    International Nuclear Information System (INIS)

    Massoud, M.

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients

  19. Transient Effects in Fischer-Tropsch Reactor with a Fixed Bed of Catalyst Particles

    Directory of Open Access Journals (Sweden)

    I. V. Derevich

    2015-01-01

    Full Text Available Based on analysis of small temperature disturbances in the Fischer-Tropsch reactor with a fixed bed of catalyst particles various scenarios of thermal instability were investigated. There are two possible scenarios of thermal instability of the reactor. First, thermal explosion may occur due to growth of temperature disturbances inside a catalytic granule. Second scenario connected with loss of thermal stability as a result of an initial increase in temperature in the reactor volume. The boundaries of thermal stability of the reactor were estimated by solving the eigenvalue problems for spherical catalyst particles and cylindrical reactor. Processes of diffusional resistance inside the catalytic granule and heat transfer from wall of the reactor tube are taken into account. Estimation of thermal stability area is compared with the results of numerical simulation of behavior of temperature and concentration of synthesis gas.

  20. RETRAN operational transient analysis of the Big Rock Point plant boiling water reactor

    International Nuclear Information System (INIS)

    Sawtelle, G.R.; Atchison, J.D.; Farman, R.F.; VandeWalle, D.J.; Bazydlo, H.G.

    1983-01-01

    Energy Incorporated used the RETRAN computer code to model and calculate nine Consumers Power Company Big Rock Point Nuclear Power Plant transients. RETRAN, a best-estimate, one-dimensional, homogeneous-flow thermal-equilibrium code, is applicable to FSAR Chapter 15 transients for Conditions 1 through IV. The BWR analyses were performed in accordance with USNRC Standard Review Plan criteria and in response to the USNRC Systematic Evaluation Program. The RETRAN Big Rock Point model was verified by comparison to plant startup test data. This paper discusses the unique modeling techniques used in RETRAN to model this steam-drum-type BWR. Transient analyses results are also presented

  1. Preliminary evaluation of SACI-O code for the analysis of transients in a pressurized water reactor core

    International Nuclear Information System (INIS)

    Soares, P.A.; Sirimarco, L.F.; Veloso, M.A.F.

    1979-03-01

    SACI-O is a computer code which deals with the dynamics of the core of pressurized light water reactors (PWR). Its applicability is determined by the evaluation of the models used in the simulation of the several phenomena and processes which occur in the core during transients. This report presents a comparison between the results obtained with SACI-O and those presented in the Final Safety Analysis Report (FSAR) of Angra dos Reis Nuclear Station, Unit 1. Although some data used in the calculations done by Westinghouse are not known, there was a good agreement between the mentioned results. (Author) [pt

  2. Development of a preliminary PIRT (Phenomena Identification and Ranking Table) of thermal-hydraulic phenomena for 330MWt SMART integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, W. J.; Sim, S. K.; Song, J. H.; Kim, H. C.

    1997-09-01

    The work reported in this document identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in a 330 MWt SMART integral reactor which is under development at KAERI. The result of this efforts is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary Phenomena Identification and Ranking Table (PIRT) has been developed based on the experts` knowledge and experience. The preliminary PIRT has been developed by the consensus of KAERI expert panelists and AHP (Analytical Hierarchy Process). Preliminary PIRT developed in this report is intended for use to identify and integrate development areas of further experimental tests needed and thermal-hydraulic models and correlations and code improvements for the safety analysis of the SMART integral reactor. (author). 7 refs., 21 tabs., 22 figs.

  3. INTRANS. A computer code for the non-linear structural response analysis of reactor internals under transient loads

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The 'INTRANS' system is a general purpose computer code, designed to perform linear and non-linear structural stress and deflection analysis of impacting or non-impacting nuclear reactor internals components coupled with reactor vessel, shield building and external as well as internal gapped spring support system. This paper describes in general a unique computational procedure for evaluating the dynamic response of reactor internals, descretised as beam and lumped mass structural system and subjected to external transient loads such as seismic and LOCA time-history forces. The computational procedure is outlined in the INTRANS code, which computes component flexibilities of a discrete lumped mass planar model of reactor internals by idealising an assemblage of finite elements consisting of linear elastic beams with bending, torsional and shear stiffnesses interacted with external or internal linear as well as non-linear multi-gapped spring support system. The method of analysis is based on the displacement method and the code uses the fourth-order Runge-Kutta numerical integration technique as a basis for solution of dynamic equilibrium equations of motion for the system. During the computing process, the dynamic response of each lumped mass is calculated at specific instant of time using well-known step-by-step procedure. At any instant of time then, the transient dynamic motions of the system are held stationary and based on the predicted motions and internal forces of the previous instant. From which complete response at any time-step of interest may then be computed. Using this iterative process, the relationship between motions and internal forces is satisfied step by step throughout the time interval

  4. Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

    International Nuclear Information System (INIS)

    Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

    2015-01-01

    Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

  5. Utilizing elements of the CSAU phenomena identification and ranking table (PIRT) to qualify a PWR non-LOCA transients system code

    Energy Technology Data Exchange (ETDEWEB)

    Greene, K.R.; Fletcher, C.D.; Gottula, R.C.; Lindquist, T.R.; Stitt, B.D. [Framatome ANP, Richland, WA (United States)

    2001-07-01

    Licensing analyses of Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) Chapter 15 non-LOCA transients are an important part of establishing operational safety limits and design limits for nuclear power plants. The applied codes and methods are generally qualified using traditional methods of benchmarking and assessment, sample problems, and demonstration of conservatism. Rigorous formal methods for developing code and methodology have been created and applied to qualify realistic methods for Large Break Loss-of-Coolant Accidents (LBLOCA's). This methodology, Code Scaling, Applicability, and Uncertainty (CSAU), is a very demanding, resource intensive, process to apply. It would be challenging to apply a comprehensive and complete CSAU level of analysis, individually, to each of the more than 30 non-LOCA transients that comprise Chapter 15 events. However, certain elements of the process can be easily adapted to improve quality of the codes and methods used to analyze non- LOCA transients. One of these elements is the Phenomena Identification and Ranking Table (PIRT). This paper presents the results of an informally constructed PIRT that applies to non-LOCA transients for Pressurized Water Reactors (PWR's) of the Westinghouse and Combustion Engineering design. A group of experts in thermal-hydraulics and safety analysis identified and ranked the phenomena. To begin the process, the PIRT was initially performed individually by each expert. Then through group interaction and discussion, a consensus was reached on both the significant phenomena and the appropriate ranking. The paper also discusses using the PIRT as an aid to qualify a 'conservative' system code and methodology. Once agreement was obtained on the phenomena and ranking, the table was divided into six functional groups, by nature of the transients, along the same lines as Chapter 15. Then, assessment and disposition of the significant phenomena was performed. The PIRT and

  6. Transient Model of a 10 MW Supercritical CO{sub 2} Brayton Cycle for Light Water Reactors by using MARS Code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo-Hyun; Park, Hyun Sun; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of); Bae, Sung Won; Cha, Jae-Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this study, recuperation cycle was chosen as a reference loop design and the MARS code was chosen as the transient cycle analysis code. Cycle design condition is focus on operation point of the light-water reactor. Development of a transient model was performed for 10MW-electron SCO{sub 2} coupled with light water reactors. In order to perform transient analysis, cycle transient model was developed and steady-state run was performed and presented in the paper. In this study, the transient model of SCO{sub 2} recuperation Brayton cycle was developed and implemented in MARS to study the steady-state simulation. We performed nodalization of the transient model using MARS code and obtained steady-state results. This study is shown that the supercritical CO{sub 2} Brayton cycle can be used as a power conversion system for light water reactors. Future work will include transient analysis such as partial road operation, power swing, start-up, and shutdown. Cycle control strategy will be considered for various control method.

  7. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  8. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

    2016-01-15

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  9. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.

    2016-01-01

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  10. Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

  11. Contraction of information and its inverse problem in reactor system identification and stochastic diagnosis

    International Nuclear Information System (INIS)

    Kishida, K.

    1996-01-01

    Research concerning power reactor noise analysis makes rapid progress in the areas of the system identification, prediction and diagnosis. Keywords in these studies are artificial intelligence, neural network, fuzzy, and chaos. Nonlinear, nonstationary, or non-Gaussian processes as well as linear and steady processes are also studied in fluctuation analysis. However, we have not enough time to study a fundamental theory, since we are urged to obtain results or applications in power reactor fluctuations. Furthermore, we have no systematic approach to handle observed time series data in the linear process, since power reactor noise phenomena are complicated. Hence, it is important to study it from the fundamental viewpoint. It is a main aim of the present review paper to describe a unified formalism for reactor system identification and stochastic diagnosis

  12. Stability analysis of the Backward Euler time discretization for the pin-resolved transport transient reactor calculation

    International Nuclear Information System (INIS)

    Zhu, Ang; Xu, Yunlin; Downar, Thomas

    2016-01-01

    Three-dimensional, full core transport modeling with pin-resolved detail for reactor dynamic simulation is important for some multi-physics reactor applications. However, it can be computationally intensive due to the difficulty in maintaining accuracy while minimizing the number of time steps. A recently proposed Transient Multi-Level (TML) methodology overcomes this difficulty by use multi-level transient solvers to capture the physical phenomenal in different time domains and thus maximize the numerical accuracy and computational efficiency. One major problem with the TML method is the negative flux/precursor number density generated using large time steps for the MOC solver, which is due to the Backward Euler discretization scheme. In this paper, the stability issue of Backward Euler discretization is first investigated using the Point Kinetics Equations (PKEs), and the predicted maximum allowed time step for SPERT test 60 case is shown to be less than 10 ms. To overcome this difficulty, linear and exponential transformations are investigated using the PKEs. The linear transformation is shown to increase the maximum time step by a factor of 2, and the exponential transformation is shown to increase the maximum time step by a factor of 5, as well as provide unconditionally stability above a specified threshold. The two sets of transformations are then applied to TML scheme in the MPACT code, and the numerical results presented show good agreement for standard, linear transformed, and exponential transformed maximum time step between the PKEs model and the MPACT whole core transport solution for three different cases, including a pin cell case, a 3D SPERT assembly case and a row of assemblies (“striped assembly case”) from the SPERT model. Finally, the successful whole transient execution of the stripe assembly case shows the ability of the exponential transformation method to use 10 ms and 20 ms time steps, which all failed using the standard method.

  13. Transient feedback from fuel motion in metal IFR [Integral Fast Reactor] fuel

    International Nuclear Information System (INIS)

    Rhodes, E.A.; Stanford, G.S.; Regis, J.P.; Bauer, T.H.; Dickerman, C.E.

    1990-01-01

    Results from hodoscope data analyses are presented for TREAT transient-overpower tests M5 through M7 with emphasis on transient feedback mechanisms, including prefailure expansion at the tops of the fuel pins, subsequent dispersive axial fuel motion, and losses in relative worth of the fuel pins during the tests. Tests M5 and M6 were the first TOP tests of margin to cladding branch and prefailure elongation of D9-clad ternary (U-Pu-Zr) IFR-type fuel. Test M7 extended these results to high-burnup fuel and also initiated transient testing of HT9-clad binary (U-Zr) FFTF-driver fuel. Results show significant prefailure negative reactivity feedback and strongly negative feedback from fuel driven to failure. 4 refs., 6 figs

  14. Empirical method to calculate Clinch River Breeder Reactor (CRBR) inlet plenum transient temperatures

    International Nuclear Information System (INIS)

    Howarth, W.L.

    1976-01-01

    Sodium flow enters the CRBR inlet plenum via three loops or inlets. An empirical equation was developed to calculate transient temperatures in the CRBR inlet plenum from known loop flows and temperatures. The constants in the empirical equation were derived from 1/4 scale Inlet Plenum Model tests using water as the test fluid. The sodium temperature distribution was simulated by an electrolyte. Step electrolyte transients at 100 percent model flow were used to calculate the equation constants. Step electrolyte runs at 50 percent and 10 percent flow confirmed that the constants were independent of flow. Also, a transient was tested which varied simultaneously flow rate and electrolyte. Agreement of the test results with the empirical equation results was good which verifies the empirical equation

  15. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors

    International Nuclear Information System (INIS)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2015-01-01

    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  16. On the major DYN3D developments for fast reactor design and transient analysis

    International Nuclear Information System (INIS)

    Merk, B.; Kliem, S.

    2013-01-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  17. On the major DYN3D developments for fast reactor design and transient analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

    2013-07-01

    Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

  18. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  19. The importance of transient analysis in the light water reactor licensing procedure

    International Nuclear Information System (INIS)

    Izouierdo, J.M.; Villadoniga, J.I.

    1979-01-01

    The basic principles of the Nuclear Regulation are developed in the first part of this report. The achievement of the safety objective by establishing protections -that prevent or reduce the barriers failure- is analyzed. An iterative method for the definition of the systems and components safety design bases is proposed, analyzing the role of Technical Specifications in this process. The second part shows how this methodology can be used in the case of the first barrier: the fuel cladding. The safety criteria, transient clasification problems, transient analysis and its relation with surveillance and protection systems, and the role of such analysis in fuel protection design verification are discused. (author)

  20. Thermal-hydraulics analysis of a PWR reactor using zircaloy and carbide silicon reinforced with type S fibers as fuel claddings: Simulation of a channel blockage transient

    Energy Technology Data Exchange (ETDEWEB)

    Matuck, Vinicius; Ramos, Mario C.; Faria, Rochkhudson B.; Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia, E-mail: rochkdefaria@yahoo.com.br, E-mail: matuck747@gmail.com, E-mail: patricialire@yahoo.com.br, E-mail: marc5663@gmail.com, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Departamento de Engenharia Nuclear

    2017-11-01

    A detailed thermal-hydraulic reactor model using as reference data from the Angra 2 Final Safety Analysis Report (FSAR) has been developed and SiC reinforced with Hi-Nicalon type S fibers (SiC HNS) was used as fuel cladding. The goal is to compare its behavior from the thermal viewpoint with the Zircaloy, at the steady- state and transient conditions. The RELAP-3D was used to perform the thermal-hydraulic analysis and a blockage transient has been investigated at full power operation. The transient considered is related to total obstruction of a core cooling channel of one fuel assembly. The calculations were performed using a point kinetic model. The reactor behavior after this transient was analyzed and the time evolution of cladding and coolant temperatures mass flow and void fraction are presented. (author)

  1. An endothermic chemical process facility coupled to a high temperature reactor. Part II: Transient simulation of accident scenarios within the chemical plant

    International Nuclear Information System (INIS)

    Brown, Nicholas R.; Revankar, Shripad T.

    2012-01-01

    Highlights: ► Seven quantitative transient case studies were analyzed in a coupled PBMR and thermochemical sulfur cycle based hydrogen plant. ► Positive power excursion in the nuclear reactor were found for helium-inlet overcoolings. ► In all cases studied the maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. - Abstract: Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. Transient study of the operational or accident events within the coupled plant is largely absent from the literature. In this paper, seven quantitative transient case studies are analyzed. The case studies consist of: (1) feed flow failure from one section of the chemical plant to another with an accompanying parametric study of the temperature in an individual reaction chamber, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without emergency nuclear reactor shutdown, (6) total failure of the chemical plant, (7) control rod insertion in the nuclear reactor. Various parametric studies based on the magnitude of the events were also performed. The only chemical plant initiated events that caused a positive power excursion in the nuclear reactor were helium-inlet overcoolings due to process holding tank failures or reaction chamber ruptures. Even for a severe sustained overcooling, the calculated maximum fuel temperatures in the nuclear reactor were 200 K below the design basis limit. The qualitative basis for the case studies and the analysis models are summarized in part I of this paper.

  2. A proof-of-concept transient diagnostic expert system for BWRs [Boiling Water Reactors

    International Nuclear Information System (INIS)

    Yoshida, K.; Naser, J.A.

    1988-05-01

    A proof-of-concept transient diagnostic expert system has been developed to identify the cause and the type of an abnormal transient in a boiling water nuclear power plant. For this expert system development, the calculational results of the simulation code RETRAN were used as the knowledge source. The knowledge extracted from the RETRAN analyses was transformed into IF-THEN rules in the knowledge base for the expert system. An important feature of this expert system is the introduction of certainty factors to allow diagnosis even in the cases where data may be either missing or marked as invalid. To increase the capability of this diagnostic system to distinguish between similiar transients, backward chaining reasoning is used to support the forward chaining reasoning with certainty factors. Through this effort, it has been demonstrated that an expert system can be successfully used to create a transient diagnostic system. It has also successfully demonstrated that RETRAN can be used as the knowledge source for developing the knowledge base of the diagnostic system

  3. Reactor sabotage vulnerability and vital-equipment identification

    International Nuclear Information System (INIS)

    Boudreau, J.M.; Haarman, R.A.

    1982-01-01

    Two ongoing programs at Los Alamos, the Vital Area Analysis Program and the Reactor Sabotage Vulnerability Program, are discussed. The Laboratory has been providing the Nuclear Regulatory Commission with technical support in identifying the vital areas at nuclear power plants through the use of sabotage fault trees. This procedure is being expanded to provide support for the Reactor Sabotage Vulnerability Assessment Program. A re-examination of some of the original system modeling assumptions, including a survey of the applicable research, is underway. A description of the survey work and the computerized data bases being used is provided. This program is expected to result in refinements in the existing procedures

  4. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Belblidia, L.A.; Chevrier, A.

    1994-01-01

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  5. Validation of models for the analysis of the transient behavior of metallic fast reactor fuel

    International Nuclear Information System (INIS)

    Kramer, J.M.; Hughes, T.H.; Gruber, E.E.

    1989-01-01

    The Integral Fast Reactor (IFR) concept being developed at Argonne National Laboratory has prompted a renewed interest in U-Pu-Zr metal alloys as a fuel for sodium-cooled fast reactors. Part of the attractiveness of the IFR concept is the improvement in reactor safety margins through inherent features of a metal-fueled LMR core. In order to demonstrate these safety margins it is necessary to have computer codes available to analyze the detailed response of metallic fuel to a wide range of accident initiators. Two of the codes that play a key role in assessing this response are the STARS fission gas behavior code and the FPIN2 fuel pin mechanics code. Verification and validation are two important components in the development of models and computer codes. Verification demonstrates through comparison of calculations with analytical solutions that the methodology and algorithms correctly solve the equations that govern the phenomena being modeled. Validation, on the other hand, demonstrates through comparison with data that the phenomena are being modeled correctly. Both components are necessary in order to have the confidence to extrapolate the calculations to reactor accident conditions. This paper presents the results of recent progress in the validation of models for the analysis of the behavior of metallic fast reactor fuel. 9 refs., 7 figs

  6. Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Raymond, P.; Caruge, D.; Paik, H.J.

    1994-01-01

    The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

  7. Application of synthesis methods to two-dimensional fast reactor transient study

    International Nuclear Information System (INIS)

    Izutsu, Sadayuki; Hirakawa, Naohiro

    1978-01-01

    Space time synthesis and time synthesis codes were developed and applied to the space-dependent kinetics benchmark problem of a two-dimensional fast reactor model, and it was found both methods are accurate and economical for the fast reactor kinetics study. Comparison between the space time synthesis and the time synthesis was made. Also, in space time synthesis, the influence of the number of trial functions on the error and on the computing time and the effect of degeneration of expansion coefficients are investigated. The matrix factorization method is applied to the inversion of the matrix equation derived from the synthesis equation, and it is indicated that by the use of this scheme space-dependent kinetics problem of a fast reactor can be solved efficiently by space time synthesis. (auth.)

  8. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  9. Identification of metabolites during biodegradation of pendimethalin in bioslurry reactor

    International Nuclear Information System (INIS)

    Ramakrishna, M.; Venkata Mohan, S.; Shailaja, S.; Narashima, R.; Sarma, P.N.

    2008-01-01

    Bioslurry phase reactor was used for the degradation of pendimethalin, a pre-emergence herbicide in the contaminated soil under aerobic environment. More than 91% degradation of pendimethalin was observed for 5 days of reactor operation augmented with sewage from effluent treatment plant (ETP). The performance of the reactor was monitored regularly by measuring pH and colony forming units (CFU). The metabolites of pendimethalin formed during degradation were identified using various analytical techniques, viz., thin layer chromatography (TLC), high performance liquid chromatography (HPLC) and liquid chromatography-mass spectroscopy (LC-MS/MS). Four metabolites were formed and identified as N-(1-ethylpropyl)-3,4-dicarboxy 2,6-dinitrobenzenamine-N-oxide, N-(1-ethylpropyl)-3,4-dimethoxy-2,6-dinitrobenzenamine and benezimadazole-7-carboxyaldehyde. The reactions involved were monohydrolysis of 2-methyl groups followed by dihydrolysis. Further oxidation of amine groups and hydroxylation of propyl groups produced the above said metabolites. Degradation pathway of pendimethalin has been proposed in the bioslurry phase reactor

  10. Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

    International Nuclear Information System (INIS)

    1996-10-01

    This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

  11. Utilization of the voltage frequency converter or digital representation and documentation of transient reactor operation

    International Nuclear Information System (INIS)

    Doane, Harry J.

    1986-01-01

    The ease and speed of handling transient data is enhanced by the use of a voltage to frequency converter (VFC). This analogue to digital semiconductor device provides an inexpensive and portable alternative to electro-mechanical recorders and hand entry of data into computer codes. The VFC used at The University of Arizona is a Teledyne Philbrick 4705/01. A zero to positive ten volt input signal provides a zero to one megahertz output signal which is TTL/DTL compatible. VFC is used at the University of Arizona to collect data for super prompt critical TRIGA excursions. The VFC provides a low cost, convenient method of transient data storage and retrieval for experimentation and laboratory demonstration

  12. The dynamic behavior of the SUPER-PHENIX reactor under unprotected transient

    International Nuclear Information System (INIS)

    Gouriou, A.; Francillon, E.; Kayser, G.; Malenfer, G.; Languille, A.

    1982-01-01

    Due to design changes and progress on the knowledge of feed-back effects, a reactualization of the dynamic behavior of SUPER-PHENIX under unprotected transients was undertaken. We present the main data on feed-back characteristics and the results of dynamic calculations. With the present state of knowledge, the former conclusion is confirmed: the dynamic evolution is very slow and no irreversible phenomena happen in the short term

  13. CRISTE - a subcomputer code for axial distribution, transient, of temperatures in a reactor channel of PWR

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Roberty, N.C.; Carmo, E.G.D. do.

    1983-12-01

    The subroutine CRISTE was developed to calculate the temperature distribution for transients in a PWR coolant. The Crank-Nicholson approximation was used for the temporal discretization and a semi-analytical spatial solution was obtained. The temperature in the cladding was simulated by a routine adapted from the permanent distribution, and was used in on iterative method, following CRISTE subroutine. (E.G.) [pt

  14. 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

  15. Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

    2003-01-01

    The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

  16. Patterns identification in supervisory systems of nuclear reactors installations and gas pipelines systems using self-organizing maps; Identificacao de padroes em sistemas supervisorios de instalacoes de reatores nucleares e em sistemas de gasodutos utilizando mapas auto-organizaveis

    Energy Technology Data Exchange (ETDEWEB)

    Doraskevicius Junior, Waldemar

    2005-07-01

    Self-Organizing Maps, SOM, of Kohonen were studied, implemented and tested with the aim of developing, for the energy branch, an effective tool especially for transient identification in nuclear reactors and for gas pipelines networks logistic supervision, by classifying operations and identifying transients or abnormalities. The digital system for the test was developed in Java platform, for the portability and scalability, and for belonging to free development platforms. The system, executed in personal computers, showed satisfactory results to aid in decision taking, by classifying IRIS (International Reactor Innovative and Secure) reactor operation conditions (data from simulator) and by classifying Southeast (owner: TRANSPETRO - Brazil) gas pipeline network. Various adaptations were needed for such business, as new topologies for the output layer of artificial neural network and particular preparation for the input data. (author)

  17. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Peng, C.M.; Maly, J.

    1988-01-01

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  18. Application of sensitivity analysis to a simplified coupled neutronic thermal-hydraulics transient in a fast reactor using Adjoint techniques

    International Nuclear Information System (INIS)

    Gilli, L.; Lathouwers, D.; Kloosterman, J.L.; Van der Hagen, T.H.J.J.

    2011-01-01

    In this paper a method to perform sensitivity analysis for a simplified multi-physics problem is presented. The method is based on the Adjoint Sensitivity Analysis Procedure which is used to apply first order perturbation theory to linear and nonlinear problems using adjoint techniques. The multi-physics problem considered includes a neutronic, a thermo-kinetics, and a thermal-hydraulics part and it is used to model the time dependent behavior of a sodium cooled fast reactor. The adjoint procedure is applied to calculate the sensitivity coefficients with respect to the kinetic parameters of the problem for two reference transients using two different model responses, the results obtained are then compared with the values given by a direct sampling of the forward nonlinear problem. Our first results show that, thanks to modern numerical techniques, the procedure is relatively easy to implement and provides good estimation for most perturbations, making the method appealing for more detailed problems. (author)

  19. The code DYN3DR for steady-state and transient analyses of light water reactor cores with Cartesian geometry

    International Nuclear Information System (INIS)

    Grundmann, U.

    1995-11-01

    The code DYN3D/M2 was developed for 3-dimensional steady-state and transient analyses of reactor cores with hexagonal fuel assemblies. The neutron kinetics of the new version DYN3DR is based on a nodal method for the solution of the 3-dimensional 2-group neutron diffusion equation for Cartesian geometry. The thermal-hydraulic model FLOCAL simulating the two phase flow of coolant and the fuel rod behaviour is used in the two versions. The fundamentals for the solution of the neutron diffusion equations in DYN3DR are described. The 3-dimensional NEACRP benchmarks for rod ejections in LWR with quadratic fuel assemblies were calculated and the results were compared with the published solutions. The developed algorithm for neutron kinetics are suitable for using parallel processing. The behaviour of speed-up versus the number of processors is demonstrated for calculations of a static neutron flux distribution using a workstation with 4 processors. (orig.) [de

  20. Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

    1996-01-01

    The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

  1. A possibilistic approach for transient identification with 'don't know' response capability optimized by genetic algorithm

    International Nuclear Information System (INIS)

    Almeida, Jose Carlos S. de; Schirru, Roberto; Pereira, Claudio M.N.A.; Universidade Federal, Rio de Janeiro, RJ

    2002-01-01

    This work describes a possibilistic approach for transient identification based on the minimum centroids set method, proposed in previous work, optimized by genetic algorithm. The idea behind this method is to split the complex classification problem into small and simple ones, so that the performance in the classification can be increased. In order to accomplish that, a genetic algorithm is used to learn, from realistic simulated data, the optimized time partitions, which the robustness and correctness in the classification are maximized. The use of a possibilistic classification approach propitiates natural and consistent classification rules, leading naturally to a good heuristic to handle the 'don't know 'response, in case of unrecognized transient, which is fairly desirable in transient classification systems where safety is critical. Application of the proposed approach to a nuclear transient indentification problem reveals good capability of the genetic algorithm in learning optimized possibilistic classification rules for efficient diagnosis including 'don't know' response. Obtained results are shown and commented. (author)

  2. Identification of improvements of advanced light water reactor concepts

    International Nuclear Information System (INIS)

    Frisch, W.; Liesch, K.; Riegel, B.

    1993-01-01

    The scope of this report is to identify the improvement of reactor developments with respect to reactor safety. This includes the collection of non-proprietary information on the description of the advanced design characteristics, especially summary design descriptions and general publications. This documentation is not intended to include a safety evaluation of the advanced concepts; however, it is structured in such a way that it can serve as a basis for a future safety evaluation. This is taken into account in the structure of the information regarding the distinction of the various concepts with respect to their 'advancement' and the classification of design characteristics according to some basic safety aspects. The overall description concentrates on those features which are relevant to safety. Other aspects, such as economy, operational features, maintenance, the construction period, etc...are not considered explicitly in this report

  3. Modal method for crack identification applied to reactor recirculation pump

    International Nuclear Information System (INIS)

    Miller, W.H.; Brook, R.

    1991-01-01

    Nuclear reactors have been operating and producing useful electricity for many years. Within the last few years, several plants have found cracks in the reactor coolant pump shaft near the thermal barrier. The modal method and results described herein show the analytical results of using a Modal Analysis test method to determine the presence, size, and location of a shaft crack. The authors have previously demonstrated that the test method can analytically and experimentally identify shaft cracks as small as five percent (5%) of the shaft diameter. Due to small differences in material property distribution, the attempt to identify cracks smaller than 3% of the shaft diameter has been shown to be impractical. The rotor dynamics model includes a detailed motor rotor, external weights and inertias, and realistic total support stiffness. Results of the rotor dynamics model have been verified through a comparison with on-site vibration test data

  4. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  5. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    International Nuclear Information System (INIS)

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

  6. Transient signal analysis in power reactors by means of the wavelet technique

    International Nuclear Information System (INIS)

    Wentzeis, Luis

    1999-01-01

    The application of the wavelet technique, had enabled to study the time evolution of the properties (amplitude and frequency content) of a signals set, measured in the Embalse nuclear power plant (CANDU 600 M we), in the low frequency range and for different operating conditions. Particularly, by means of this technique, we studied the time evolution of the signals in the non-stationary state of the reactor (during a raise in power), where the Fourier analysis results inadequate. (author)

  7. Safety analysis of Atucha 1 reactor pressure vessel for a typical transient

    International Nuclear Information System (INIS)

    Chomik, E.; Jinchuk, D.

    1994-01-01

    As a consequence of disturbances on the CNA I external electric grid some incidents were produced in a 6 minutes lapse, causing a sudden cooling of the primary system, while pressure was maintained nearly constant. On the basis of this event, a safety analysis based on the LInear Elastic Fracture Mechanics was carried out. This paper presents an alternative method for the calculation of transients; the Finite Element Method, particularly, the OCA-II FEM code. By using this method it was possible to demonstrate, for this event, a safe operating condition for the end of life of the RPV, with regard to brittle fracture risk. 6 refs, 11 figs, 1 tab

  8. Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor

    International Nuclear Information System (INIS)

    Saavedra, Fernando M.

    2001-01-01

    The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es

  9. Transient thermal-hydraulic/neutronic analysis in a VVER-1000 reactor core

    International Nuclear Information System (INIS)

    Seyed khalil Mousavian; Mohammad Mohsen Ertejaei; Majid Shahabfar

    2005-01-01

    Full text of publication follows: Nowadays, coupled thermal-hydraulic and three-dimensional neutronic codes in order to consider different feedback effects is state of the art subject in nuclear engineering researches. In this study, RELAP5/COBRA and WIMS/CITATION codes are implemented to investigate the VVER-1000 reactor core parameters during Large Break Loss of Coolant Accident (LB-LOCA). In a LB-LOCA, the primary side pressure, coolant density and fuel temperature strongly decrease but the cladding temperature experiences a strong peak. For this purpose, the RELAP5 Best Estimate (BE) system code is used to simulate the LB-LOCA analysis in VVER-1000 nuclear thermal-hydraulic loops. Also, the modified COBRA-IIIc software as a sub-channel analysis code is applied for modeling of VVER-1000 reactor core. Moreover, WIMS and CITATION as a cross section and 3-D neutron flux codes are coupled with thermal-hydraulic codes with the aim of consider the spatial effects through the reactor core. For this reason, suitable software is developed to link and speed up the coupled thermalhydraulic and three-dimensional neutronic calculations. This software utilizes of external coupling concept in order to integrate thermal-hydraulic and neutronic calculations. (authors)

  10. Transient regimes in a heavy water reactor; Regimes transitoires dans un reacteur a eau lourde

    Energy Technology Data Exchange (ETDEWEB)

    Raievski, V [Commissariat a l' Energie Atomique, Saclay(France). Centre d' Etudes Nucleaires

    1953-07-01

    We studied the variations of power and reactivity of a reactor when we raise in a continuous way the starting plates. During the subcritical regime (negative reactivity), the power is determined by reactivity and by the intensity of the sources of photo neutrons, produced during the previous work of the reactor. When, during the rise of the plates, the reactor, pass by the critical regime (zero reactivity), one notes that the reached power is independent of the initial reactivity. During the sur-critical regime (positive reactivity), the elevation of temperature of the uranium bars slows down the growth of reactivity due to the movements of the plates. The power stretches then toward a value that depends only on the regime of cooling of the reactor and the excess of the available reactivity. This survey permits to choose such a rise speed, that reactivity remains constantly lower to a value beyond which the piloting of the reactor becomes difficult. This result is not more valid, if the intensity of the sources is insufficient, what takes place during the first divergences and after a stop of long length. (author) [French] On etudie les variations de puissance et de reactivite d'un reacteur quand on leve d'une facon continue les plaques de demarrage. Pendant le regime subcritique (reactivite negative), la puissance est determinee par la reactivite et par l'intensite des sources de photoneutrons, produites pendant la marche anterieure du reacteur. Quand, au cours de la montee des plaques, le reacteur passe par le regime critique (reactivite nulle), on constate que la puissance atteinte est independante de la reactivite initiale. Pendant le regime surcritique (reactivite positive), l'elevation de temperature des barres d'uranium ralentit l'accroissement de reactivite due aux mouvements des plaques. La puissance tend alors vers une valeur qui ne depend plus que du regime de refroidissement du reacteur et de l'exces de la reactivite disponible. Cette etude permet de

  11. Reloading pattern optimization of VVER-1000 reactors in transient cycles using genetic algorithm

    International Nuclear Information System (INIS)

    Rahmani, Yashar

    2017-01-01

    Highlights: • The genetic algorithm (GA) and the innovative weighting factors method were used. • The coupling of WIMSD5-B and CITATION-LDI2 neutronic codes with the thermohydraulic WERL code was employed. • Optimization of reloading patterns was carried out in two states. • First an arrangement with satisfactory excess reactivity and the flattest power distribution was searched. • Second, it is tried to obtain an arrangement with satisfactory safety threshold and the maximum K_e_f_f. - Abstract: The present paper proposes application of the genetic algorithm (GA) and the innovative weighting factor method to optimize the reloading pattern of Bushehr VVER-1000 reactor in the second cycle. To estimate the composition of fuel assemblies remaining from the first cycle and precisely calculate the objective parameters of each reloading pattern in the second cycle, coupling of WIMSD5-B and CITATION-LDI2 codes in the neutronic section and the WERL code in the thermo-hydraulic section was employed. Optimization of the reloading patterns was carried out in two states. To meet the mentioned objective, with application of the weighting factor method in the first state, the type and quantity of the loadable fresh assemblies were determined to enable the reactor core to maintain the core criticality over the entire cycle length. Afterwards, the genetic algorithm was used to optimize the reloading pattern of the reactor to obtain an arrangement with flat radial power distribution. In the second state, the optimization algorithm was free to select the type and number of fresh fuel assemblies to be able to search for an arrangement with the maximum effective multiplication factor and the safe power peaking factor. In addition, in order to ensure the safety and desirability of the proposed patterns in both states, a time-dependent examination of the thermo-neutronic behavior of the reactor core was carried out during the second cycle. With consideration of the new

  12. Identification of the reduced order models of a BWR reactor; Identificacion de modelos de orden reducido de un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez S, A. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: augusto@correo.unam.mx

    2004-07-01

    The present work has as objective to analyze the relative stability of a BWR type reactor. It is analyzed that so adaptive it turns out to identify the parameters of a model of reduced order so that this it reproduces a condition of given uncertainty. This will take of a real fact happened in the La Salle plant under certain operation conditions of power and flow of coolant. The parametric identification is carried out by means of an algorithm of recursive least square and an Output Error model (Output Error), measuring the output power of the reactor when the instability is present, and considering that it is produced by a change in the reactivity of the system in the same way that a sign of type step. Also it is carried out an analytic comparison of the relative stability, analyzing two types of answers: the original answer of the uncertainty of the reactor vs. the obtained response identifying the parameters of the model of reduced order, reaching the conclusion that it is very viable to adapt a model of reduced order to study the stability of a reactor, under the only condition to consider that the dynamics of the reactivity is of step type. (Author)

  13. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    International Nuclear Information System (INIS)

    Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ernestova, M.; Zamboch, M.; Seifert, H.P.; Ritter, S.; Roth, A.; Devrient, B.; Ehrnsten, U.

    2004-01-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  14. Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

    International Nuclear Information System (INIS)

    Marra Neto, A.; Silva, A.T. e; Sabundjian, G.; Freitas, R.L.; Neves Conti, T. das.

    1991-09-01

    The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

  15. Proposal of an ISO Standard: Classification of Transients and Accidents for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Chung, Bub Dong; Lee, Doo-Jeong; Kim, Jong In; Yoon, Ju Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Jeong, Jae Jun [Pusan National Univ., Busan (Korea, Republic of); Kim, An Sup; Lee, Sang Yoon [Korea Electric Association, Seoul (Korea, Republic of)

    2016-05-15

    Classification of the events for a nuclear power plant is a fundamental basis for defining nuclear safety functions, safety systems performing those functions, and specific acceptance criteria for safety analyses. Presently, the approaches for the event classification adopted by the nuclear suppliers are different, which makes a nuclear technology trade barrier. The IAEA and WENRA are making efforts to establish general requirements or guidelines on the classification of either plant states or defence-in-depth levels for the design of nuclear power plants. However, the requirements and guidelines do not provide the details for practical application to various types of commercial PWRs. Recently, Korea proposed a new ISO standardisation project to develop a harmonized or consolidated international standard for classifying the events in PWRs and for defining (or imposing) the acceptance criteria for reactor design and/or radiation protection corresponding to each event class. This paper briefs the method with strategies for developing the standard, the current various practices of the PWR event classification and acceptance criteria developed or adopted by several organizations in USA and Europe, and a draft of the proposed standard. The proposed standard will affect all the relevant stakeholders such as reactor designers, vendors, suppliers, utilities, regulatory bodies, and publics of the leading countries in the area of nuclear industry as well as utilities, regulatory bodies, and publics of the newly entering (starting) countries. It is expected that all of the stakeholders will benefit from the proposed deliverable which provides an internationally harmonized standard for classifying the PWR events as follows: The reactor design bases for assuring safety and related technical information can be effectively communicated and shared among them resulting in enhancement of the global nuclear safety and fosterage of the global nuclear trade. The countries starting

  16. Identification of speech transients using variable frame rate analysis and wavelet packets.

    Science.gov (United States)

    Rasetshwane, Daniel M; Boston, J Robert; Li, Ching-Chung

    2006-01-01

    Speech transients are important cues for identifying and discriminating speech sounds. Yoo et al. and Tantibundhit et al. were successful in identifying speech transients and, emphasizing them, improving the intelligibility of speech in noise. However, their methods are computationally intensive and unsuitable for real-time applications. This paper presents a method to identify and emphasize speech transients that combines subband decomposition by the wavelet packet transform with variable frame rate (VFR) analysis and unvoiced consonant detection. The VFR analysis is applied to each wavelet packet to define a transitivity function that describes the extent to which the wavelet coefficients of that packet are changing. Unvoiced consonant detection is used to identify unvoiced consonant intervals and the transitivity function is amplified during these intervals. The wavelet coefficients are multiplied by the transitivity function for that packet, amplifying the coefficients localized at times when they are changing and attenuating coefficients at times when they are steady. Inverse transform of the modified wavelet packet coefficients produces a signal corresponding to speech transients similar to the transients identified by Yoo et al. and Tantibundhit et al. A preliminary implementation of the algorithm runs more efficiently.

  17. A two-dimensional model for transients calculations with phase changes in sodium cooled reactors

    International Nuclear Information System (INIS)

    Granziera, M.R.

    1981-01-01

    A computer code (NATOF2D) for the numerical simulation of situations where the radial non-uniformity in the sodium flow is an important factor, was developed. This computer code uses the two-fluid model, in which each phase is described by a complete set of mass conservation equations, energy equations and momentum equations. The experiment SLSF-P3A realized in the Engineering Test Reactor, Idaho, during the period of july to september of 1977, was simulated. (E.G.) [pt

  18. Computer methods for transient fluid-structure analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Belytschko, T.; Liu, W.K.

    1985-01-01

    Fluid-structure interaction problems in nuclear engineering are categorized according to the dominant physical phenomena and the appropriate computational methods. Linear fluid models that are considered include acoustic fluids, incompressible fluids undergoing small disturbances, and small amplitude sloshing. Methods available in general-purpose codes for these linear fluid problems are described. For nonlinear fluid problems, the major features of alternative computational treatments are reviewed; some special-purpose and multipurpose computer codes applicable to these problems are then described. For illustration, some examples of nuclear reactor problems that entail coupled fluid-structure analysis are described along with computational results

  19. Comparison of one-dimensional and point kinetics for various light water reactor transients

    International Nuclear Information System (INIS)

    Naser, J.A.; Lin, C.; Gose, G.C.; McClure, J.A.; Matsui, Y.

    1985-01-01

    The object of this paper is to compare the results from the three kinetics options: 1) point kinetics; 2) point kinetics by not changing the shape function; and 3) one-dimensional kinetics for various transients on both BWRs and PWRs. A systematic evaluation of the one-dimensional kinetics calculation and its alternatives is performed to determine the status of these models and to identify additional development work. In addition, for PWRs, the NODEP-2 - NODETRAN and SIMULATE - SIMTRAN paths for calculating kinetics parameters are compared. This type of comparison has not been performed before and is needed to properly evaluate the RASP methodology of which these codes are a part. It should be noted that RASP is in its early pre-release stage and this is the first serious attempt to examine the consistency between these two similar but different methods of generating physics parameters for the RETRAN computer code

  20. Evidence of circadian and extended shift effects on reactor transient frequency

    International Nuclear Information System (INIS)

    Maloney, S.

    1992-01-01

    An extensive body of knowledge exists documenting the significant swings in error rates, perception, judgment, and overall alertness levels in the course of a day. The literature also demonstrates pronounced differences in performance levels whereby night workers routinely underperform day workers in any set of tasks. The performance split widens with task complexity. Rotating shift workers, such as nuclear power plant operators, have been shown to experience performance problems comparable to straight night workers. Finally, sleep research also documents that extended hours, such as the so-called 4-day work week, can undermine alertness levels, particularly in the last 2 to 3 h of a daily shift. These issues suggest that cyclical performance by shift workers may be evident in nuclear plant operators. This paper seeks evidence of such cyclical performance by examining operating transients at nuclear power plants involving human error

  1. Development and Application of a Plant Code to the Analysis of Transients in Integrated Reactors

    International Nuclear Information System (INIS)

    Rabiti, A.; Gimenez, M.; Delmastro, D.; Zanocco, P.

    2003-01-01

    In this work, a secondary system model for a CAREM-25 type nuclear power plant was developed.A two-phase flow homogenous model was used and found adequate for the scope of the present work.A finite difference scheme was used for the numerical implementation of the model.This model was coupled to the HUARPE code, a primary circuit code, in order to obtain a plant code.This plant code was used to analyze the inherent response of the system, without control feedback loops, for a transient of steam generator feed-water mass flow reduction.The results obtained are satisfactory, but a validation against other plant codes is still necessary

  2. CORA. A thermal and hydraulic transient analysis computer code for a cluster of reactor core assemblies

    International Nuclear Information System (INIS)

    Johnson, H.G.

    1982-01-01

    The Fast Flux Test Facility (FFTF) is arranged for natural circulation emergency core cooling in the event of loss of all plant electrical power. This design feature was conclusively demonstrated in a series of four natural circulation transient tests during the plant startup testing program in 1980 and 1981. Predictions, of core performance during these tests were made using the Westinghouse Hanford Company CORA computer program. The predictions, which compared well with measured plant data, were used in the extrapolation process to demonstrate the validity of the FFTF plant safety models and codes. This paper provides a brief description of the CORA code and includes typical comparisons of predictions to measured plant test data

  3. An improved model to simulate pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    2004-01-01

    Among those theories to interpret the PWR iodine spiking behaviors, the most accepted concept is based on steam formation and condensation in damaged fuel rods. Due to the complex nature of the phenomenon, a comprehensive model of the iodine behavior has not yet been successfully developed. In 1992 a new empirical model was introduced to establish a correlation with the operating parameters. The comparison results of the predicted iodine-131 equivalent activity value with the operating radiochemistry database was off by 23%. This paper presents an improved model. Although it is still an empirical model which also gives a first order estimation of the peak iodine spiking magnitude, the deviation between prediction and measurement was reduced to ∼7%. It is believed that this improved model can be used for better prediction and control of the iodine spiking magnitude resulted from failed fuel rods during power transients or plant shutdown. (author)

  4. Analysis on transient hydrodynamic characteristics of cavitation process for reactor coolant pump

    International Nuclear Information System (INIS)

    Wang Xiuli; Wang Peng; Yuan Shouqi; Zhu Rongsheng; Fu Qiang

    2014-01-01

    The reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX, and the corresponding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform. The results show that gas content presents the law of exponential function with the pressure reduction or time increase. In the cavitation primary condition, the pulsation frequency of head for the reactor coolant pump is mainly low frequency, and the main frequency of pressure pulsation is still rotation frequency while the effect of the pressure pulsation caused by cavitation on main frequency is not obvious. With the development of cavitation, the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency, secondary frequency and pulsating amplitude while the head pulsation frequency is given priority to low frequency pulse. Under serious cavitation condition, the head pulsation frequency is given priority to irregular changes of pulse high frequency, and also contains almost regular changes of low frequency. (authors)

  5. Decay heat removal and transient analysis in accidental conditions in the EFIT reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Meloni, P.; Polidori, M.; Casamirra, M.; Castiglia, F.; Giardina, M.

    2007-01-01

    The development of a conceptual design of an industrial scale transmutation facility (EFIT) of several 100 MW thermal power based on Accelerator Driven System (ADS) is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related Decay Heat Removal (DHR) system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which lead to the Loss of Heat Sink (LOHS). In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1-D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios. (author)

  6. Decay Heat Removal and Transient Analysis in Accidental Conditions in the EFIT Reactor

    Directory of Open Access Journals (Sweden)

    Giacomino Bandini

    2008-01-01

    Full Text Available The development of a conceptual design of an industrial-scale transmutation facility (EFIT of several 100 MW thermal power based on accelerator-driven system (ADS is addressed in the frame of the European EUROTRANS Integral Project. In normal operation, the core power of EFIT reactor is removed through steam generators by four secondary loops fed by water. A safety-related decay heat removal (DHR system provided with four independent inherently safe loops is installed in the primary vessel to remove the decay heat by natural convection circulation under accidental conditions which are caused by a loss-of-heat sink (LOHS. In order to confirm the adequacy of the adopted solution for decay heat removal in accidental conditions, some multi-D analyses have been carried out with the SIMMER-III code. The results of the SIMMER-III code have been then used to support the RELAP5 1D representation of the natural circulation flow paths in the reactor vessel. Finally, the thermal-hydraulic RELAP5 code has been employed for the analysis of LOHS accidental scenarios.

  7. Improvements and validation of the transient analysis code MOREL for molten salt reactors

    International Nuclear Information System (INIS)

    Zhuang Kun; Zheng Youqi; Cao Liangzhi; Hu Tianliang; Wu Hongchun

    2017-01-01

    The liquid fuel salt used in the molten salt reactors (MSRs) serves as the fuel and coolant simultaneously. On the one hand, the delayed neutron precursors circulate in the whole primary loop and part of them decay outside the core. On the other hand, the fission heat is carried off directly by the fuel flow. These two features require new analysis method with the coupling of fluid flow, heat transfer and neutronics. In this paper, the recent update of MOREL code is presented. The update includes: (1) the improved quasi-static method for the kinetics equation with convection term is developed. (2) The multi-channel thermal hydraulic model is developed based on the geometric feature of MSR. (3) The Variational Nodal Method is used to solve the neutron diffusion equation instead of the original analytic basis functions expansion nodal method. The update brings significant improvement on the efficiency of MOREL code. And, the capability of MOREL code is extended for the real core simulation with feedback. The numerical results and experiment data gained from molten salt reactor experiment (MSRE) are used to verify and validate the updated MOREL code. The results agree well with the experimental data, which prove the new development of MOREL code is correct and effective. (author)

  8. Cassandre : a two-dimensional multigroup diffusion code for reactor transient analysis

    International Nuclear Information System (INIS)

    Arien, B.; Daniels, J.

    1986-12-01

    CASSANDRE is a two-dimensional (x-y or r-z) finite element neutronics code with thermohydraulics feedback for reactor dynamics prior to the disassembly phase. It uses the multigroup neutron diffusion theory. Its main characteristics are the use of a generalized quasistatic model, the use of a flexible multigroup point-kinetics algorithm allowing for spectral matching and the use of a finite element description. The code was conceived in order to be coupled with any thermohydraulics module, although thermohydraulics feedback is only considered in r-z geometry. In steady state criticality search is possible either by control rod insertion or by homogeneous poisoning of the coolant. This report describes the main characterstics of the code structure and provides all the information needed to use the code. (Author)

  9. Local Fission Gas Release and Swelling in Water Reactor Fuel during Slow Power Transients

    DEFF Research Database (Denmark)

    Mogensen, Mogens Bjerg; Walker, C.T.; Ray, I.L.F.

    1985-01-01

    Gas release and fuel swelling caused by a power increase in a water reactor fuel (burn-up 2.7–4.5% FIMA) is described. At a bump terminal level of about 400 W/cm (local value) gas release was 25–40%. The formation of gas bubbles on grain boundaries and their degree of interlinkage are the two...... factors that determine the level of fission gas release during a power bump. Release begins when gas bubbles on grain boundaries start o interlink. This occurred at r/r0 ~ 0.75. Release tunnels were fully developed at r/r0 ~ 0.55 with the result that gas release was 60–70% at this position....

  10. Lecture background notes on transient sodium boiling and voiding in fast reactors

    International Nuclear Information System (INIS)

    Okrent, D.; Fauske, H.K.

    1972-01-01

    This set of lecture background notes includes the following: (1) Introductory remarks on fast reactor safety, which are intended to provide some perspective on the role played by sodium boiling. (2) A discussion of superheat which reviews the experimental data and nucleation models with emphasis on the pressure-temperature history effect on radius of active cavity sites, including the role played by inert gas. (3) A discussion of the growth and collapse of spherical bubbles. (4) A historical description of the development of computer codes to describe voiding and a detailed description of the analytical formulation of typical models for calculating voiding due to boiling, fission gas release, and molten fuel-coolant interaction. (U.S.)

  11. CYLFUX, Fast Reactor Reactivity Transients Simulation in LWR by 2-D 2 Group Diffusion

    International Nuclear Information System (INIS)

    Schmidt, A.

    1973-01-01

    1 - Nature of physical problem solved: A 2-dimensional calculation of the 2-group, space-dependent neutron diffusion equations is performed in r-z geometry using an arbitrary number of groups of delayed neutron precursors. The program is designed to simulate fast reactivity excursions in light water reactors taking into account Doppler feedback via adiabatic heatup of fuel. Axial motions of control rods may be considered including scram action on option. 2 - Method of solution: The differential equations are solved at each time step by an explicit finite difference method using two time levels. The stationary distributions are obtained by using the same algorithm. 3 - Restrictions on the complexity of the problem: No restriction to the number of space points and delayed neutron energy groups besides the computer size

  12. Limitations of power conversion systems under transient loads and impact on the pulsed tokamak power reactor

    International Nuclear Information System (INIS)

    Sager, G.T.; Wong, C.P.C.; Kapich, D.D.; McDonald, C.F.; Schleicher, R.W.

    1993-11-01

    The impact of cyclic loading of the power conversion system of a helium-cooled, pulsed tokamak power plant is assessed. Design limits of key components of heat transport systems employing Rankie and Brayton thermodynamic cycles are quantified based on experience in gas-cooled fission reactor design and operation. Cyclic loads due to pulsed tokamak operation are estimated. Expected performance of the steam generator is shown to be incompatible with pulsed tokamak operation without load leveling thermal energy storage. The close cycle gas turbine is evaluated qualitatively based on performance of existing industrial and aeroderivative gas turbines. Advances in key technologies which significantly improve prospects for operation with tokamak fusion plants are reviewed

  13. Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

    International Nuclear Information System (INIS)

    Sankaranarayanan, S.

    2004-01-01

    Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

  14. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    Science.gov (United States)

    Gross, Kenny C.

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as "background" gases, further reducing the number of trial node combinations. Lastly, a "fuzzy" set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements.

  15. Expert system for identification of simultaneous and sequential reactor fuel failures with gas tagging

    International Nuclear Information System (INIS)

    Gross, K.C.

    1994-01-01

    Failure of a fuel element in a nuclear reactor core is determined by a gas tagging failure detection system and method. Failures are catalogued and characterized after the event so that samples of the reactor's cover gas are taken at regular intervals and analyzed by mass spectroscopy. Employing a first set of systematic heuristic rules which are applied in a transformed node space allows the number of node combinations which must be processed within a barycentric algorithm to be substantially reduced. A second set of heuristic rules treats the tag nodes of the most recent one or two leakers as ''background'' gases, further reducing the number of trial node combinations. Lastly, a ''fuzzy'' set theory formalism minimizes experimental uncertainties in the identification of the most likely volumes of tag gases. This approach allows for the identification of virtually any number of sequential leaks and up to five simultaneous gas leaks from fuel elements. 14 figs

  16. Study for identification of control rod drops in PWR reactors at any burnup step

    International Nuclear Information System (INIS)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C.

    2013-01-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  17. Study for identification of control rod drops in PWR reactors at any burnup step

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Thiago J.; Martinez, Aquilino S.; Medeiros, Jose A.C.C.; Goncalves, Alessandro C., E-mail: tsouza@nuclear.ufrj.br, E-mail: aquilino@lmp.ufrj.br, E-mail: canedo@lmp.ufrj.br, E-mail: alessandro@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The control rod drop event in PWR reactors induces an unsafe operating condition. Therefore, in a scenario of a control rod drop is important to quickly identify the rod to minimize undesirable effects. The objective of this work is to develop an on-line method for identification of control rod drop in PWR reactors. The method consists on the construction of a tool that is based on the ex-core detector responses. Therefore, it is proposed to recognize patterns in the neutron ex-core detectors responses and thus to identify on-line a control rod drop in the core during the reactor operation. The results of the study, as well as the behavior of the detector responses, demonstrated the feasibility of this method. (author)

  18. Identification of long-duration noise transients in LIGO and Virgo

    International Nuclear Information System (INIS)

    Coughlin, Michael W

    2011-01-01

    The LIGO and Virgo detectors are sensitive to a variety of noise sources, such as instrumental artifacts and environmental disturbances. The Stochastic Transient Analysis Multi-detector Pipeline has been developed to search for long-duration (t ≥ 1 s) gravitational-wave (GW) signals. This pipeline can also be used to identify environmental noise transients. Here, we present an algorithm to determine when long-duration noise sources couple into the interferometers, as well as identify what these noise sources are. We analyze the cross-power between a GW strain channel and an environmental sensor, using pattern recognition tools to identify statistically significant structure in cross-power time-frequency maps. We identify interferometer noise from airplanes, helicopters, thunderstorms and other sources. Examples from LIGO's sixth science run, S6, and Virgo's third scientific run, VSR3, are presented. (paper)

  19. Identification and localization of absorbers of variable strength in nuclear reactors

    International Nuclear Information System (INIS)

    Demaziere, C.; Andhill, G.

    2005-01-01

    This paper investigates the possibility of localising a noise source of the type 'absorber of variable strength' (or reactor oscillator) from as few as five neutron detectors evenly distributed throughout the core of a commercial nuclear reactor. The novelty of this investigation lies with the fact that the calculations are performed for a realistic 2-D heterogeneous reactor in the 2-group diffusion approximation, via the prior determination of the corresponding reactor transfer function. It is first demonstrated that the response of such a reactor to a localized perturbation deviates significantly from point-kinetics. The space-dependence of the induced neutron noise thus carries enough information about the location of the noise source, which makes it possible to determine its position from a few detector readings. The identification of the type of noise source is easily performed from the in-phase behaviour of the induced neutron noise. Different unfolding techniques are finally tested. All these techniques rely on the use of the reactor transfer function. One of these techniques is based on the comparison between the actual measured neutron noise and the neutron noise calculated for every possible location of the noise source. This technique is very reliable and almost insensitive to the contamination of the detector signals by background noise, but also extremely CPU consuming. Another technique, based on the piece-wise inversion of the reactor transfer function and requiring little CPU effort, was developed. Although this technique is much less reliable when background noise is present, this technique is useful to indicate a region of the reactor where a noise source is likely to be located

  20. Apparent resistivity for transient electromagnetic induction logging and its correction in radial layer identification

    Science.gov (United States)

    Meng, Qingxin; Hu, Xiangyun; Pan, Heping; Xi, Yufei

    2018-04-01

    We propose an algorithm for calculating all-time apparent resistivity from transient electromagnetic induction logging. The algorithm is based on the whole-space transient electric field expression of the uniform model and Halley's optimisation. In trial calculations for uniform models, the all-time algorithm is shown to have high accuracy. We use the finite-difference time-domain method to simulate the transient electromagnetic field in radial two-layer models without wall rock and convert the simulation results to apparent resistivity using the all-time algorithm. The time-varying apparent resistivity reflects the radially layered geoelectrical structure of the models and the apparent resistivity of the earliest time channel follows the true resistivity of the inner layer; however, the apparent resistivity at larger times reflects the comprehensive electrical characteristics of the inner and outer layers. To accurately identify the outer layer resistivity based on the series relationship model of the layered resistance, the apparent resistivity and diffusion depth of the different time channels are approximately replaced by related model parameters; that is, we propose an apparent resistivity correction algorithm. By correcting the time-varying apparent resistivity of radial two-layer models, we show that the correction results reflect the radially layered electrical structure and the corrected resistivities of the larger time channels follow the outer layer resistivity. The transient electromagnetic fields of radially layered models with wall rock are simulated to obtain the 2D time-varying profiles of the apparent resistivity and corrections. The results suggest that the time-varying apparent resistivity and correction results reflect the vertical and radial geoelectrical structures. For models with small wall-rock effect, the correction removes the effect of the low-resistance inner layer on the apparent resistivity of the larger time channels.

  1. Domain decomposition parallel computing for transient two-phase flow of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Choi, Hyoung Gwon [Seoul National University, Seoul (Korea, Republic of)

    2016-05-15

    KAERI (Korea Atomic Energy Research Institute) has been developing a multi-dimensional two-phase flow code named CUPID for multi-physics and multi-scale thermal hydraulics analysis of Light water reactors (LWRs). The CUPID code has been validated against a set of conceptual problems and experimental data. In this work, the CUPID code has been parallelized based on the domain decomposition method with Message passing interface (MPI) library. For domain decomposition, the CUPID code provides both manual and automatic methods with METIS library. For the effective memory management, the Compressed sparse row (CSR) format is adopted, which is one of the methods to represent the sparse asymmetric matrix. CSR format saves only non-zero value and its position (row and column). By performing the verification for the fundamental problem set, the parallelization of the CUPID has been successfully confirmed. Since the scalability of a parallel simulation is generally known to be better for fine mesh system, three different scales of mesh system are considered: 40000 meshes for coarse mesh system, 320000 meshes for mid-size mesh system, and 2560000 meshes for fine mesh system. In the given geometry, both single- and two-phase calculations were conducted. In addition, two types of preconditioners for a matrix solver were compared: Diagonal and incomplete LU preconditioner. In terms of enhancement of the parallel performance, the OpenMP and MPI hybrid parallel computing for a pressure solver was examined. It is revealed that the scalability of hybrid calculation was enhanced for the multi-core parallel computation.

  2. GeN-Foam: a novel OpenFOAM{sup ®} based multi-physics solver for 2D/3D transient analysis of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fiorina, Carlo, E-mail: carlo.fiorina@psi.ch [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland); Clifford, Ivor [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland); Aufiero, Manuele [LPSC-IN2P3-CNRS/UJF/Grenoble INP, 53 avenue des Martyrs, 38026 Grenoble Cedex (France); Mikityuk, Konstantin [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland)

    2015-12-01

    Highlights: • Development of a new multi-physics solver based on OpenFOAM{sup ®}. • Tight coupling of thermal-hydraulics, thermal-mechanics and neutronics. • Combined use of traditional RANS and porous-medium models. • Mesh for neutronics deformed according to the predicted displacement field. • Use of three unstructured meshes, adaptive time step, parallel computing. - Abstract: The FAST group at the Paul Scherrer Institut has been developing a code system for reactor analysis for many years. For transient analysis, this code system is currently based on a state-of-the-art coupled TRACE-PARCS routine. This work presents an attempt to supplement the FAST code system with a novel solver characterized by tight coupling between the different equations, parallel computing capabilities, adaptive time-stepping and more accurate treatment of some of the phenomena involved in a reactor transient. The new solver is based on OpenFOAM{sup ®}, an open-source C++ library for the solution of partial differential equations using finite-volume discretization. It couples together a multi-scale fine/coarse mesh sub-solver for thermal-hydraulics, a multi-group diffusion sub-solver for neutronics, a displacement-based sub-solver for thermal-mechanics and a finite-difference model for the temperature field in the fuel. It is targeted toward the analysis of pin-based reactors (e.g., liquid metal fast reactors or light water reactors) or homogeneous reactors (e.g., fast-spectrum molten salt reactors). This paper presents each “single-physics” sub-solver and the overall coupling strategy, using the sodium-cooled fast reactor as a test case, and essential code verification tests are described.

  3. GeN-Foam: a novel OpenFOAM"® based multi-physics solver for 2D/3D transient analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Fiorina, Carlo; Clifford, Ivor; Aufiero, Manuele; Mikityuk, Konstantin

    2015-01-01

    Highlights: • Development of a new multi-physics solver based on OpenFOAM"®. • Tight coupling of thermal-hydraulics, thermal-mechanics and neutronics. • Combined use of traditional RANS and porous-medium models. • Mesh for neutronics deformed according to the predicted displacement field. • Use of three unstructured meshes, adaptive time step, parallel computing. - Abstract: The FAST group at the Paul Scherrer Institut has been developing a code system for reactor analysis for many years. For transient analysis, this code system is currently based on a state-of-the-art coupled TRACE-PARCS routine. This work presents an attempt to supplement the FAST code system with a novel solver characterized by tight coupling between the different equations, parallel computing capabilities, adaptive time-stepping and more accurate treatment of some of the phenomena involved in a reactor transient. The new solver is based on OpenFOAM"®, an open-source C++ library for the solution of partial differential equations using finite-volume discretization. It couples together a multi-scale fine/coarse mesh sub-solver for thermal-hydraulics, a multi-group diffusion sub-solver for neutronics, a displacement-based sub-solver for thermal-mechanics and a finite-difference model for the temperature field in the fuel. It is targeted toward the analysis of pin-based reactors (e.g., liquid metal fast reactors or light water reactors) or homogeneous reactors (e.g., fast-spectrum molten salt reactors). This paper presents each “single-physics” sub-solver and the overall coupling strategy, using the sodium-cooled fast reactor as a test case, and essential code verification tests are described.

  4. Development of a digital card to simulate period transients in research reactors; Desenvolvimento de um cartao digital para simulacao da variacao do periodo em reatores de pesquisa

    Energy Technology Data Exchange (ETDEWEB)

    Masotti, Paulo Henrique Ferraz

    1999-07-01

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

  5. Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author) [pt

  6. A pilot application of the RELAP file to the steady state and transient analysis of a test section inside the BR2 reactor

    International Nuclear Information System (INIS)

    Ferri, M. G.; D'Auria, F.; Forasassi, G.; Giot, M.

    2000-01-01

    BR2 is a material test reactor sited in the Belgian Nuclear Research Centre in Mol. The main research programs carried out in BR2 are related to the safety of nuclear reactor structural materials and fuels, in normal and accidental conditions, plant lifetime evaluation and ageing of components. In this framework, a computer program that allows the performance of detailed, steady state analysis of several kinds of in-pile sections with an axisymmetrical geometry has been developed. Furthermore, comparing its results with those of the well known, extensively used, Relap5/Mod 3.2 code on a test problem has validated this program. This was performed in three steps: 1. modalisation development of a subsystem of a typical in-pile section. 2. steady state analysis and comparison with the above-mentioned program. 3. transient simulation of the same system; the considered transient consists of a loss of coolant flow. (author)

  7. Online failed fuel identification using delayed neutron detector signals in pool type reactors

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

    2011-01-01

    In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

  8. Adaptation and implementation of the TRACE code for transient analysis on designs of cooled lead fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2014-07-01

    The article describes the changes implemented in the TRACE code to include thermodynamic tables of liquid lead drawn from experimental results. He then explains the process for developing a thermohydraulic model for the prototype ALFRED and analysis of a selection of representative transient conducted within the framework of international research projects. The study demonstrates the applicability of TRACE code to simulate designs of cooled lead fast reactors and exposes the high safety margins are there in this technology to accommodate the most severe transients identified in their security study. (Author)

  9. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  11. The proposals on cooperation to foreign centers of science on thermophysical properties of reactor materials in a broad band of pressure and temperatures realized at normal transient and emergency operation activity of nuclear power plants

    International Nuclear Information System (INIS)

    Fortov, V.E.

    1996-01-01

    The proposals on cooperation in the area of thermophysical properties of reactor materials in a broad band of pressure and temperature realized at normal transient and emergency operation activity of nuclear power plants are discussed. 1 fig

  12. A novel application of wavelet based SVM to transient phenomena identification of power transformers

    International Nuclear Information System (INIS)

    Jazebi, S.; Vahidi, B.; Jannati, M.

    2011-01-01

    A novel differential protection approach is introduced in the present paper. The proposed scheme is a combination of Support Vector Machine (SVM) and wavelet transform theories. Two common transients such as magnetizing inrush current and internal fault are considered. A new wavelet feature is extracted which reduces the computational cost and enhances the discrimination accuracy of SVM. Particle swarm optimization technique (PSO) has been applied to tune SVM parameters. The suitable performance of this method is demonstrated by simulation of different faults and switching conditions on a power transformer in PSCAD/EMTDC software. The method has the advantages of high accuracy and low computational burden (less than a quarter of a cycle). The other advantage is that the method is not dependent on a specific threshold. Sympathetic and recovery inrush currents also have been simulated and investigated. Results show that the proposed method could remain stable even in noisy environments.

  13. Statistical and optimization methods to expedite neural network training for transient identification

    International Nuclear Information System (INIS)

    Reifman, J.; Vitela, E.J.; Lee, J.C.

    1993-01-01

    Two complementary methods, statistical feature selection and nonlinear optimization through conjugate gradients, are used to expedite feedforward neural network training. Statistical feature selection techniques in the form of linear correlation coefficients and information-theoretic entropy are used to eliminate redundant and non-informative plant parameters to reduce the size of the network. The method of conjugate gradients is used to accelerate the network training convergence and to systematically calculate the Teaming and momentum constants at each iteration. The proposed techniques are compared with the backpropagation algorithm using the entire set of plant parameters in the training of neural networks to identify transients simulated with the Midland Nuclear Power Plant Unit 2 simulator. By using 25% of the plant parameters and the conjugate gradients, a 30-fold reduction in CPU time was obtained without degrading the diagnostic ability of the network

  14. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  15. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  16. HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

    International Nuclear Information System (INIS)

    Giles, G.E.; DeVault, R.M.; Turner, W.D.; Becker, B.R.

    1976-05-01

    A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to provide more complete analysis of the reactor system during accidents involving reactor trip and cooling down on the auxiliary systems. Included are brief descriptions of the components of the HEXEREI code and sample HEXEREI analyses compared with analytical solutions and other heat transfer codes

  17. Scientific-technical cooperation with Russia. Transient analyses for alternative types of water-cooled reactors. Final report; WTZ mit Russland. Transientenanalysen fuer wassergekuehlte Kernreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Rohde, Ulrich [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Kozmenkov, Yaroslav [Forschungszentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung; Institute of Physics and Power Engineering, Obninsk (Russian Federation); Pivovarov, Valeri; Matveev, Yurij [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    2010-12-15

    The recently developed multi-group version DYN3D-MG of the reactor dynamics code DYN3D has been qualified for applications to water-cooled reactor concepts different from industrial PWR and BWR. An extended DYN3D version was applied to the graphite-moderated pressure tube reactor EGP-6 (NPP Bilibino) and conceptual design studies of an advanced Boiling Water Reactor with reduced moderation (RMWR) as well as the RUTA-70 reactor for low temperature heat supply. Concerning the RUTA reactor, safe heat removal by natural circulation of the coolant at low pressure has to be shown. For the corresponding validation of thermo-hydraulic system codes like ATHLET and RELAP5, experiments on flashing-induced natural circulation instabilities performed at the CIRCUS test facility at the TU Delft were simulated using the RELAP5 code. For the application to alternative water-cooled reactors, DYN3D model extensions and modifications were implemented, in particular adaptations of heat conduction and heat transfer models. Performing code-to-code comparisons with the Russian fine-mesh neutron diffusion code ACADEM contributed to the verification of DYN3D-MG. Validation has been performed by calculating reactor dynamics experiments at the NPP Bilibino. For the reactors EGP-6, RMWR and RUTA, analyses of various protected and unprotected control rod withdrawal and ejection transients were performed. The beyond design basis accident (BDBA) scenario ''Coast-down of all main coolant pumps at nominal power without scram'' for the RUTA reactor was analyzed using the code complexes DYN3D/ATHLET and DYN3D/RELAP5. It was shown, that the reactor passes over to a save asymptotic state at reduced power with coolant natural circulation. Analyzing the BDBA ''Unprotected withdrawal of a control rod group'' for the RMWR, the safety against Departure from Nucleate Boiling (DNB) could not be shown with the necessary confidence. Finally, conclusions have been drawn

  18. Identification of a cis-regulatory element by transient analysis of co-ordinately regulated genes

    Directory of Open Access Journals (Sweden)

    Allan Andrew C

    2008-07-01

    Full Text Available Abstract Background Transcription factors (TFs co-ordinately regulate target genes that are dispersed throughout the genome. This co-ordinate regulation is achieved, in part, through the interaction of transcription factors with conserved cis-regulatory motifs that are in close proximity to the target genes. While much is known about the families of transcription factors that regulate gene expression in plants, there are few well characterised cis-regulatory motifs. In Arabidopsis, over-expression of the MYB transcription factor PAP1 (PRODUCTION OF ANTHOCYANIN PIGMENT 1 leads to transgenic plants with elevated anthocyanin levels due to the co-ordinated up-regulation of genes in the anthocyanin biosynthetic pathway. In addition to the anthocyanin biosynthetic genes, there are a number of un-associated genes that also change in expression level. This may be a direct or indirect consequence of the over-expression of PAP1. Results Oligo array analysis of PAP1 over-expression Arabidopsis plants identified genes co-ordinately up-regulated in response to the elevated expression of this transcription factor. Transient assays on the promoter regions of 33 of these up-regulated genes identified eight promoter fragments that were transactivated by PAP1. Bioinformatic analysis on these promoters revealed a common cis-regulatory motif that we showed is required for PAP1 dependent transactivation. Conclusion Co-ordinated gene regulation by individual transcription factors is a complex collection of both direct and indirect effects. Transient transactivation assays provide a rapid method to identify direct target genes from indirect target genes. Bioinformatic analysis of the promoters of these direct target genes is able to locate motifs that are common to this sub-set of promoters, which is impossible to identify with the larger set of direct and indirect target genes. While this type of analysis does not prove a direct interaction between protein and DNA

  19. Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Mori, Takero

    2005-01-01

    In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

  20. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

    2009-10-01

    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  1. Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kruessmann, R., E-mail: regina.kruessmann@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Ponomarev, A.; Pfrang, W.; Struwe, D. [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Champigny, J.; Carluec, B. [AREVA, 10, rue J. Récamier, 69456 Lyon Cedex 06 (France); Schmitt, D.; Verwaerde, D. [EDF R& D, 1 avenue du général de Gaulle, 92140 Clamart (France)

    2015-04-15

    Highlights: • Comparison of different core designs for a sodium-cooled fast reactor. • Safety assessment with the code system SAS-SFR. • Unprotected Loss of Flow (ULOF) scenario. • Sodium boiling and core melting cannot be avoided. • A net negative Na void effect provides more grace time prior to local SA destruction. - Abstract: In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

  2. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  3. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); 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.; Schmitt, D. [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); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  4. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    International Nuclear Information System (INIS)

    Lazaro, A.; Schikorr, M.; Mikityuk, K.; Ammirabile, L.; Bandini, G.; Darmet, G.; Schmitt, D.; Dufour, Ph.; Tosello, A.; Gallego, E.; Jimenez, G.; Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.

    2014-01-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

  5. RA-6 reactor's probabilistic safety evaluation. Identification and selection of starting events

    International Nuclear Information System (INIS)

    Kay, J.; Chiossi, C.; Felizia, E.; Vallerga, H.; Kalejman, G.; Navarro, R.; Caruso, G.J.

    1987-01-01

    A summary of the 'Identification and selection of starting events' stage of the previous probabilistic safety evaluation of RA-6 reactor is presented. This evaluation was performed to verify if the safety criteria required for the licensing of RA-6 are met and to promote the diffusion of its meaning and usefulness with educational purposes. At this stage the starting events of RA-6 are determined and the probability that such events occur is calculated. The identification and selection of starting events is performed in two steps: determination of proposed starting events and determination of postulated starting events. The proposed starting events are determined by means of the master logic diagram (MLD) method, while the postulated starting events are obtained by grouping the proposed starting events. The simplifying hypothesis required for the application of MLD to the reactor are also formulated. The probability that the proposed and postulated starting events occur is afterwards calculated, adopting different fault models, in accordance with the nature of events that are considered. Conservative hypothesis on the characteristics of these events and the uncertainty of parameter values of those models are also formulated. The numerical values of the above mentioned probabilities are obtained by giving the parameters suitable values that are extracted from specialized publications. (Author)

  6. Identification of the autotrophic denitrifying community in nitrate removal reactors by DNA-stable isotope probing.

    Science.gov (United States)

    Xing, Wei; Li, Jinlong; Cong, Yuan; Gao, Wei; Jia, Zhongjun; Li, Desheng

    2017-04-01

    Autotrophic denitrification has attracted increasing attention for wastewater with insufficient organic carbon sources. Nevertheless, in situ identification of autotrophic denitrifying communities in reactors remains challenging. Here, a process combining micro-electrolysis and autotrophic denitrification with high nitrate removal efficiency was presented. Two batch reactors were fed organic-free nitrate influent, with H 13 CO 3 - and H 12 CO 3 - as inorganic carbon sources. DNA-based stable-isotope probing (DNA-SIP) was used to obtain molecular evidence for autotrophic denitrifying communities. The results showed that the nirS gene was strongly labeled by H 13 CO 3 - , demonstrating that the inorganic carbon source was assimilated by autotrophic denitrifiers. High-throughput sequencing and clone library analysis identified Thiobacillus-like bacteria as the most dominant autotrophic denitrifiers. However, 88% of nirS genes cloned from the 13 C-labeled "heavy" DNA fraction showed low similarity with all culturable denitrifiers. These findings provided functional and taxonomical identification of autotrophic denitrifying communities, facilitating application of autotrophic denitrification process for wastewater treatment. Copyright © 2017 Elsevier Ltd. All rights reserved.

  7. Poison 1 - a programme for calculation of reactivity transients due to fission product poisoning and its application in continuous determination of xenon and samarium poisoning in reactor KS-150

    International Nuclear Information System (INIS)

    Rana, S.B.

    1973-12-01

    The report contains a user's description of the 3-dimensional programme POISON 1 for calculating reactivity transients due to fission-product poisoning during changes of reactor power. The chapter dealing with Xe poisoning contains a description of Xe tables, the method of operational determination of Xe poisoning, use of Xe transients for calibrating control rods and means of shutting down the reactor without being overriden by Xe poisoning. Sm poisoning is determined continuously on the basis of the power diagram of reactor operation. In conclusion a possibility of using the programme in a process computer in combination with self-powered detectors as local power sensors is indicated. (author)

  8. User's manual for ASTERIX-2: A two-dimensional modular code system for the steady state and xenon transient analysis of a pebble bed high temperature reactor

    International Nuclear Information System (INIS)

    Wu, T.; Cowan, C.L.; Lauer, A.; Schwiegk, H.J.

    1982-03-01

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analysis from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution. (orig.)

  9. Anticipated transients without scram for WWER reactors. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1999-12-01

    Anticipated transients without scram (ATWS) are anticipated operational occurrences followed by the failure of one reactor scram function. Current international practice requires that the capability of pressurized water reactors (PWRs) to cope with ATWS be demonstrated following a systematic evaluation of plants' defence in depth. Countries operating PWRs require design consideration of ATWS events on a deterministic basis. The regulatory requirements may concern either specific mitigating systems or acceptable plant performance during these events. The prevailing international practice for performing transient analysis of ATWS for licensing is the best estimate approach. Available transient analyses of ATWS events indicate that WWER reactors, like PWRs, have the tendency to shut themselves down if the inherent nuclear feedback is sufficiently negative. Various control and limitation functions of the WWER plants also provide a degree of defence against ATWS. However, for most WWER plants, complete and systematic ATWS analyses have yet to be submitted for rigorous review by the regulatory authorities and preventive or mitigative measures have not been established. In addition, it has also been recognized that plant behaviour in case of ATWS also relies on certain system functions (use of pressurizer safety valves for liquid discharge, availability of steam dump valve to both the condenser (BRU-K) and the atmosphere (BRU-A) for secondary side pressure control, and others) which have been identified as safety issues and need to be qualified for accident conditions. In all countries operating WWERs, the need for ATWS investigations is recognized and reflected in the safety improvement programmes. ATWS analysis for WWERs is not required for the licensing process in Bulgaria, the Czech Republic (with the exception of the Temelin nuclear power plant) and Russia. Design consideration of ATWS is required if expert assessments of probabilistic safety assessment (PSA) results

  10. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-06-01

    A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input

  11. Summary of transient analysis

    International Nuclear Information System (INIS)

    Saha, P.

    1984-01-01

    This chapter reviews the papers on the pressurized water reactor (PWR) and boiling water reactor (BWR) transient analyses given at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Most of the papers were based on the systems calculations performed using the TRAC-PWR, RELAP5 and RETRAN codes. The status of the nuclear industry in the code applications area is discussed. It is concluded that even though comprehensive computer codes are available for plant transient analysis, there is still a need to exercise engineering judgment, simpler tools and even hand calculations to supplement these codes

  12. A model for steady-state and transient determination of subcooled boiling for calculations coupling a thermohydraulic and a neutron physics calculation program for reactor core calculation

    International Nuclear Information System (INIS)

    Mueller, R.G.

    1987-06-01

    Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de

  13. Modeling and analysis of thermal-hydraulic response of uranium-aluminum reactor fuel plates under transient heatup conditions

    Energy Technology Data Exchange (ETDEWEB)

    Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.

  14. Effects of generation and optimization of libraries of effective sections in the analysis of transient in PWR reactors; Efectos de generacion y optimizacion de librerias de secciones eficaces en el analisis de transitorios en reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Cervera, S.; Garcia Herranz, N.; Cuervo, D.; Ahnert, C.

    2014-07-01

    In this paper evaluates the impact that has a certain mesh on a transient in a PWR reactor in the expulsion of a control bar. Have been used for this purpose the coupled codes neutronic and Thermo-hydraulic COBAYA3/COBRA-TF. This objective has been chosen the OECD/NEA PWR MOX/UO{sub 2} rod ejection transient benchmark provides isotopic compositions and defined geometric configurations that allow the use of codes lattice to generate own bookstores. The code used for this transport has been the code APOLLO2.8. The results show large discrepancies when using the benchmark library or libraries own by comparing them to the other participants solutions. The source of these discrepancies is the nodal effective sections provided in the benchmark. (Author)

  15. Bifurcation analysis of the simplified models of boiling water reactor and identification of global stability boundary

    Energy Technology Data Exchange (ETDEWEB)

    Pandey, Vikas; Singh, Suneet, E-mail: suneet.singh@iitb.ac.in

    2017-04-15

    Highlights: • Non-linear stability analysis of nuclear reactor is carried out. • Global and local stability boundaries are drawn in the parameter space. • Globally stable, bi-stable, and unstable regions have been demarcated. • The identification of the regions is verified by numerical simulations. - Abstract: Nonlinear stability study of the neutron coupled thermal hydraulics instability has been carried out by several researchers for boiling water reactors (BWRs). The focus of these studies has been to identify subcritical and supercritical Hopf bifurcations. Supercritical Hopf bifurcation are soft or safe due to the fact that stable limit cycles arise in linearly unstable region; linear and global stability boundaries are same for this bifurcation. It is well known that the subcritical bifurcations can be considered as hard or dangerous due to the fact that unstable limit cycles (nonlinear phenomena) exist in the (linearly) stable region. The linear stability leads to a stable equilibrium in such regions, only for infinitesimally small perturbations. However, finite perturbations lead to instability due to the presence of unstable limit cycles. Therefore, it is evident that the linear stability analysis is not sufficient to understand the exact stability characteristics of BWRs. However, the effect of these bifurcations on the stability boundaries has been rarely discussed. In the present work, the identification of global stability boundary is demonstrated using simplified models. Here, five different models with different thermal hydraulics feedback have been investigated. In comparison to the earlier works, current models also include the impact of adding the rate of change in temperature on void reactivity as well as effect of void reactivity on rate of change of temperature. Using the bifurcation analysis of these models the globally stable region in the parameter space has been identified. The globally stable region has only stable solutions and

  16. Bifurcation analysis of the simplified models of boiling water reactor and identification of global stability boundary

    International Nuclear Information System (INIS)

    Pandey, Vikas; Singh, Suneet

    2017-01-01

    Highlights: • Non-linear stability analysis of nuclear reactor is carried out. • Global and local stability boundaries are drawn in the parameter space. • Globally stable, bi-stable, and unstable regions have been demarcated. • The identification of the regions is verified by numerical simulations. - Abstract: Nonlinear stability study of the neutron coupled thermal hydraulics instability has been carried out by several researchers for boiling water reactors (BWRs). The focus of these studies has been to identify subcritical and supercritical Hopf bifurcations. Supercritical Hopf bifurcation are soft or safe due to the fact that stable limit cycles arise in linearly unstable region; linear and global stability boundaries are same for this bifurcation. It is well known that the subcritical bifurcations can be considered as hard or dangerous due to the fact that unstable limit cycles (nonlinear phenomena) exist in the (linearly) stable region. The linear stability leads to a stable equilibrium in such regions, only for infinitesimally small perturbations. However, finite perturbations lead to instability due to the presence of unstable limit cycles. Therefore, it is evident that the linear stability analysis is not sufficient to understand the exact stability characteristics of BWRs. However, the effect of these bifurcations on the stability boundaries has been rarely discussed. In the present work, the identification of global stability boundary is demonstrated using simplified models. Here, five different models with different thermal hydraulics feedback have been investigated. In comparison to the earlier works, current models also include the impact of adding the rate of change in temperature on void reactivity as well as effect of void reactivity on rate of change of temperature. Using the bifurcation analysis of these models the globally stable region in the parameter space has been identified. The globally stable region has only stable solutions and

  17. 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

  18. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

  19. TORT-TD/ATTICA3D: a coupled neutron transport and thermal hydraulics code system for 3-D transient analysis of gas cooled high temperature reactors

    International Nuclear Information System (INIS)

    Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

    2011-01-01

    Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)

  20. Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors

    International Nuclear Information System (INIS)

    Toti, A.; Vierendeels, J.; Belloni, F.

    2017-01-01

    Highlights: • A system thermal-hydraulic/CFD coupling methodology is proposed for high-fidelity transient flow analyses. • The method is based on domain decomposition and implicit numerical scheme. • A novel interface Quasi-Newton algorithm is implemented to improve stability and convergence rate. • Preliminary validation analyses on the TALL-3D experiment. - Abstract: The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK• CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.

  1. A fast reactor transient analysis methodology for PCs: Volume 3, LTC program manual of the QuickBASIC code

    International Nuclear Information System (INIS)

    Ott, K.O.; Chung, L.

    1992-06-01

    This manual augments the detailed manual of the GW-BASIC version of the LTC code for an application in QuickBASIC. As most of the GW-BASIC coding of this program for ''LMR Transient Calculations'' is compatible with QuickBASIC, this manual pertains primarily to the required changes, such as the handling of input and output. The considerable reduction in computation time achieved by this conversion is demonstrated for two sample problems, using a variety of hardware and execution options. The revised code is listed. Although the severe storage limitations of GW-BASIC no longer apply, the LOF transient path has not been completed in this QuickBASIC code. Its advantages are thus primarily in the much faster running time for TOP and LOHS transients. For the fastest PC hardware (486) and execution option the computation time is reduced by a factor of 124 compared to GW-BASIC on a 386/20

  2. RAZORBACK - A Research Reactor Transient Analysis Code Version 1.0 - Volume 3: Verification and Validation Report.

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G.

    2017-04-01

    This report describes the work and results of the verification and validation (V&V) of the version 1.0 release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This V&V effort was intended to confirm that the code shows good agreement between simulation and actual ACRR operations.

  3. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

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

  4. Reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  5. TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

    International Nuclear Information System (INIS)

    Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

    1988-01-01

    TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

  6. TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

    1988-03-01

    TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

  7. Long-term ammonia removal in a coconut fiber-packed biofilter: analysis of N fractionation and reactor performance under steady-state and transient conditions.

    Science.gov (United States)

    Baquerizo, Guillermo; Maestre, Juan P; Machado, Vinicius C; Gamisans, Xavier; Gabriel, David

    2009-05-01

    A comprehensive study of long-term ammonia removal in a biofilter packed with coconut fiber is presented under both steady-state and transient conditions. Low and high ammonia loads were applied to the reactor by varying the inlet ammonia concentration from 90 to 260 ppm(v) and gas contact times ranging from 20 to 36 s. Gas samples and leachate measurements were periodically analyzed and used for characterizing biofilter performance in terms of removal efficiency (RE) and elimination capacity (EC). Also, N fractions in the leachate were quantified to both identify the experimental rates of nitritation and nitratation and to determine the N leachate distribution. Results showed stratification in the biofilter activity and, thus, most of the NH(3) removal was performed in the lower part of the reactor. An average EC of 0.5 kg N-NH(3)m(-3)d(-1) was obtained for the whole reactor with a maximum local average EC of 1.7 kg N-NH(3)m(-3)d(-1). Leachate analyses showed that a ratio of 1:1 of ammonium and nitrate ions in the leachate was obtained throughout steady-state operation at low ammonia loads with similar values for nitritation and nitratation rates. Low nitratation rates during high ammonia load periods occurred because large amounts of ammonium and nitrite accumulated in the packed bed, thus causing inhibition episodes on nitrite-oxidizing bacteria due to free ammonia accumulation. Mass balances showed that 50% of the ammonia fed to the reactor was oxidized to either nitrite or nitrate and the rest was recovered as ammonium indicating that sorption processes play a fundamental role in the treatment of ammonia by biofiltration.

  8. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain); Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  9. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    International Nuclear Information System (INIS)

    Ernestova, M.; Zamboch, M.; Devrient, B.; Roth, A.; Ehrnsten, U.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ritter, S.; Seifert, H.P.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  10. Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

    International Nuclear Information System (INIS)

    Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.; Hughes, E.D.; Solbrig, C.W.

    1975-11-01

    A mathematical model and a numerical solution scheme for thermal-hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media

  11. Development of a Reduced-Order Three-Dimensional Flow Model for Thermal Mixing and Stratification Simulation during Reactor Transients

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui

    2017-09-03

    Mixing, thermal-stratification, and mass transport phenomena in large pools or enclosures play major roles for the safety of reactor systems. Depending on the fidelity requirement and computational resources, various modeling methods, from the 0-D perfect mixing model to 3-D Computational Fluid Dynamics (CFD) models, are available. Each is associated with its own advantages and shortcomings. It is very desirable to develop an advanced and efficient thermal mixing and stratification modeling capability embedded in a modern system analysis code to improve the accuracy of reactor safety analyses and to reduce modeling uncertainties. An advanced system analysis tool, SAM, is being developed at Argonne National Laboratory for advanced non-LWR reactor safety analysis. While SAM is being developed as a system-level modeling and simulation tool, a reduced-order three-dimensional module is under development to model the multi-dimensional flow and thermal mixing and stratification in large enclosures of reactor systems. This paper provides an overview of the three-dimensional finite element flow model in SAM, including the governing equations, stabilization scheme, and solution methods. Additionally, several verification and validation tests are presented, including lid-driven cavity flow, natural convection inside a cavity, laminar flow in a channel of parallel plates. Based on the comparisons with the analytical solutions and experimental results, it is demonstrated that the developed 3-D fluid model can perform very well for a wide range of flow problems.

  12. Transient thermal hydraulic analysis of the IAEA 10 MW MTR reactor during Loss of Flow Accident to investigate the flow inversion

    International Nuclear Information System (INIS)

    AL-Yahia, Omar S.; Albati, Mohammad A.; Park, Jonghark; Chae, Heetaek; Jo, Daeseong

    2013-01-01

    Highlights: • Transient analyses of a slow and fast LOFA were investigated. • A reactor kinetic and thermal hydraulic coupled model was developed. • Based on force balance, the flow rate during flow inversion was determined. • Flow inversion in a hot channel occurred earlier than in an average channel. • Two temperature peaks were observed during both slow and fast LOFA. - Abstract: Transient analyses of the IAEA 10 MW MTR reactor are investigated during a fast and slow Loss of Flow Accident (LOFA) with a neutron kinetic and thermal hydraulic coupling model. A spatial-dependent thermal hydraulic technique is adopted for analyzing the local thermal hydraulic parameters and hotspot location during a flow inversion. The flow rate through the channel is determined in terms of a balance between driving and preventing forces. Friction and buoyancy forces act as resistance of the flow before a flow inversion while buoyancy force becomes the driving force after a flow inversion. By taking into account the buoyancy effect to determine the flow rate, the difference in the flow inversion time between hot and average channels is investigated: a flow inversion occurs earlier in the hot channel than in an average channel. Furthermore, the movement of the hotspot location before and after a flow inversion is investigated for a slow and fast LOFA. During a flow inversion, two temperature peaks are observed: (1) the first temperature peak is at the initiation of the LOFA, and (2) the second temperature peak is when a flow inversion occurs. The maximum temperature of the cladding is found at the second temperature peak for both LOFA analyses, and is lower than the saturation temperature

  13. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  14. Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chenu, A.

    2011-10-01

    Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

  15. Effects of transient and non-uniform distribution of heat flux on intensity of heat transfer and burnout conditions in the channels of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vitaly Osmachkin [Russian Research Center ' Kurchatov Institute' 1, Kurchatov sq, Moscow 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: The influence of power transient, changes of flow rate, inlet temperatures or pressure in cores of nuclear reactors on heat transfer and burnout conditions in channels depend on rate of such violations. Non-uniform distribution of the heat flux is also important factor for heat transfer and development of crisis phenomenon. Such effects may be significant for NPPs safety. But they have not yet generally accepted interpretation. Steady state approach is often recommended for use in calculations. In the paper a review of experimental observed so-called non-equilibrium effects is presented. The effects of space and time factors are displaying due delay in reformation turbulence intensity, velocity, temperatures or void fraction profiles, water film flow on the surface of heated channels. For estimation of such effect different methods are used. Modern computer codes based on two or three fluids approaches are considered as most effective. But simple and clear correlations may light up the mechanics of effects on heat transfer and improve general understanding of scale and significance of the transient events. In the paper the simplified methods for assessment the influence of lags in the development of distributions of parameters of flow, the relaxation of temporal or space violations are considered. They are compared with more sophisticated approaches. Velocities of disturbance fronts moving along the channels are discussed also. (author)

  16. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  17. Study of sodium boiling during a power transient in a heating duct. Investigation carried out within the scope of a study on the safety of fast neutron reactors

    International Nuclear Information System (INIS)

    Seiler, J.-M.

    1977-01-01

    This work comprises an experimental study and a physical and theoretical interpretation of sodium boiling during a power transient in conditions simulating a power excursion accident in a breeder reactor. The experimental study was performed on an 'out of pile' system with forced sodium convection. It enabled a detailed study to be made of the draining of the heating duct representing a part of a reactor assembly (expansion of the double-phase area on the heating part of the pin). The physical and theoretical interpretation rests on a very simple calculation model solely describing the heat transfers in the duct, assuming that the fluid stays liquid. The extension of the boiling is governed by the initial overheating and the boiling conditioned by thermal inertias of the walls. The extent of the overheating appears to depend on the degree of initial under-saturation. In the absence of overheating, the rate at which the double-phase front moves is controlled by the saturated boiling. This front is well localised when the central temperature gradient is significant. The draining rate is then directly proportional to the growth rate of the temperature in the liquid. The model employed in the CASPAR code for describing power excursion experiments in single liquid phase is described. This code makes it possible to calculate the heat transfers occurring in a forced system by a heating part, a coolant in movement and possibly a part representing an outer cover [fr

  18. Comparison of capability between two versions of reactor transient diagnosis expert system 'DISKET' programmed in different languages

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo

    1991-01-01

    An expert system DISKET has been developed at JAERI to apply knowledge engineering techniques to the transient diagnosis of nuclear power plant. The first version of DISKET programmed in UTILISP has been developed with the main-frame computer FACOM M-780 at JAERI. The LISP language is not suitable for on-line diagnostic systems because it is highly dependent on computer to be used and requires a large computer memory. The large mainframe computer is also not suitable because there are various restrictions as a multi-user computer system. The second version of DISKET for a practical use has been developed in FORTRAN to realize on-line real time diagnoses with limited computer resources. These two versions of DISKET with the same knowledge base have been compared in running capability, and it has been found that the LISP version of DISKET needs more than two times of memory and CPU time of FORTRAN version. From this result, it is shown that this approach is a practical one to develop expert systems for on-line real time diagnosis of transients with limited computer resources. (author)

  19. Experimental study of the attenuation waves oriented to transients caused by the sodium-water explosive reaction in fast reactors

    International Nuclear Information System (INIS)

    Pedroso, L.J.

    1990-01-01

    One of the problems related to fluid-structure interaction that can compromise the structural integrity of components of a fast reactor is the explosion caused by the sodium-water reaction, in the case of a flood at the level of the thermic exchange wall at the steam generator. In this paper we have considered the aspects of the pressure-waves damping caused by the reaction, when these waves transverse certain perforated structures. In order to solve this problem, we also adopted a parametric experimental approach, using a scale model (RIO test rig). (author)

  20. The design of electrical heater pins to simulate transient dryout and post-dryout of water reactor fuel

    International Nuclear Information System (INIS)

    Burgess, M.H.; Butcher, A.A.; Sidoli, J.E.A.

    1978-11-01

    A theoretical assessment of indirect and direct filled heater simulations of nuclear reactor fuel pins is described. For reasons of fast temperature response, a direct unfilled heater, with thermocouples buried in the walls, is recommended for studies of Loss-of-Coolant Accidents leading to dryout, post-dryout and rewetting. A design of heater pins, for use in SGHWR or PWR experiments, and compatible with existing 9MW power supplies, is described. Experiments to confirm collapse pressure calculations at 1000 0 C and thermocouple response times are also reported. (author)

  1. Time-step selection considerations in the analysis of reactor transients with DIF3D-K

    International Nuclear Information System (INIS)

    Taiwo, T.A.; Khalil, H.S.; Cahalan, J.E.; Morris, E.E.

    1993-01-01

    The DIF3D-K code solves the three-dimensional, time-dependent multigroup neutron diffusion equations by using a nodal approach for spatial discretization and either the theta method or one of three space-time factorization approaches for temporal integration of the nodal equations. The three space-time factorization options (namely, improved quasistatic, adiabatic, and conventional point kinetics) were implemented because of their potential efficiency advantage for the analysis of transients in which the flux shape changes more slowly than its amplitude. In this paper, we describe the implementation of DIF3D-K as the neutronics module within the SAS-HWR accident analysis code. We also describe the neuronic-related time-step selection algorithms and their influence on the accuracy and efficiency of the various solution options

  2. On the rate determining step in fission gas release from high burn-up water reactor fuel during power transients

    International Nuclear Information System (INIS)

    Walker, C.T.; Mogensen, M.

    1987-01-01

    The radial distribution of grain boundary gas in a PWR and a BWR fuel is reported. The measurements were made using a new approach involving X-ray fluorescence analysis and electron probe microanalysis. In both fuels the concentration of grain boundary gas was much higher than hitherto suspected. The gas was mainly contained in the bubble/pore structure. The factors that determined the fraction of gas released from the grains and the level of gas retention on the grain boundaries are identified and discussed. The variables involved are the local fuel stoichiometry, the amount of open porosity, the magnitude of the local compressive hydrostatic stress and the interaction of metallic precipitates with gas bubbles on the grain faces. It is concluded that under transient conditions the interlinkage of gas bubbles on the grain faces and the subsequent formation of grain edge tunnels is the rate determining step for gas release; at least when high burn-up fuel is involved. (orig.)

  3. Time-step selection considerations in the analysis of reactor transients with DIF3D-K

    International Nuclear Information System (INIS)

    Taiwo, T.A.; Khalil, H.S.; Cahalan, J.E.; Morris, E.E.

    1993-01-01

    The DIF3D-K code solves the three-dimensional, time-dependent multigroup neutron diffusion equations by using a nodal approach for spatial discretization and either the theta method or one of three space-time factorization approaches for temporal integration of the nodal equations. The three space-time factorization options (namely, improved quasistatic, adiabatic and conventional point kinetics) were implemented because of their potential efficiency advantage for the analysis of transients in which the flux shape changes more slowly than its amplitude. Here we describe the implementation of DIF3D-K as the neutronics module within the SAS-HWR accident analysis code. We also describe the neutronics-related time step selection algorithms and their influence on the accuracy and efficiency of the various solution options

  4. A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

    International Nuclear Information System (INIS)

    Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho; Pyeon, Cheol Ho

    2015-01-01

    In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r g , E g , t g ) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the neutron sources

  5. A Derivation of Source-based Kinetics Equation with Time Dependent Fission Kernel for Reactor Transient Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Song Hyun; Woo, Myeong Hyun; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Pyeon, Cheol Ho [Kyoto University, Osaka (Japan)

    2015-10-15

    In this study, a new balance equation to overcome the problems generated by the previous methods is proposed using source-based balance equation. And then, a simple problem is analyzed with the proposed method. In this study, a source-based balance equation with the time dependent fission kernel was derived to simplify the kinetics equation. To analyze the partial variations of reactor characteristics, two representative methods were introduced in previous studies; (1) quasi-statics method and (2) multipoint technique. The main idea of quasistatics method is to use a low-order approximation for large integration times. To realize the quasi-statics method, first, time dependent flux is separated into the shape and amplitude functions, and shape function is calculated. It is noted that the method has a good accuracy; however, it can be expensive as a calculation cost aspect because the shape function should be fully recalculated to obtain accurate results. To improve the calculation efficiency, multipoint method was proposed. The multipoint method is based on the classic kinetics equation with using Green's function to analyze the flight probability from region r' to r. Those previous methods have been used to analyze the reactor kinetics analysis; however, the previous methods can have some limitations. First, three group variables (r{sub g}, E{sub g}, t{sub g}) should be considered to solve the time dependent balance equation. This leads a big limitation to apply large system problem with good accuracy. Second, the energy group neutrons should be used to analyze reactor kinetics problems. In time dependent problem, neutron energy distribution can be changed at different time. It can affect the change of the group cross section; therefore, it can lead the accuracy problem. Third, the neutrons in a space-time region continually affect the other space-time regions; however, it is not properly considered in the previous method. Using birth history of the

  6. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  7. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  8. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L.

    1996-02-01

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components.

  9. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-01-01

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to: (1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, (2) assess the RELAP5 and TRACE computer code against the experimental data, and (3) develop mathematical model and heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

  10. Development and application of a Fourier transform based methodology for the identification of instability in boiling water reactors at a local scale

    Energy Technology Data Exchange (ETDEWEB)

    Walser, Stefan Franz

    2017-04-06

    This thesis addresses the development of an analysis methodology for BWR instability phenomena and aims at the identification of in-core, local thermal-hydraulic processes during a transient. The analysis methodology is designed to apply as input data the simulation results of time domain coupled system codes. For the application described in this thesis, a coupled TRACE/PARCS model representing the Oskarshamn-2 (O2) NPP with a one-to-one core channel representation has been used. The coupled model simulates the O2-1999 feedwater transient; an instability event characterized by an in-phase mode of oscillation with reactor power amplitudes up to 132 %. The analysis methodology is a two-step approach and uses in the first step the fast-fourier transform algorithm applied on normalized core parameters in twodimensional spacial direction of the core. The normalization of the data implies the advantage of directly comparable results in spectral representation. The spectral analysis results show for each data node the oscillation amplitude to its corresponding frequency. In the second step the dominating frequency of each single parameter is determined and the relative phase shift of the dominating components is calculated. The application of the developed methodology on the simulation results of the O2-1999 feedwater transient show that the channel mass flow rates have among all investigated parameters the clearest differences in the local expression of oscillation and are a governing indicator for BWR instability due to the density wave mechanism. The spectral analysis of the core channel in planar direction points out a heterogeneous oscillation behavior of the fuel assemblies mass fl ow rates. A certain pattern of core channels with striking mass flow rate oscillations is prevailing and the pattern shows proportionality to the fuel assemblies relative power ratio. Moreover the mass ow rate oscillations of the peripheral core channels are observed to employ a quasi

  11. Identification of reactor failure states using noise methods, and spatial power distribution

    International Nuclear Information System (INIS)

    Vavrin, J.; Blazek, J.

    1981-01-01

    A survey is given of the results achieved. Methodical means and programs were developed for the control computer which may be used in noise diagnostics and in the control of reactor power distribution. Statistical methods of processing the noise components of the signals of measured variables were used for identifying failures of reactors. The method of the synthesis of the neutron flux was used for modelling and evaluating the reactor power distribution. For monitoring and controlling the power distribution a mathematical model of the reactor was constructed suitable for control computers. The uses of noise analysis methods are recommended and directions of further development shown. (J.P.)

  12. Vibration system identification of Paks and Kozloduy reactor buildings on the basis of the blast test results

    International Nuclear Information System (INIS)

    Varpasuo, P.

    1999-01-01

    System identification allows to build mathematical models of a dynamic system based on measured data. System identification is carried out by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The aim of this study is to investigate and model the behavior of complex vibratory systems on the basis of measured excitation and response. The first part of the study describes the theory used in the analysis and the software tools used in the analysis. The second part of the study describes the investigation and modeling of the response of single degree of freedom oscillator excited by sinusoidal and blast excitation. In the third part of the study the system identification of the Kozloduy NPP unit 5 reactor building and Paks NPP unit 1 reactor building is studied and the models are estimated using the method of segmentation of excitation and response. System identification is carried out using MATLAB software by adjusting parameters within a given model until its output coincides as well as possible with the measured output. The types of models used for the were: l) ARX models; 2) ARMAX model; 3) Output-Error (OE) models; 4) Box-Jenkins (BJ) models; 5) State-space models. The model coefficients for different models were calculated using the least-squares and maximum likelihood estimation methods available in MATLAB system identification toolbox. Excitation was in both Paks and Kozloduy case the measured free-field excitation and responses were the vibration responses of the building on the foundation slab level and top of the building. By examining the established models the frequency characteristics of vibration systems were determined with 95 % accuracy and the amplitude response with 80 % accuracy. In case of the steady state response of sinusoidally excited single dof oscillator the modelling gave almost exact results. But in the case of the blast response of the reactor building the obtaining of the

  13. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

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

  14. A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94

    International Nuclear Information System (INIS)

    Orr, B. R.

    1999-01-01

    Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA)

  15. On the application of the method of turbulent flow to transient and quasi-stationary flow calculations in high temperature reactors with spherical fuel elements

    International Nuclear Information System (INIS)

    Hoefer, I.

    1980-12-01

    For the calculation of flows in high-temperature reactors and of their temperature behavior the equations of the method of turbulent flow in the primitive form are derived for inhomogeneous regions. This system of equations is appropriate for the investigation of transient and quasi-stationary phenomena in pebble beds. By modification of the flow function in parallel arranged reflector channels a parallel flow can be simulated. For simplification the flow in region with a smaller pressure loss is assumed to be a potential flow. For the numerical solution of the time-dependent convective parts of the system of equations a number of explicit and implicit difference methods are compared. If the method using UP-WIND differences is taken to be an interpolation method the introduction of an extension becomes possible, which together with preliminary integration of the fictional terms allows to apply larger time steps. The algebraic system of equations for numerical calculation of a steady flow field also is established by formation of UP-WIND differences for the convective terms. By mathematical verification of some examples the applicability of the mathematical model for flow problems in pebble beds with forced or natural convection is shown. (orig.) [de

  16. Using reactor network for global identification based on residence time distribution theory

    Energy Technology Data Exchange (ETDEWEB)

    Hocine, S.; Pibouleau, L.; Azzaro-Pantel, C.; Domenech, S. [Laboratoire de Genie Chimique - UMR 5503 CNRS/ INPT ENSIACET, 31 - Toulouse (France)

    2006-07-01

    In the ventilation systems, the control of transfer contaminants is one of the principal problems during the design and control phases. The installation of a suitable ventilation system for the control of contaminant transfer is essential in industry, because it makes it possible to detect and to prevent chemical and radiological risks. Research on air distribution in ventilated rooms traditionally involves full-scale experiments, scale -model experiments and application of the computational fluid dynamics (C.F.D.) tools. Most of the time, particularly in our case of large and cluttered enclosures, the predictive approach based on C.F.D. codes can not be used. The solution retained here is the establishment of a model based on the well known residence time distribution. This model is widely used in chemical engineering to treat non-ideal flows. The proposed method is based on the experimental determination of the residence time distribution curve, generally obtained through the response of the system to tracer release. A superstructure involving the set of all the possible solutions corresponding to the physical reactor is then defined, and the model will be selected from this superstructure according to its simulated response. The superstructure is identified as a combination of elementary systems, representing ideal flow patterns, as perfect mixed flows, plug flows, continuous stirred tank reactors, etc. The selected model is derived from the comparison between the simulated response to a stimulus, and the experimental response. The structure and parameters of the model are simultaneously optimized in order to fit the experimental curve with a minimal number of elementary units, constituting a key point for future control purposes of the process. This problem is a dynamic M.I.N.L.P. (Mixed Integer Non Linear Programming) problem with bilinear equality constraints. Generally, these constraints lead to numerical difficulties for reaching an optimum solution (even a

  17. The Hanford Site N Reactor buildings task identification and evaluation of historic properties

    International Nuclear Information System (INIS)

    Stapp, D.C.; Marceau, T.E.

    1996-01-01

    The New Production Reactor complex at Hanford (hereafter referred to as N Reactor) is proposed to be deactivated, decommissioned, and demolished in the coming years. Recognizing that the Hanford Site has been important to the nation, state, and local community, a task was funded to examine the effects that these activities may have on the historic properties of N Reactor. The objectives of the N Reactor buildings task were to identify potential historic properties at N Reactor, to complete Historic Property Inventory forms for all structures considered eligible and ineligible for listing in the National Register of Historic Places, and to prepare a Memorandum of Agreement that identifies the measures required to mitigate any adverse effects

  18. 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.

  19. Identification and management of plant aging and life extension issues for a liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    King, R.W.; Perry, W.H.

    1991-06-01

    Experimental Breeder Reactor 2 (EBR-2) is a pool-type sodium-cooled fast reactor that supports extensive experimental, test and demonstration programs while providing electrical power to the local grid. EBR-2 is a US Department of Energy Facility located at the Idaho National Engineering Laboratory and operated by Argonne National Laboratory (ANL). The current mission of EBR-2 is to serve as the operational prototype for the Integral Fast Reactor demonstration program. This mission and other programs require EBR-2 to operate reliability to a 40-year lifetime, a significant extension beyond the five to ten year life originally planned for the facility. The benefits of operating EBR-2 in the extended-life mode are important for providing long-term operational performance data for a sodium-cooled fast reactor that is not available elsewhere. Identification and preliminary assessment of potential life-limiting factors indicate that, with appropriate consideration given in the design phase, the sodium-cooled plant has potential for a very long operational lifetime. Achievement of a 40-year lifetime with high reliability is important not only for achieving the near-term goals of the EBR-2/IFR programs, but for the advancement of the liquid-metal-cooled reactor concept to the demonstration/commercialization phase. Key features make extended-life operation feasible based on the use of sodium as the primary coolant: low-pressure, high thermal capacity primary system and a low-pressure secondary system requiring no active valves; and limited corrosion of components. 2 refs

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

    International Nuclear Information System (INIS)

    El-Kafas, A.A.

    1997-01-01

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

  1. Numerical simulation and analysis of axial instabilities occurrence and development in turbomachines. Application to a break transient in a helium nuclear reactor

    International Nuclear Information System (INIS)

    Tauveron, N.

    2006-02-01

    The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge

  2. Steady- and transient-state analyses of fully ceramic microencapsulated fuel loaded reactor core via two-temperature homogenized thermal-conductivity model

    International Nuclear Information System (INIS)

    Lee, Yoonhee; Cho, Nam Zin

    2015-01-01

    Highlights: • Fully ceramic microencapsulated fuel-loaded core is analyzed via a two-temperature homogenized thermal-conductivity model. • The model is compared to harmonic- and volumetric-average thermal conductivity models. • The three thermal analysis models show ∼100 pcm differences in the k eff eigenvalue. • The three thermal analysis models show more than 70 K differences in the maximum temperature. • There occur more than 3 times differences in the maximum power for a control rod ejection accident. - Abstract: Fully ceramic microencapsulated (FCM) fuel, a type of accident-tolerant fuel (ATF), consists of TRISO particles randomly dispersed in a SiC matrix. In this study, for a thermal analysis of the FCM fuel with such a high heterogeneity, a two-temperature homogenized thermal-conductivity model was applied by the authors. This model provides separate temperatures for the fuel-kernels and the SiC matrix. It also provides more realistic temperature profiles than those of harmonic- and volumetric-average thermal conductivity models, which are used for thermal analysis of a fuel element in VHTRs having a composition similar to the FCM fuel, because such models are unable to provide the fuel-kernel and graphite matrix temperatures separately. In this study, coupled with a neutron diffusion model, a FCM fuel-loaded reactor core is analyzed via a two-temperature homogenized thermal-conductivity model at steady- and transient-states. The results are compared to those from harmonic- and volumetric-average thermal conductivity models, i.e., we compare k eff eigenvalues, power distributions, and temperature profiles in the hottest single-channel at steady-state. At transient-state, we compare total powers, reactivity, and maximum temperatures in the hottest single-channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized thermal

  3. Proceedings of the 8. Brazilian Meeting on Reactor Physics and Thermal Hydraulics

    International Nuclear Information System (INIS)

    1991-01-01

    Some papers about pressurized light water reactors, fast reactors, accident analysis, transients, research reactors, nuclear data collection, thermal hydraulics, reactor monitoring, neutronics are presented. (E.G.)

  4. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

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

  5. Identification of significant process variables for a flow-through supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Rossi, R.E.

    1992-05-01

    The effects of four process variables on the destruction efficiency of a flow-through supercritical water oxidation reactor were investigated. These process variables included: (1) reactor throughput (GPH), (2) concentration of the surrogate waste (% acetone), (3) maximum reactor tube-wall temperature (OC), and (4) applied stoichiometric oxygen. The analysis was conducted utilizing two-level factorial experiments, steepest ascent methods, and central composite designs. This experimental protocol assures efficient experimentation and allows for an empirical response surface model of the system to be developed. This experimentation identified a significant positive effect for stoichiometric oxygen applied and temperature variations between 400 to 500 degrees C. The increase in destruction efficiency due to stoichiometric 0 2 provides strong evidence that supercritical water oxidations are catalyzed by excess oxygen, and the strong temperature effect is a result of large increases in the kinetic rates for this temperature range. However, increasing temperature between 550 to 650 degrees C does not provide substantial increases in destruction efficiency. In addition, destruction efficiency is significantly unproved by increasing the Reynolds number and residence time. The destruction efficiency of the reactor is also dependent upon the initial concentration of surrogate waste. This concentration dependence may indicate first-order supercritical CO kinetics is inadequate for describing all waste types and reactor configurations. Alternatively, it may indicate reactant mixing, caused by local turbulence at the oxidation fronts of these higher concentration waste streams, results in higher destruction efficiencies

  6. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [ORNL; Mays, Gary T [ORNL; Omitaomu, Olufemi A [ORNL; Poore III, Willis P [ORNL

    2013-12-01

    This analysis identifies candidate locations, in a broad sense, where there are high concentrations of federal government agency use of electricity, which are also suitable areas for near-term SMRs. Near-term SMRs are based on light-water reactor (LWR) technology with compact design features that are expected to offer a host of safety, siting, construction, and economic benefits. These smaller plants are ideally suited for small electric grids and for locations that cannot support large reactors, thus providing utilities or governement entities with the flexibility to scale power production as demand changes by adding additional power by deploying more modules or reactors in phases. This research project is aimed at providing methodologies, information, and insights to assist the federal government in meeting federal clean energy goals.

  7. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    International Nuclear Information System (INIS)

    Charak, I.; Kier, P.H.

    1995-04-01

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

  8. An approach using quantum ant colony optimization applied to the problem of identification of nuclear power plant transients

    International Nuclear Information System (INIS)

    Silva, Marcio H.; Schirru, Roberto; Medeiros, Jose A.C.C.

    2009-01-01

    Using concepts and principles of the quantum computation, as the quantum bit and superposition of states, coupled with the biological metaphor of a colony of ants, used in the Ant Colony Optimization algorithm (ACO), Wang et al developed the Quantum Ant Colony Optimization (QACO). In this paper we present a modification of the algorithm proposed by Wang et al. While the original QACO was used just for simple benchmarks functions with, at the most, two dimensions, QACO A lfa was developed for application where the original QACO, due to its tendency to converge prematurely, does not obtain good results, as in complex multidimensional functions. Furthermore, to evaluate its behavior, both algorithms are applied to the real problem of identification of accidents in PWR nuclear power plants. (author)

  9. Identification and assay of radionuclides in the environmental samples following Chernobyl reactor accident

    International Nuclear Information System (INIS)

    Prakash, S.

    1987-01-01

    In India radioactivity was detected from 2 May 1986 onwards. A variety of samples were assayed for radionuclides in the environmental samples originating from the Chernobyl reactor accident. These are: Cotton swipe samples from aircrafts, air filters, milk, goat thyroids, grass, vegetables and tap waters. The results are presented in the paper

  10. Improved identification to prevent transposition during operation of 900 MWe PWR reactors

    International Nuclear Information System (INIS)

    Leckner, J.M.; Dien, Y.; Cernes, A.

    1986-04-01

    Detailed human factors analysis of 900 MWe PWR control room identification systems was carried out by the Nuclear and Fossil Generation Division of Electricite de France (EDF) consequent to a series of incidents where personnel confused one plant unit, room or piece of equipment for another. Preliminary analysis uncovered coding inadequacies and suggested possible remedies. This data was used to prepare specifications for identification redesign at a pilot plant on which detailed investigations could be carried out. Recommended solutions were submitted to pilot plant operators and their opinion sollicited. Operator recommendations will be tried out on the pilot plant and adopted on a grid-wide basis if trials prove satisfactory

  11. Learning from anticipated and abnormal plant transients

    International Nuclear Information System (INIS)

    Varnado, B.

    1983-01-01

    A report is given of the American Nuclear Society topical meeting on Anticipated and Abnormal Transients in Light Water Reactors held in Jackson, Wyoming in September 1983. Industry involvement in the evaluation of operating experience, human error contributions, transient management, thermal hydraulic modelling, the role of probabilistic risk assessment and the cost of transient incidents are discussed. (U.K.)

  12. 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

  13. Using a combination of weighting factor method and imperialist competitive algorithm to improve speed and enhance process of reloading pattern optimization of VVER-1000 reactors in transient cycles

    Energy Technology Data Exchange (ETDEWEB)

    Rahmani, Yashar, E-mail: yashar.rahmani@gmail.com [Department of Physics, Faculty of Engineering, Islamic Azad University, Sari Branch, Sari (Iran, Islamic Republic of); Shahvari, Yaser [Department of Computer Engineering, Payame Noor University (PNU), P.O. Box 19395-3697, Tehran (Iran, Islamic Republic of); Kia, Faezeh [Golestan Institute of Higher Education, Gorgan 49139-83635 (Iran, Islamic Republic of)

    2017-03-15

    Highlights: • This article was an attempt to optimize reloading pattern of Bushehr VVER-1000 reactor. • A combination of weighting factor method and the imperialist competitive algorithm was used. • The speed of optimization and desirability of the proposed pattern increased considerably. • To evaluate arrangements, a coupling of WIMSD5-B, CITATION-LDI2 and WERL codes was used. • Results reflected the considerable superiority of the proposed method over direct optimization. - Abstract: In this research, an innovative solution is described which can be used with a combination of the new imperialist competitive algorithm and the weighting factor method to improve speed and increase globality of search in reloading pattern optimization of VVER-1000 reactors in transient cycles and even obtain more desirable results than conventional direct method. In this regard, to reduce the scope of the assumed searchable arrangements, first using the weighting factor method and based on values of these coefficients in each of the 16 types of loadable fuel assemblies in the second cycle, the fuel assemblies were classified in more limited groups. In consequence, the types of fuel assemblies were reduced from 16 to 6 and consequently the number of possible arrangements was reduced considerably. Afterwards, in the first phase of optimization the imperialist competitive algorithm was used to propose an optimum reloading pattern with 6 groups. In the second phase, the algorithm was reused for finding desirable placement of the subset assemblies of each group in the optimum arrangement obtained from the previous phase, and thus the retransformation of the optimum arrangement takes place from the virtual 6-group mode to the real mode with 16 fuel types. In this research, the optimization process was conducted in two states. In the first state, it was tried to obtain an arrangement with the maximum effective multiplication factor and the smallest maximum power peaking factor. In

  14. Hierarchically structured identification and classification method for vibrational monitoring of reactor components

    International Nuclear Information System (INIS)

    Saedtler, E.

    1981-01-01

    The dissertation discusses: 1. Approximative filter algorithms for identification of systems and hierarchical structures. 2. Adaptive statistical pattern recognition and classification. 3. Parameter selection, extraction, and modelling for an automatic control system. 4. Design of a decision tree and an adaptive diagnostic system. (orig./RW) [de

  15. User's manual for ASTERIX-2: a two-dimensional modular-code system for the steady-state and xenon-transient analysis of a pebble-bed high-temperature reactor

    International Nuclear Information System (INIS)

    Lauer, A.; Schwiegk, H.J.; Wu, T.; Cowan, C.L.

    1982-03-01

    The ASTERIX modular code package was developed at KFA Laboratory-Juelich for the steady state and xenon transient analysis of a pebble bed high temperature reactor. The code package was implemented on the Stanford Linear Accelerator Center Computer in August, 1980, and a user's manual for the current version of the code, identified as ASTERIX-2, was prepared as a cooperative effort by KFA Laboratory and GE-ARSD. The material in the manual includes the requirements for accessing the program, a description of the major subroutines, a listing of the input options, and a listing of the input data for a sample problem. The material is provided in sufficient detail for the user to carry out a wide range of analyses from steady state operations to the xenon induced power transients in which the local xenon, temperature, buckling and control feedback effects have been incorporated in the problem solution

  16. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor

    International Nuclear Information System (INIS)

    Victoria R, M.A.; Morales S, J.B.

    2005-01-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  17. Experience with automatic reactor control at EBR-II

    International Nuclear Information System (INIS)

    Lehto, W.K.; Larson, H.A.; Christensen, L.J.

    1985-01-01

    Satisfactory operation of the ACRDS has extended the capabilities of EBR-II to a transient test facility, achieving automatic transient control. Test assemblies can now be irradiated in transient conditions overlapping the slower transient capability of the TREAT reactor

  18. Phenomena identification ranking table and knowledge base gaps and needs for the modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Tokuhiro, Akira; Potirniche, Gabriel; Rink, Karl

    2009-01-01

    The U.S. is developing a modular high-temperature gas-cooled reactor (MHTGR) under the Next Generation Nuclear Plant (NGNP); also known as the Very High Temperature Reactor (VHTR). The generic MHTGR is a graphite-moderated, gas-cooled reactor (GCR) of either a prismatic modular (block-type, PMR) or pebble-bed (PBR) core configuration. The pebble-bed design requires new attention with respect to neutronics, materials, thermal hydraulic, safety and licensing relative to the set of phenomena and engineering analyses associated with the current fleet of legacy LWRs. In fact, the relative knowledge and experiential base on gas reactors is small in comparison to the LWR. There is a dated body of knowledge from some 25+ years ago on GCRs; recently there is a renewed interest. Thus in the present design and development phase of the NGNP/VHTR, there are relevant thermohydraulic safety issues surrounding the MHTGR with issues impacting foremost the design review process. A common phenomena with respect to PMR and PBR core design, is that concerning 'graphite dust' and its interaction and transport with potential fission products (FP) that may be present within the graphite and subsequently in the primary system. The nature of the graphite and FPs, when circulated or transported in the primary, and possibly beyond, is of concern as potentially an relevant 'source term' (radionuclide inventory) of the MHTGR. Based on NUREG/CR-6944, Volumes 1-5, the author briefly describes the state-of-the art knowledge base on graphite dust and FP transport with respect to the anticipated design of the MHTGR. In addition, from the Phenomena Identification and Ranking Tables (PIRTs) developed in these reports we concurrently identify and describe 'gaps and needs' of the knowledge base. That is, we also present the knowledge base gaps and needs with respect to the following: 1) R and D needs relative to PIRTs, 2) (experimental) database needs relative to PIRTs, and 3) simulation and modeling

  19. Methodology for Identification of the Coolant Thermalhydraulic Regimes in the Core of Nuclear Reactors

    International Nuclear Information System (INIS)

    Sharaevsky, L.G.; Sharaevskaya, E.I.; Domashev, E.D.; Arkhypov, A.P.; Kolochko, V.N.

    2002-01-01

    The paper deals with one of the acute for the nuclear energy problem of accident regimes of NPPs recognition diagnostics using noise signal diagnostics methodology. The methodology intends transformation of the random noise signals of the main technological parameters at the exit of a nuclear facility (neutron flow, dynamic pressure etc.) which contain the important information about the technical status of the equipment. The effective algorithms for identification of random processes wore developed. After proper transformation its were considered as multidimensional random vectors. Automatic classification of these vectors in the developed algorithms is realized on the basis of the probability function in particular Bayes classifier and decision functions. Till now there no mathematical models for thermalhydraulic regimes of fuel assemblies recognition on the acoustic and neutron noises parameters in the core of nuclear facilities. The two mathematical models for analysis of the random processes submitted to the automatic classification is proposed, i.e. statistical (using Bayes classifier of acoustic spectral density diagnosis signals) and geometrical (on the basis of formation in the featured space of dividing hyper-plane). The theoretical basis of the bubble boiling regimes in the fuel assemblies is formulated as identification of these regimes on the basis of random parameters of auto spectral density of acoustic noise (ASD) measured in the fuel assemblies (dynamic pressure in the upper plenum in the paper). The elaborated algorithms allow recognize realistic status of the fuel assemblies. For verification of the proposed mathematical models the analysis of experimental measurements was carried out. The research of the boiling onset and definition of the local values of the flow parameters in the seven-beam fuel assembly (length of 1.3 m, diameter of 6 mm) have shown the correct identification of the bubble boiling regimes. The experimental measurements on

  20. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

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

  1. Identification and adoptive multivariable control method. Application to fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Dang Van Mien, H.; Irving, E.; Rousseau, I.

    1982-01-01

    Motivated by the limitations inherent in the standard approach, a new model reference multivariable adaptive control method is described. This latter control method utilizes as a design tool a simple vector difference equation of the controlled system. The adaptive control method is of the series-parallel direct reference model type and the adjustment algorithm is the standard least squares estimation technics with hyperstability conditions, controlled convergence speed forgetting, regularization and threshold operations. Numerical results are presented which illustrate the interest of the latter approach. The precise problem which has been tackled is the control of the steam generator of the second fast breeder French nuclear reactor called Super-Phenix. After a short description of the plant and its responses at different loads, the principles and the performances of the standard technique control scheme which has been adopted are discussed [fr

  2. Pulse-based electron spin transient nutation measurement of BaTiO3 fine particle: Identification of controversial signal around g = 2.00

    Science.gov (United States)

    Sawai, Takatoshi; Yamaguchi, Yoji; Kitamura, Noriko; Date, Tomotsugu; Konishi, Shinya; Taga, Kazuya; Tanaka, Katsuhisa

    2018-05-01

    Two dimensional pulse-based electron spin transient nutation (2D-ESTN) spectroscopy is a powerful tool for determining the spin quantum number and has been applied to BaTiO3 fine powder in order to identify the origin of the continuous wave electron spin resonance (CW-ESR) signal around g = 2.00. The signal is frequently observed in BaTiO3 ceramics, and the correlation between the signal intensity and positive temperature coefficient of resistivity (PTCR) properties has been reported to date. The CW-ESR spectrum of BaTiO3 fine particles synthesized by the sol-gel method shows a typical asymmetric signal at g = 2.004. The 2D-ESTN measurements of the sample clearly reveal that the signal belongs to the S = 5/2 high spin state, indicating that the signal is not due to a point defect as suggested by a number of researchers but rather to a transition metal ion. Our elemental analysis, as well as previous studies, indicates that the origin of the g = 2.004 signal is due to the presence of an Fe3+ impurity. The D value (second-order fine structure parameter) reveals that the origin of the signal is an Fe3+ center with distant charge compensation. In addition, we show a peculiar temperature dependence of the CW-ESR spectrum, suggesting that the phase transition behavior of a BaTiO3 fine particle is quite different from that of a bulk single crystal. Our identification does not contradict a vacancy-mediated mechanism for PTCR. However, it is incorrect to use the signal at g = 2.00 as evidence to support the vacancy-mediated mechanism.

  3. Development of PARA-ID Code to Simulate Inelastic Constitutive Equations and Their Parameter Identifications for the Next Generation Reactor Designs

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, J. H.

    2006-03-01

    The establishment of the inelastic analysis technology is essential issue for a development of the next generation reactors subjected to elevated temperature operations. In this report, the peer investigation of constitutive equations in points of a ratcheting and creep-fatigue analysis is carried out and the methods extracting the constitutive parameters from experimental data are established. To perform simulations for each constitutive model, the PARA-ID (PARAmeter-IDentification) computer program is developed. By using this code, various simulations related with the parameter identification of the constitutive models are carried out

  4. Identification and evaluation of facilitation techniques for decommissioning light water power reactors

    International Nuclear Information System (INIS)

    LaGuardia, T.S.; Risley, J.F.

    1986-06-01

    This report describes a study sponsored by the US Nuclear Regulatory Commission to identify practical techniques to facilitate the decommissioning of nuclear power generating facilities. The objective of these ''facilitation techniques'' is to reduce the radioactive exposures and/or volumes of waste generated during the decommissioning process. The report presents the possible facilitation techniques identified during the study and discusses the corresponding facilitation of the decommissioning process. Techniques are categorized by their applicability of being implemented during the three stages of power reactor life: design/construction, operation, or decommissioning. Detailed cost-benefit analyses were performed for each technique to determine the anticipated exposure and/or radioactive waste reduction; the estimated costs for implementing each technique were then calculated. Finally, these techniques were ranked by their effectiveness in facilitating the decommissioning process. This study is a part of the Nuclear Regulatory Commission's evaluation of decommissioning policy and its modification of regulations pertaining to the decommissioning process. The findings can be used by the utilities in the planning and establishment of activities to ensure that all objectives of decommissioning will be achieved

  5. A model for the description of the coolant mixing and its application to the analysis of boron dilution transients in pressurized water reactors; Ein Modell zur Beschreibung der Kuehlmittelvermischung und seine Anwendung auf die Analyse von Borverduennungstransienten in Druckwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren

    2010-08-15

    A model for the realistic description of the coolant mixing inside the pressure vessel of a pressurized water reactors has been developed and validated. This fast running model is based on the linear superposition of response functions on Dirac-pulse-like perturbations of the coolant parameters. It has been implemented into the coupled code system DYN3D/ATHLET nad serves as the interface between the one-dimensional thermal hydraulic system code ATHLETE and the 3D neutron kinetic core model DYN3D. By help of this model, the coolant mixing inside the reactor pressure vessel can e simulated in an efficient manner. A methodology for this analysis of hypothetical boron dilution accidents has been developed, which is based on the newly developed model for the coolant mixing. This methodology consists of a combination of stationary and transient calculations including a realistic treatment of the mixing of deborated slugs on the way towards the reactor core. The degree of conservatism can be adjusted by the variation of the initial size of the deborated slug. This new method was applied to two different boron dilution accidents. Besides the start of the first main coolant pump with a deborated slug of coolant in the cold leg of the primary circuit, a deboration event during the operation of the residual heat removal system was investigated. The results of the parameter study for a reactor core with a generic loading pattern demonstrated in both cases, that although the shut-down reactor becomes re-critical safety relevant cladding temperature limits are not reached, even if maximum possible volumes of the deborated slug are considered. The main reason for these results is the use of realistic time-dependent distributions of the boron concentration at the inlet of each fuel assembly.

  6. Identification of two-phase flow patterns in a nuclear reactor by the high-frequency contribution fraction

    International Nuclear Information System (INIS)

    Wang, Y.W.; Pei, B.S.; King, C.H.; Lee, S.C.

    1989-01-01

    Recently, King et al. and Wang et al. analyzed the fluctuating characteristics of differential pressure and void fraction by the optimum modeling method and by spectral analysis, respectively. These two investigations presented some new concepts and deterministic criteria, which are based on purely empirical formulas, to identify two-phase flow patterns. These deterministic criteria on two-phase flow patterns' identification seem to show reasonable performance. In King's and Wang's studies, there are at least three problems that need further investigations for the applications to the nuclear reactor engineering field. These three problems are the following: 1. Is the response to a certain two-phase flow pattern, i.e., the fluctuating characteristics, of neutrons the same as that of differential pressure or void fraction? 2. Could those criteria developed from air/water flow be allowed to identify steam/water two-phase flow patterns? 3. Could those criteria be applied to identify two-phase flow patterns in rod bundles? In this paper, parts of the investigated results answer the first problem, and detailed comparisons with the previous work of the authors are given on a variety of items

  7. Simulation of steady states of an integral PWR and power change transients using RELAP5 MOD3

    International Nuclear Information System (INIS)

    Aronne, Ivan Dionysio Aronne; Palmieri, Elcio Tadeu; Azwvedo, Carlos Vicente Goulart de; Baptista Filho, Benedito Dias; Barroso, Antonio Carlos de Oliveira

    2005-01-01

    An integral pressurized water reactor presents several differences in relation to conventional PWRs. The metal and cooling fluid masses of integral reactors are larger than those of a conventional reactor and, on the other hand, bombs tend to be smaller and the pressurizer should present characteristics proper of that arrangement. These characteristics, representing inertias different from the usual ones, makes obtaining the stationary state of the integral reactor a task with particularities that demand strategies different from the usually employed. This paper presents, initially, the main physical characteristics of the reactor in study and then the options adopted in developing the model and that were used to obtain the simulation of stationary states with the code RELAP5-MOD3. The results of the simulation of the steady state show the effects of the fore mentioned differences, where the times lags are significantly larger, as well as the suitability and efficiency of the defined approach. Two transients were simulated for changing the reactor power from steady state power of 100% to steady state power of 90%. The power change of these transients were one in step and the other in ramp with a rate of 5%/min. These calculations represent a first step for the definition and tests of parts of a preliminary control system for this reactor. The two transient simulated were based on plausible control hypotheses whose results are presented and commented. The final objective of this study is the use of results of simulations of steady states as much as of transients in support to the development of a transient identification and classification system, based on a neural network using self organizing maps whose basic proposition is presented in this paper. (author)

  8. Transient analyzer

    International Nuclear Information System (INIS)

    Muir, M.D.

    1975-01-01

    The design and design philosophy of a high performance, extremely versatile transient analyzer is described. This sub-system was designed to be controlled through the data acquisition computer system which allows hands off operation. Thus it may be placed on the experiment side of the high voltage safety break between the experimental device and the control room. This analyzer provides control features which are extremely useful for data acquisition from PPPL diagnostics. These include dynamic sample rate changing, which may be intermixed with multiple post trigger operations with variable length blocks using normal, peak to peak or integrate modes. Included in the discussion are general remarks on the advantages of adding intelligence to transient analyzers, a detailed description of the characteristics of the PPPL transient analyzer, a description of the hardware, firmware, control language and operation of the PPPL transient analyzer, and general remarks on future trends in this type of instrumentation both at PPPL and in general

  9. HECTR [Hydrogen Event Containment Transient Response] Version 1.5N: A modification of HECTR Version 1.5 for application to N Reactor

    International Nuclear Information System (INIS)

    Camp, A.L.; Dingman, S.E.

    1987-05-01

    This report describes HECTR Version 1.5N, which is a special version of HECTR developed specifically for application to the N Reactor. HECTR is a fast-running, lumped-parameter containment analysis computer program that is most useful for performing parametric studies. The main purpose of HECTR is to analyze nuclear reactor accidents involving the transport and combustion of hydrogen, but HECTR can also function as an experiment analysis tool and can solve a limited set of other types of containment problems. Version 1.5N is a modification of Version 1.5 and includes changes to the spray actuation logic, and models for steam vents, vacuum breakers, and building cross-vents. Thus, all of the key features of the N Reactor confinement can be modeled. HECTR is designed for flexibility and provides for user control of many important parameters, if built-in correlations and default values are not desired

  10. Analyses of anticipated transient without scram events in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Chun, Ji Han; Kim, Soo Hyoung; Yang, Soo Hyung; Bae, Kyoo Hwan

    2012-01-01

    SMART is a small integral reactor, which was developed at KAERI and acquired standard design approval in 2012. SMART works like a pressurized light water reactor in principle though it is more compact than loop type large commercial reactors. ATWS(Anticipated Transient Without Scram) event is an AOO(Anticipated Operational Occurrence) where RPS fails to trip the reactor when requested. SMART incorporated a DPS(diverse protection system) to protect the reactor system when RPS(reactor protection system) fails to trip the reactor. The results of transient analyses show that DPS in SMART effectively mitigates the consequence of ATWS

  11. Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors; Adaptacion y aplicacion del codigo TRACE para el analisis de transitorios en disenos de reactores rapidos refrigerados por plomo

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, A.; Ammirabile, L.; Martorell, S.

    2015-07-01

    Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

  12. Parameter Identification with the Random Perturbation Particle Swarm Optimization Method and Sensitivity Analysis of an Advanced Pressurized Water Reactor Nuclear Power Plant Model for Power Systems

    Directory of Open Access Journals (Sweden)

    Li Wang

    2017-02-01

    Full Text Available The ability to obtain appropriate parameters for an advanced pressurized water reactor (PWR unit model is of great significance for power system analysis. The attributes of that ability include the following: nonlinear relationships, long transition time, intercoupled parameters and difficult obtainment from practical test, posed complexity and difficult parameter identification. In this paper, a model and a parameter identification method for the PWR primary loop system were investigated. A parameter identification process was proposed, using a particle swarm optimization (PSO algorithm that is based on random perturbation (RP-PSO. The identification process included model variable initialization based on the differential equations of each sub-module and program setting method, parameter obtainment through sub-module identification in the Matlab/Simulink Software (Math Works Inc., Natick, MA, USA as well as adaptation analysis for an integrated model. A lot of parameter identification work was carried out, the results of which verified the effectiveness of the method. It was found that the change of some parameters, like the fuel temperature and coolant temperature feedback coefficients, changed the model gain, of which the trajectory sensitivities were not zero. Thus, obtaining their appropriate values had significant effects on the simulation results. The trajectory sensitivities of some parameters in the core neutron dynamic module were interrelated, causing the parameters to be difficult to identify. The model parameter sensitivity could be different, which would be influenced by the model input conditions, reflecting the parameter identifiability difficulty degree for various input conditions.

  13. Identification of tertiary butyl alcohol (TBA)-utilizing organisms in BioGAC reactors using 13C-DNA stable isotope probing.

    Science.gov (United States)

    Aslett, Denise; Haas, Joseph; Hyman, Michael

    2011-09-01

    Biodegradation of the gasoline oxygenates methyl tertiary-butyl ether (MTBE) and ethyl tertiary-butyl ether (ETBE) can cause tertiary butyl alcohol (TBA) to accumulate in gasoline-impacted environments. One remediation option for TBA-contaminated groundwater involves oxygenated granulated activated carbon (GAC) reactors that have been self-inoculated by indigenous TBA-degrading microorganisms in ground water extracted from contaminated aquifers. Identification of these organisms is important for understanding the range of TBA-metabolizing organisms in nature and for determining whether self-inoculation of similar reactors is likely to occur at other sites. In this study (13)C-DNA-stable isotope probing (SIP) was used to identify TBA-utilizing organisms in samples of self-inoculated BioGAC reactors operated at sites in New York and California. Based on 16S rRNA nucleotide sequences, all TBA-utilizing organisms identified were members of the Burkholderiales order of the β-proteobacteria. Organisms similar to Cupriavidus and Methylibium were observed in both reactor samples while organisms similar to Polaromonas and Rhodoferax were unique to the reactor sample from New York. Organisms similar to Hydrogenophaga and Paucibacter strains were only detected in the reactor sample from California. We also analyzed our samples for the presence of several genes previously implicated in TBA oxidation by pure cultures of bacteria. Genes Mpe_B0532, B0541, B0555, and B0561 were all detected in (13)C-metagenomic DNA from both reactors and deduced amino acid sequences suggested these genes all encode highly conserved enzymes. One gene (Mpe_B0555) encodes a putative phthalate dioxygenase-like enzyme that may be particularly appropriate for determining the potential for TBA oxidation in contaminated environmental samples.

  14. Nuclear power plant transients: where are we

    International Nuclear Information System (INIS)

    Majumdar, D.

    1984-05-01

    This document is in part a postconference review and summary of the American Nuclear Society sponsored Anticipated and Abnormal Plant Transients in Light Water Reactors Conference held in Jackson, Wyoming, September 26-29, 1983, and in part a reflection upon the issues of plant transients and their impact on the viability of nuclear power. This document discusses state-of-the-art knowledge, deficiencies, and future directions in the plant transients area as seen through this conference. It describes briefly what was reported in this conference, emphasizes areas where it is felt there is confidence in the nuclear industry, and also discusses where the experts did not have a consensus. Areas covered in the document include major issues in operational transients, transient management, transient events experience base, the status of the analytical tools and their capabilities, probabilistic risk assessment applications in operational transients, and human factors impact on plant transients management

  15. Construction of a graphic interface for a nuclear reactor modelling and simulation

    International Nuclear Information System (INIS)

    Cadrdenas C, Carlos Roberto; Riquelme R, Raul Antonio.

    1995-01-01

    A graphic interface is presented for real time transient analysis under reactivity insertion, reactor operators training, and the RECH-1 reactor licensing, using the Paret (Program for Analysis of Reactor Transients) computer code. 17 refs., 29 figs

  16. Experience with transients in German NPPs

    International Nuclear Information System (INIS)

    Lindauer, E.

    1984-01-01

    This chapter examines reactor accidents in the Federal Republic of Germany based on the formal reporting system for licensee event reports (LERs) and a special investigation on all unplanned power variations in 3 PWRs. The significant transients experienced by BWR type reactors are analyzed. The main goal is to find weak points which caused the transient or influenced its course in an unfavorable way in order to improve the affected plant and others. The complete survey of all transients, with normally little or no safety relevance, allows statistical evaluations and the analysis of trends. It is concluded that significant transients were mainly experienced at older plants, whereas plants of an advanced design produced very few significant transients. The most frequent human errors which lead to transients are failure search in electronic systems and errors during design and commissioning

  17. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor; Algoritmo neuro-difuso para la identificacion de la potencia neutronica del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.; Benitez R, J. S. [Instituto Tecnologico de Toluca, Division de Estudios de Posgrado e Investigacion, Av. Tecnologico s/n, Ex-Rancho La Virgen, 50140 Metepec, Estado de Mexico (Mexico); Segovia de los Rios, J. A.; Rivero G, T. [ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)], e-mail: jorge.benitez@inin.gob.mx

    2009-10-15

    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  18. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-09-01

    This portion of the RELAP4/MOD5 User's Manual presents the details of setting up and entering the reactor model to be evaluated. The input card format and arrangement is presented in depth, including not only cards for data but also those for editing and restarting. Problem initalization including pressure distribution and energy balance is discussed. A section entitled ''User Guidelines'' is included to provide modeling recommendations, analysis and verification techniques, and computational difficulty resolution. The section is concluded with a discussion of the computer output form and format

  19. An analysis, using the CLAPTRAP code, of the pressure transients developed in the Carolinas Virginia Tube Reactor during containment performance tests

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1982-11-01

    To check containment performance of the CVTR, steam was injected above the operating floor through a 10 foot pipe cap containing the 1 inch diameter holes, at a steady rate of 102.8 lb/sec for a period of 166 seconds. This steam had an enthalpy of 1195 Btu/lb and was therefore not entirely typical of the much wetter material which would be rejected for the greater part of a true breached circuit accident. Pressure transients measured experimentally within the containment were compared with results calculated by the American code CONTEMPT and these results in turn have allowed the Winfrith code CLAPTRAP to be tested for consistency and to establish that the use of this code would have led to similar conclusions about the heat transfer coefficients at the heat absorbent surfaces. (U.K.)

  20. Identification and evaluation of alternatives for the disposition of fluoride fuel and flush salts from the molten salt reactor experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-01-01

    This document presents an initial identification and evaluation of the alternatives for disposition of the fluoride fuel and flush salts stored in the drain tanks at the Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory (ORNL). It will serve as a resource for the U.S. Department of Energy contractor preparing the feasibility study for this activity under the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA). This document will also facilitate further discussion on the range of credible alternatives, and the relative merits of alternatives, throughout the time that a final alternative is selected under the CERCLA process

  1. Molten Salt Reactor Experiment Facility (Building 7503) standards/requirements identification document adherence assessment plan at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    International Nuclear Information System (INIS)

    1996-02-01

    This is the Phase 2 (adherence) assessment plan for the Building 7503 Molten Salt Reactor Experiment (MSRE) Facility standards/requirements identification document (S/RID). This document outlines the activities to be conducted from FY 1996 through FY 1998 to ensure that the standards and requirements identified in the MSRE S/RID are being implemented properly. This plan is required in accordance with the Department of Energy Implementation Plan for Defense Nuclear Facilities Safety Board Recommendation 90-2, November 9, 1994, Attachment 1A. This plan addresses the major aspects of the adherence assessment and will be consistent with Energy Systems procedure QA-2. 7 ''Surveillances.''

  2. FINEDAN - an explicit finite-element calculation code for two-dimensional analyses of fast dynamic transients in nuclear reactor technology

    International Nuclear Information System (INIS)

    Adamik, V.; Matejovic, P.

    1989-01-01

    The problems are discussed of nonstationary, nonlinear dynamics of the continuum. A survey is presented of calculation methods in the given area with emphasis on the area of impact problems. A description is presented of the explicit finite elements method and its application to two-dimensional Cartesian and cylindrical configurations. Using the method the explicit calculation code FINEDAN was written which was tested in a series of verification calculations for different configurations and different types of continuum. The main characteristics are presented of the code and of some, of its practical applications. Envisaged trends of the development of the code and its possible applications in the technology of nuclear reactors are given. (author). 9 figs., 4 tabs., 10 refs

  3. Identification of core components and transient interactors of the peroxisomal importomer by dual-track stable isotope labeling with amino acids in cell culture analysis.

    Science.gov (United States)

    Oeljeklaus, Silke; Reinartz, Benedikt S; Wolf, Janina; Wiese, Sebastian; Tonillo, Jason; Podwojski, Katharina; Kuhlmann, Katja; Stephan, Christian; Meyer, Helmut E; Schliebs, Wolfgang; Brocard, Cécile; Erdmann, Ralf; Warscheid, Bettina

    2012-04-06

    The importomer complex plays an essential role in the biogenesis of peroxisomes by mediating the translocation of matrix proteins across the organellar membrane. A central part of this highly dynamic import machinery is the docking complex consisting of Pex14p, Pex13p, and Pex17p that is linked to the RING finger complex (Pex2p, Pex10p, Pex12p) via Pex8p. To gain detailed knowledge on the molecular players governing peroxisomal matrix protein import and, thus, the integrity and functionality of peroxisomes, we aimed at a most comprehensive investigation of stable and transient interaction partners of Pex14p, the central component of the importomer. To this end, we performed a thorough quantitative proteomics study based on epitope tagging of Pex14p combined with dual-track stable isotope labeling with amino acids in cell culture-mass spectrometry (SILAC-MS) analysis of affinity-purified Pex14p complexes and statistics. The results led to the establishment of the so far most extensive Pex14p interactome, comprising 9 core and further 12 transient components. We confirmed virtually all known Pex14p interaction partners including the core constituents of the importomer as well as Pex5p, Pex11p, Pex15p, and Dyn2p. More importantly, we identified new transient interaction partners (Pex25p, Hrr25p, Esl2p, prohibitin) that provide a valuable resource for future investigations on the functionality, dynamics, and regulation of the peroxisomal importomer.

  4. Calculation programme for the accidental transients in reactors of the gas-graphite type; Programme de calcul des transitoires accidentels des piles de la filiere graphite-gaz

    Energy Technology Data Exchange (ETDEWEB)

    Henri, Ch.; Bayard, J.P. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The study of the behaviour of the fuel during certain incidents or accidents in reactors is closely connected to the study of the changes in temperature. This document describes in the first part the main physical phenomena governing the kinetics of the accident. The aim is to know the temperatures at all points and at all times during the irregular regime which can follow the initial stable regime. In the second part an explanation is given of the numerical methods used. (authors) [French] L'etude du comportement du combustible lors de certains incidents ou accidents de pile est etroitement liee a l'etude de l'evolution des temperatures. Dans sa premiere partie, ce document decrit les phenomenes physiques principaux intervenant dans la cinetique de l'accident. Le but recherche est la connaissance des temperatures en tout point et a tout instant d'un regime varie, faisant suite a un regime initial stable. Dans la deuxieme partie les methodes numeriques employees sont explicitees. (auteurs)

  5. Process Stability Identification Through Dynamic Study of Single-bed Ammonia Reactor with Feed-Effluent Heat Exchanger (FEHE

    Directory of Open Access Journals (Sweden)

    Adhi Tri Partono

    2018-01-01

    Full Text Available In ammonia reactor system, a feed-effluent heat exchanger (FEHE is typically installed to utilize reaction-generated heat to heat the reactor’s feed. Utilizing energy from exothermic reaction to the incoming feed stream is often called “autothermic operation”. Despite the advantage of FEHE, there is a risk of utilizing FEHE in a reactor system such as instability of process temperature or known as hysteresis. Hysteresis phenomena in chemical process could cause operational problems, for example it could damage the integrity of the equipment’s material. This paper aims to evaluate the dynamic behavior of a single-bed ammonia reactor with FEHE, particularly to propose a way to prevent instability within the system. The dynamic simulation of the single-bed ammonia reactor with FEHE was performed with Aspen HYSYS v8.8. The result of the simulation result shows that hysteresis phenomenon in the ammonia reactor system occurs when the feed’s temperature is below a certain value. If the feed temperature reaches that value, the temperature of the reactor’s outlet oscillates. One of the solution to keep the feed temperature above that critical value is by installing a trim heater within the system. Based on the simulation, trim heater installation within the system is able to prevent hysteresis in the system evaluated.

  6. A COMETHE version with transient capability

    International Nuclear Information System (INIS)

    Vliet, J. van; Lebon, G.; Mathieu, P.

    1980-01-01

    A version of the COMETHE code is under development to simulate transient situations. This paper focuses on some aspects of the transient heat transfer models. Initially the coupling between transient heat transfer and other thermomechanical models is discussed. An estimation of the thermal characteristic times shows that the cladding temperatures are often in quasi-steady state. In order to reduce the computing time, calculations are therefore switched from a transient to a quasi-static numerical procedure as soon as such a quasi-equilibrium is detected. The temperature calculation is performed by use of the Lebon-Lambermont restricted variational principle, with piecewise polynoms as trial functions. The method has been checked by comparison with some exact results and yields good agreement for transient as well as for quasi-static situations. This method therefore provides a valuable tool for the simulation of the transient behaviour of nuclear reactor fuel rods. (orig.)

  7. HEDL experimental transient overpower program

    International Nuclear Information System (INIS)

    Hikido, T.; Culley, G.E.

    1976-01-01

    HEDL is conducting a series of experiments to evaluate the performance of Fast Flux Test Facility (FFTF) prototypic fuel pins up to the point of cladding breach. A primary objective of the program is to demonstrate the adequacy of fuel pin and Plant Protective System (PPS) designs for terminated transients. Transient tests of prototypic FFTF fuel pins previously irradiated in the Experimental Breeder Reactor-II (EBR-II) have demonstrated the adequacy of the PPS and fuel pin designs and indicate that a very substantial margin exists between PPS-terminated transients and that required to produce fuel pin cladding failure. Additional experiments are planned to extend the data base to high burnup, high fluence fuel pin specimens

  8. Transient analysis on the SMART-P anticipated transients without scram

    International Nuclear Information System (INIS)

    Yang, S. H.; Bae, K. H.; Kim, H. C.; Zee, S. Q.

    2005-01-01

    Anticipated transients without scram (ATWS) are anticipated operational occurrences accompanied by a failure of an automatic reactor trip when required. Although the occurrence probability of the ATWS events is considerably low, these events can result in unacceptable consequences, i.e. the pressurization of the reactor coolant system (RCS) up to an unacceptable range and a core-melting situation. Therefore, the regulatory body requests the installation of a protection system against the ATWS events. According to the request, a diverse protection system (DPS) is installed in the SMART-P (System-integrated Modular Advanced ReacTor-Pilot). This paper presents the results of the transient analysis performed to identify the performance of the SMART-P against the ATWS. In the analysis, the TASS/SMR (Transients And Setpoint Simulation/Small and Medium Reactor) code is applied to identify the thermal hydraulic response of the RCS during the transients

  9. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  10. Identification of the two-phase flow in the upper part of a boiling water reactor core using reactor noise analysis

    International Nuclear Information System (INIS)

    Miteff, L.

    1983-08-01

    The starting point of this work were neutron flux correlation measurements in the core of the Muehleberg boiling water reactor. During these measurements a new effect was observed i.e. that the cross power spectral density (CPSD) phases could be approximated by two straight lines of different slopes. The two linear domains of the CPSD-phases are separated along the frequency axis by a transition interval. It was supposed that such CPSD-phases could account for the propagation in the axial direction of two different thermohydraulic perturbations in the upper part of the Muehleberg BWR-core. It was also assumed that the second linear domain of these CPSD-phases could be related to a characteristic property of an annular flow regime of the steam-water two-phase flows in the bundles around the neutron detectors. The author attempts to give an explanation for the existence of a second transport phenomenon. This was achieved by in-core and out-core correlation measurements as well as by theoretical work. The out-core measurements were performed on an water-air simulation loop by use of laser beams. (Auth.)

  11. Application of the MOVE algorithm for the identification of reduced order models of a core of a BWR type reactor; Aplicacion del algoritmo MOVE para la identificacion de modelos de orden reducido del nucleo de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Victoria R, M.A.; Morales S, J.B. [UNAM, DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: angelvr@gmail.com

    2005-07-01

    Presently work is applied the modified algorithm of the ellipsoid of optimal volume (MOVE) to a reduced order model of 5 differential equations of the core of a boiling water reactor (BWR) with the purpose of estimating the parameters that model the dynamics. The viability is analyzed of carrying out an analysis that calculates the global dynamic parameters that determine the stability of the system and the uncertainty of the estimate. The modified algorithm of the ellipsoid of optimal volume (MOVE), is a method applied to the parametric identification of systems, in particular to the estimate of groups of parameters (PSE for their initials in English). It is looked for to obtain the ellipsoid of smaller volume that guarantees to contain the real value of the parameters of the model. The PSE MOVE is a recursive identification method that can manage the sign of noise and to ponder it, the ellipsoid represents an advantage due to its easy mathematical handling in the computer, the results that surrender are very useful for the design of Robust Control since to smaller volume of the ellipsoid, better is in general the performance of the system to control. The comparison with other methods presented in the literature to estimate the reason of decline (DR) of a BWR is presented. (Author)

  12. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  13. Pius, self-protective thermohydraulics transient without safety system intervention

    International Nuclear Information System (INIS)

    Fredell, J.; Bredolt, V.

    1989-01-01

    In this paper, the self-protective thermohydraulic feedback of the PIUS reactor system is illustrated by an in-depth discussion of one typical transient. The selected transient is an undetected total loss of feedwater in the complete absence of conventional safety system intervention. The reactor shuts itself down to residual power in two steps. First, the power decreases due to the strongly negative moderator temperature reactivity coefficient, and then a complete shutdown occurs by ingress of cold, highly borated water from the reactor pool. The transient is terminated without any harm to the fuel or paint systems

  14. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  15. Identification of important phenomena under sodium fire accidents based on PIRT process with factor analysis in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

    2016-01-01

    The PIRT (Phenomena Identification and Ranking Table) process is an effective method to identify key phenomena involved in safety issues in nuclear power plants. The present PIRT process is aimed to validate sodium fire analysis codes. Because a sodium fire accident in sodium-cooled fast reactor (SFR) involves complex phenomena, various figures of merit (FOMs) could exist in this PIRT process. In addition, importance evaluation of phenomena for each FOM should be implemented in an objective manner under the PIRT process. This paper describes the methodology for specification of FOMs, identification of associated phenomena and importance evaluation of each associated phenomenon in order to complete a ranking table of important phenomena involved in a sodium fire accident in an SFR. The FOMs were specified through factor analysis in this PIRT process. Physical parameters to be quantified by a sodium fire analysis code were identified by considering concerns resulting from sodium fire in the factor analysis. Associated phenomena were identified through the element- and sequence-based phenomena analyses as is often conducted in PIRT processes. Importance of each associated phenomenon was evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. (author)

  16. Transient pseudohypoaldosteronism

    Directory of Open Access Journals (Sweden)

    Stajić Nataša

    2011-01-01

    Full Text Available Introduction. Infants with urinary tract malformations (UTM presenting with urinary tract infection (UTI are prone to develop transient type 1 pseudohypoaldosteronism (THPA1. Objective. Report on patient series with characteristics of THPA1, UTM and/or UTI and suggestions for the diagnosis and therapy. Methods. Patients underwent blood and urine electrolyte and acid-base analysis, serum aldosterosterone levels and plasma rennin activity measuring; urinalysis, urinoculture and renal ultrasound were done and medical and/or surgical therapy was instituted. Results. Hyponatraemia (120.9±5.8 mmol/L, hyperkalaemia (6.9±0.9 mmol/L, metabolic acidosis (plasma bicarbonate, 11±1.4 mmol/L, and a rise in serum creatinine levels (145±101 μmol/L were associated with inappropriately high urinary sodium (51.3±17.5 mmol/L and low potassium (14.1±5.9 mmol/L excretion. Elevated plasma aldosterone concentrations (170.4±100.5 ng/dL and the very high levels of the plasma aldosterone to potassium ratio (25.2±15.6 together with diminished urinary K/Na values (0.31±0.19 indicated tubular resistance to aldosterone. After institution of appropriate medical and/or surgical therapy, serum electrolytes, creatinine, and acid-base balance were normalized. Imaging studies showed ureteropyelic or ureterovesical junction obstruction in 3 and 2 patients, respectively, posterior urethral valves in 3, and normal UT in 1 patient. According to our knowledge, this is the first report on THPA1 in the Serbian literature. Conclusion. Male infants with hyponatraemia, hyperkalaemia and metabolic acidosis have to have their urine examined and the renal ultrasound has to be done in order to avoid both, the underdiagnosis of THPA1 and the inappropriate medication.

  17. LWR fuel performance during anticipated transients with scram

    International Nuclear Information System (INIS)

    Martinson, Z.R.; McCardell, R.K.; MacDonanl, P.E.; Rowland, T.C.; Tokar, M.

    1983-01-01

    Operational transients occur occasionally in light water reactors when minor malfunctions of certain system components affect the reactor core. Potential effects of such malfunctions include a loss of the secondary heat sink, an increase in system pressure, and, in boiling water reactors, void collapse and a brief increase in reactor power. The most severe postulated Boiling Water Reactor (BWR) anticipated transient is characterized by a power peak of up to 495% rated power for about 1 second (according to a recent General Electric Co., generic analysis). The results of a series of fuel behaviour tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory are presented in this paper. Four progressively higher and broader power transients at a constant coolant flow rate were performed. The first transient simulated a BWR-5 turbine trip without steam bypass with fuel rods operating at BWR-6 core average rod powers. The second transient simulated a generator load rejection without steam bypass with fuel rods operating at above core average powers. The last two transients were performed at higher powers than safety analysis predicts to be possible in commercial reactors to be defined failure threshold margins. The test rods did not fail and were not damaged during any of the four transients. (author)

  18. Transient survivability of LMR oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

    1986-01-01

    Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

  19. Identification and genome organization of saponin pathway genes from a wild crucifer, and their use for transient production of saponins in Nicotiana benthamiana.

    Science.gov (United States)

    Khakimov, Bekzod; Kuzina, Vera; Erthmann, Pernille Ø; Fukushima, Ery Odette; Augustin, Jörg M; Olsen, Carl Erik; Scholtalbers, Jelle; Volpin, Hanne; Andersen, Sven Bode; Hauser, Thure P; Muranaka, Toshiya; Bak, Søren

    2015-11-01

    The ability to evolve novel metabolites has been instrumental for the defence of plants against antagonists. A few species in the Barbarea genus are the only crucifers known to produce saponins, some of which make plants resistant to specialist herbivores, like Plutella xylostella, the diamondback moth. Genetic mapping in Barbarea vulgaris revealed that genes for saponin biosynthesis are not clustered but are located in different linkage groups. Using co-location with quantitative trait loci (QTLs) for resistance, transcriptome and genome sequences, we identified two 2,3-oxidosqualene cyclases that form the major triterpenoid backbones. LUP2 mainly produces lupeol, and is preferentially expressed in insect-susceptible B. vulgaris plants, whereas LUP5 produces β-amyrin and α-amyrin, and is preferentially expressed in resistant plants; β-amyrin is the backbone for the resistance-conferring saponins in Barbarea. Two loci for cytochromes P450, predicted to add functional groups to the saponin backbone, were identified: CYP72As co-localized with insect resistance, whereas CYP716As did not. When B. vul