WorldWideScience

Sample records for venus simulated pwr

  1. Upgrading of PWR plant simulators

    International Nuclear Information System (INIS)

    Wada, Tomonori; Sasaki, Kazunori; Nakaishi, Hirokazu.

    1989-01-01

    For the education and training of operators in electric power plants, simulators have been employed, and it is well known that their effect is great. There are operation training simulators which simulate the dynamic characteristics of plants and all the machinery and equipment that operators handle, and train the procedure of restoration at the time of abnormality in plants, education simulators which can analyze the dynamic characteristics of plants efficiently in a short time, and offer information by visualizing phenomena with three-dimensional display and others so as to be easily understandable, and forecast simulators which do the analysis forecasting plant behavior at the time of abnormality in plants, and investigate the necessity of the guide for operation procedure and the countermeasures at the time of emergency. In this explanation, the upgrading of operation training simulators which have been put already in training is discussed. The constitution of simulator system and the instructor function, the outline of PWR plant simulation models comprising thermal flow model, pump model, leak model and so on, the techniques of increasing simulator speed, and the example of analysis using the NUPAC code are reported. (K.I.)

  2. Simulation model of a PWR power plant

    International Nuclear Information System (INIS)

    Larsen, N.

    1987-03-01

    A simulation model of a hypothetical PWR power plant is described. A large number of disturbances and failures in plant function can be simulated. The model is written as seven modules to the modular simulation system for continuous processes DYSIM and serves also as a user example of this system. The model runs in Fortran 77 on the IBM-PC-AT. (author)

  3. Venus

    CERN Document Server

    Payment, Simone

    2017-01-01

    This straightforward but fascinating book takes a close look at Venus and shows young people just how different our neighboring planet is from our own. Known as the hottest planet, Venus is an example of the greenhouse effect to the extreme. Young readers will take a tour beneath the sulfur dioxide clouds and see the planet's surface up close with images taken by the Magellan and the Venus Express missions. This book will surely fascinate any young person interested in alien worlds.

  4. Microcomputer simulation of PWR power plant pressurizer

    International Nuclear Information System (INIS)

    Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

    1990-01-01

    It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

  5. MELCOR/VISOR PWR desktop simulator

    International Nuclear Information System (INIS)

    With, Anka de; Wakker, Pieter

    2010-01-01

    Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

  6. Model for transient simulation in a PWR steam circuit

    International Nuclear Information System (INIS)

    Mello, L.A. de.

    1982-11-01

    A computer code (SURF) was developed and used to simulate pressure losses along the tubes of the main steam circuit of a PWR nuclear power plant, and the steam flow through relief and safety valves when pressure reactors its thresholds values. A thermodynamic model of turbines (high and low pressure), and its associated components are simulated too. The SURF computer code was coupled to the GEVAP computer code, complementing the simulation of a PWR nuclear power plant main steam circuit. (Author) [pt

  7. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  8. Methodology for the LABIHS PWR simulator modernization

    Energy Technology Data Exchange (ETDEWEB)

    Jaime, Guilherme D.G.; Oliveira, Mauro V., E-mail: gdjaime@ien.gov.b, E-mail: mvitor@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  9. Methodology for the LABIHS PWR simulator modernization

    International Nuclear Information System (INIS)

    Jaime, Guilherme D.G.; Oliveira, Mauro V.

    2011-01-01

    The Human-System Interface Laboratory (LABIHS) simulator is composed by a set of advanced hardware and software components whose goal is to simulate the main characteristics of a Pressured Water Reactor (PWR). This simulator serves for a set of purposes, such as: control room modernization projects; designing of operator aiding systems; providing technological expertise for graphical user interfaces (GUIs) designing; control rooms and interfaces evaluations considering both ergonomics and human factors aspects; interaction analysis between operators and the various systems operated by them; and human reliability analysis in scenarios considering simulated accidents and normal operation. The simulator runs in a PA-RISC architecture server (HPC3700), developed nearby 2000's, using the HP-UX operating system. All mathematical modeling components were written using the HP Fortran-77 programming language with a shared memory to exchange data from/to all simulator modules. Although this hardware/software framework has been discontinued in 2008, with costumer support ceasing in 2013, it is still used to run and operate the simulator. Due to the fact that the simulator is based on an obsolete and proprietary appliance, the laboratory is subject to efficiency and availability issues, such as: downtime caused by hardware failures; inability to run experiments on modern and well known architectures; and lack of choice of running multiple simulation instances simultaneously. This way, there is a need for a proposal and implementation of solutions so that: the simulator can be ported to the Linux operating system, running on the x86 instruction set architecture (i.e. personal computers); we can simultaneously run multiple instances of the simulator; and the operator terminals run remotely. This paper deals with the design stage of the simulator modernization, in which it is performed a thorough inspection of the hardware and software currently in operation. Our goal is to

  10. The simulation research for the dynamic performance of integrated PWR

    International Nuclear Information System (INIS)

    Yuan Jiandong; Xia Guoqing; Fu Mingyu

    2005-01-01

    The mathematical model of the reactor core of integrated PWR has been studied and simplified properly. With the lumped parameter method, authors have established the mathematical model of the reactor core, including the neutron dynamic equation, the feedback reactivities model and the thermo-hydraulic model of the reactor. Based on the above equations and models, the incremental transfer functions of the reactor core model have been built. By simulation experimentation, authors have compared the dynamic characteristics of the integrated PWR with the traditional dispersed PWR. The simulation results show that the mathematical models and equations are correct. (authors)

  11. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR)

    International Nuclear Information System (INIS)

    Valle H, J.; Hidago H, F.; Morales S, J.B.

    2007-01-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  12. The latest full-scale PWR simulator in Japan

    International Nuclear Information System (INIS)

    Nishimuru, Y.; Tagi, H.; Nakabayashi, T.

    2004-01-01

    The latest MHI Full-scale Simulator has an excellent system configuration, in both flexibility and extendability, and has highly sophisticated performance in PWR simulation by the adoption of CANAC-II and PRETTY codes. It also has an instructive character to display the plant's internal status, such as RCS condition, through animation. Further, the simulation has been verified to meet a functional examination at model plant, and with a scale model test result in a two-phase flow event, after evaluation for its accuracy. Thus, the Simulator can be devoted to a sophisticated and broad training course on PWR operation. (author)

  13. Design of a PWR emergency core cooling simulator loop

    International Nuclear Information System (INIS)

    Melo, C.A. de.

    1982-12-01

    The preliminary design of a PWR Emergency Core Cooling Simulator Loop for investigations of the phenomena involved in a postulated Loss-of-Coolant Accident, during the Reflooding Phase, is presented. The functions of each component of the loop, the design methods and calculations, the specification of the instrumentation, the system operation sequence, the materials list and a cost assessment are included. (Author) [pt

  14. Contribution to study and design of PWR plant simulation code

    International Nuclear Information System (INIS)

    Delourme, Didier.

    1980-11-01

    This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

  15. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  16. A nodal model for the simulation of a PWR core

    International Nuclear Information System (INIS)

    Souza Pinto, R. de.

    1981-06-01

    A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

  17. Calculational limitations in PWR system simulation

    International Nuclear Information System (INIS)

    Abramson, P.B.; Kennedy, M.F.; Speis, T.P.

    1982-01-01

    Engineering transient analysis codes, which are in general more accurate than the present generation of simulator software, can be expected to yield reasonably accurate results (+-20% or so on system pressure) if carefully utilized and if the two-phase and transient flow conditions are not severe. As the severity of the transient increases, the confidence that one may have in the results decreases. None of the existing engineering analysis codes is well assessed or verified for transient analysis, but all give qualitatively the same results lending credence to their results. Recent comparisons to transients in LOFT and SEMISCALE are encouraging as are various comparisons to actual plant data

  18. Study on virtual simulation technology for operation and control of PWR

    International Nuclear Information System (INIS)

    Fang Baoguo; Zhang Dafa; Lin Yajun

    2006-01-01

    The way to build graphical models of PWR with MultiGen Creator is discussed, and the three-dimensional model used in the virtual simulation is built. The mathematical simulation model for PWR based on the platform of MFC and Vega is built through the analysis of the mathematical simulation of PWR. The way to perform the virtual effect is introduced associating with the Pressurizer. And, all above parts are connected in one with VC++ to perform the whole virtual simulation of PWR. (authors)

  19. Interface tracking simulations of bubbly flows in PWR relevant geometries

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

    2017-02-15

    Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

  20. Natural-circulation-cooling characteristics during PWR accident simulations

    International Nuclear Information System (INIS)

    Adams, J.P.; McCreery, G.E.; Berta, V.T.

    1983-01-01

    A description of natural circulation cooling characteristics is presented. Data were obtained from several pressurized water reactor accident simulations in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). The reliability of natural circulation cooling, its cooling effectiveness, and the effect of changing system conditions are described. Quantitative comparison of flow rates and time constants with theory for both single- and two-phase fluid conditions were made. It is concluded that natural circulation cooling can be relied on in plant recovery procedures in the absence of forced convection whenever the steam generator heat sink is available

  1. Abnormal transient analysis by using PWR plant simulator, (2)

    International Nuclear Information System (INIS)

    Naitoh, Akira; Murakami, Yoshimitsu; Yokobayashi, Masao.

    1983-06-01

    This report describes results of abnormal transient analysis by using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at EOL. In the simulator, malfunctions are provided for abnormal conditions of equipment failures, and in this report, 17 malfunctions for secondary system and 4 malfunctions for nuclear instrumentation systems were simulated. The abnormal conditions are turbine and generator trip, failure of condenser, feedwater system and valve and detector failures of pressure and water level. Fathermore, failure of nuclear instrumentations are involved such as source range channel, intermediate range channel and audio counter. Transient behaviors caused by added malfunctions were reasonable and detail information of dynamic characteristics for turbine-condenser system were obtained. (author)

  2. Transient analysis of multifailure conditions by using PWR plant simulator

    International Nuclear Information System (INIS)

    Morisaki, Hidetoshi; Yokobayashi, Masao.

    1984-11-01

    This report describes results of the analysis of abnormal transients caused by multifailures using a PWR plant simulator. The simulator is based on an existing 822MWe power plant with 3 loops, and designed to cover wide range of plant operation from cold shutdown to full power at the end of life. Various malfunctions to simulate abnormal conditions caused by equipment failures are provided. In this report, features of abnormal transients caused by concurrence of malfunctions are discussed. The abnormal conditions studied are leak of primary coolant, loss of charging and feedwater flows, and control systems failure. From the results, it was observed that transient responses caused by some of the malfunctions are almost same as the addition of behaviors caused by each single malfunction. Therefore, it can be said that kinds of malfunctions which are concurrent may be estimated from transient characteristics of each single malfunction. (author)

  3. High Altitude Venus Operations Concept Trajectory Design, Modeling and Simulation

    Science.gov (United States)

    Lugo, Rafael A.; Ozoroski, Thomas A.; Van Norman, John W.; Arney, Dale C.; Dec, John A.; Jones, Christopher A.; Zumwalt, Carlie H.

    2015-01-01

    A trajectory design and analysis that describes aerocapture, entry, descent, and inflation of manned and unmanned High Altitude Venus Operation Concept (HAVOC) lighter-than-air missions is presented. Mission motivation, concept of operations, and notional entry vehicle designs are presented. The initial trajectory design space is analyzed and discussed before investigating specific trajectories that are deemed representative of a feasible Venus mission. Under the project assumptions, while the high-mass crewed mission will require further research into aerodynamic decelerator technology, it was determined that the unmanned robotic mission is feasible using current technology.

  4. Implementation in free software of the PWR type university nucleo electric simulator (SU-PWR); Implementacion en software libre del simulador universitario de nucleoelectrica tipo PWR (SU-PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Valle H, J.; Hidago H, F.; Morales S, J.B. [UNAM, Laboratorio de Analisis de Ingenieria de Reactores Nucleares DEPFI, Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: julfi_jg@yahoo.com.mx

    2007-07-01

    Presently work is shown like was carried out the implementation of the University Simulator of Nucleo-electric type PWR (SU-PWR). The implementation of the simulator was carried out in a free software simulation platform, as it is Scilab, what offers big advantages that go from the free use and without cost of the product, until the codes modification so much of the system like of the program with the purpose of to improve it or to adapt it to future routines and/or more advanced graphic interfaces. The SU-PWR shows the general behavior of a PWR nuclear plant (Pressurized Water Reactor) describing the dynamics of the plant from the generation process of thermal energy in the nuclear fuel, going by the process of energy transport toward the coolant of the primary circuit the one which in turn transfers this energy to the vapor generators of the secondary circuit where the vapor is expanded by means of turbines that in turn move the electric generator producing in this way the electricity. The pressurizer that is indispensable for the process is also modeled. Each one of these stages were implemented in scicos that is the Scilab tool specialized in the simulation. The simulation was carried out by means of modules that contain the differential equation that mathematically models each stage or equipment of the PWR plant. The result is a series of modules that based on certain entrances and characteristic of the system they generate exits that in turn are the entrance to other module. Because the SU-PWR is an experimental project in early phase, it is even work and modifications to carry out, for what the models that are presented in this work can vary a little the being integrated to the whole system to simulate, but however they already show clearly the operation and the conformation of the plant. (Author)

  5. Integrated training support system for PWR operator training simulator

    International Nuclear Information System (INIS)

    Sakaguchi, Junichi; Komatsu, Yasuki

    1999-01-01

    The importance of operator training using operator training simulator has been recognized intensively. Since 1986, we have been developing and providing many PWR simulators in Japan. We also have developed some training support systems connected with the simulator and the integrated training support system to improve training effect and to reduce instructor's workload. This paper describes the concept and the effect of the integrated training support system and of the following sub-systems. We have PES (Performance Enhancement System) that evaluates training performance automatically by analyzing many plant parameters and operation data. It can reduce the deviation of training performance evaluation between instructors. PEL (Parameter and Event data Logging system), that is the subset of PES, has some data-logging functions. And we also have TPES (Team Performance Enhancement System) that is used aiming to improve trainees' ability for communication between operators. Trainee can have conversation with virtual trainees that TPES plays automatically. After that, TPES automatically display some advice to be improved. RVD (Reactor coolant system Visual Display) displays the distributed hydraulic-thermal condition of the reactor coolant system in real-time graphically. It can make trainees understand the inside plant condition in more detail. These sub-systems have been used in a training center and have contributed the improvement of operator training and have gained in popularity. (author)

  6. PWR system simulation and parameter estimation with neural networks

    International Nuclear Information System (INIS)

    Akkurt, Hatice; Colak, Uener

    2002-01-01

    A detailed nonlinear model for a typical PWR system has been considered for the development of simulation software. Each component in the system has been represented by appropriate differential equations. The SCILAB software was used for solving nonlinear equations to simulate steady-state and transient operational conditions. Overall system has been constructed by connecting individual components to each other. The validity of models for individual components and overall system has been verified. The system response against given transients have been analyzed. A neural network has been utilized to estimate system parameters during transients. Different transients have been imposed in training and prediction stages with neural networks. Reactor power and system reactivity during the transient event have been predicted by the neural network. Results show that neural networks estimations are in good agreement with the calculated response of the reactor system. The maximum errors are within ±0.254% for power and between -0.146 and 0.353% for reactivity prediction cases. Steam generator parameters, pressure and water level, are also successfully predicted by the neural network employed in this study. The noise imposed on the input parameters of the neural network deteriorates the power estimation capability whereas the reactivity estimation capability is not significantly affected

  7. PWR system simulation and parameter estimation with neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Akkurt, Hatice; Colak, Uener E-mail: uc@nuke.hacettepe.edu.tr

    2002-11-01

    A detailed nonlinear model for a typical PWR system has been considered for the development of simulation software. Each component in the system has been represented by appropriate differential equations. The SCILAB software was used for solving nonlinear equations to simulate steady-state and transient operational conditions. Overall system has been constructed by connecting individual components to each other. The validity of models for individual components and overall system has been verified. The system response against given transients have been analyzed. A neural network has been utilized to estimate system parameters during transients. Different transients have been imposed in training and prediction stages with neural networks. Reactor power and system reactivity during the transient event have been predicted by the neural network. Results show that neural networks estimations are in good agreement with the calculated response of the reactor system. The maximum errors are within {+-}0.254% for power and between -0.146 and 0.353% for reactivity prediction cases. Steam generator parameters, pressure and water level, are also successfully predicted by the neural network employed in this study. The noise imposed on the input parameters of the neural network deteriorates the power estimation capability whereas the reactivity estimation capability is not significantly affected.

  8. PWR plant operator training used full scope simulator incorporated MAAP model

    International Nuclear Information System (INIS)

    Matsumoto, Y.; Tabuchi, T.; Yamashita, T.; Komatsu, Y.; Tsubouchi, K.; Banka, T.; Mochizuki, T.; Nishimura, K.; Iizuka, H.

    2015-01-01

    NTC makes an effort with the understanding of plant behavior of core damage accident as part of our advanced training. For the Fukushima Daiichi Nuclear Power Station accident, we introduced the MAAP model into PWR operator training full scope simulator and also made the Severe Accident Visual Display unit. From 2014, we will introduce new training program for a core damage accident with PWR operator training full scope simulator incorporated the MAAP model and the Severe Accident Visual Display unit. (author)

  9. Achievement of a training simulator for PWR power plant: application to control parametric studies

    International Nuclear Information System (INIS)

    Salomon-Sigogne, A.

    1982-09-01

    A simulation tool adapted to training tasks is developed. One presents the description of the simulator. One studies the management of a model by NEPTUN X2. A general description of a 900 MW PWR power station and the modelling of the power station are presented. The results obtained on the FIDIANE version of the simulator are finally analyzed [fr

  10. Mathematical modelling of plant transients in the PWR for simulator purposes

    International Nuclear Information System (INIS)

    Hartel, K.

    1984-01-01

    This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

  11. SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

    International Nuclear Information System (INIS)

    Dias, A.F.V.

    1980-10-01

    Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

  12. SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

    International Nuclear Information System (INIS)

    Farvacque, M.; Faydide, B.; Dufeil, Ph.; Raimond, E.

    2003-01-01

    The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients

  13. Simulation and beam line experiments for the superconducting ECR ion source VENUS

    International Nuclear Information System (INIS)

    Todd, Damon S.; Leitner, Daniela; Grote, David P.; Lyneis, ClaudeM.

    2007-01-01

    The particle-in-cell code Warp has been enhanced to incorporate both two- and three-dimensional sheath extraction models giving Warp the capability of simulating entire ion beam transport systems including the extraction of beams from plasma sources. In this article we describe a method of producing initial ion distributions for plasma extraction simulations in electron cyclotron resonance (ECR) ion sources based on experimentally measured sputtering on the source biased disc. Using this initialization method, we present preliminary results for extraction and transport simulations of an oxygen beam and compare them with experimental beam imaging on a quartz viewing plate for the superconducting ECR ion source VENUS

  14. Numerical simulation of the heating and start-up of PWR nuclear power station

    International Nuclear Information System (INIS)

    Faraco-Medeiros, M.A.; Leite, C.A.T.; Ramalho, F.P.

    1992-01-01

    The start-up of a PWR nuclear power plant must be done within safety criteria and requires a simulation. The design of some equipment, cost and time can be optimized. A computer simulator, which allows control of all the equipment and variables into the operation, has been developed and is presented in this paper. The KWU procedure and an alternative for Angra II were simulated. The results are showed up. 09 refs, 03 figs. (B.C.A.)

  15. Rupther: a simulation approach applied to a PWR vessel failure during a severe accident

    International Nuclear Information System (INIS)

    Mongabure, Ph.; Nicolas, L.; Devos, J.

    2000-01-01

    The Rupther program (Rupture Under Thermal Conditions) is a part of the international researches on severe accidents in the PWR type reactors. The aim of the program is the definition of failure simulation validated by experimental data on vessel steel 16MND5 mechanical properties. The paper presents the program and the first results. (A.L.B.)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  17. CFD simulation of a four-loop PWR at asymmetric operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Jian-Ping; Yan, Li-Ming; Li, Feng-Chen, E-mail: lifch@hit.edu.cn

    2016-04-15

    Highlights: • A CFD numerical simulation procedure was established for simulating RPV of VVER-1000. • The established CFD approach was validated by comparing with available data. • Thermal hydraulic characteristics under asymmetric operation condition were investigated. • Apparent influences of the shutdown loop on its neighboring loops were obtained. - Abstract: The pressurized water reactor (PWR) with multiple loops may have abnormal working conditions with coolant pumps out of running in some loops. In this paper, a computational fluid dynamics (CFD) numerical study of the four-loop VVER-1000 PWR pressure vessel model was presented. Numerical simulations of the thermohydrodynamic characteristics in the pressure vessel were carried out at different inlet conditions with four and three loops running, respectively. At normal stead-state condition (four-loop running), different parameters were obtained for the full fluid domain, including pressure losses across different parts, pressure, velocity and temperature distributions in the reactor pressure vessel (RPV) and mass flow distribution of the coolant at the inlet of reactor core. The obtained results for pressure losses matched with the experimental reference values of the VVER-1000 PWR at Tianwan nuclear power plant (NPP). For most fuel assemblies (FAs), the inlet flow rates presented a symmetrical distribution about the center under full-loop operation conditions, which accorded with the practical distribution. These results indicate that it is now possible to study the dynamic transition process between different asymmetric operation conditions in a multi-loop PWR using the established CFD method.

  18. 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

  19. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  20. Simulation of fission products behavior in severe accidents for advanced passive PWR

    International Nuclear Information System (INIS)

    Tong, L.L.; Huang, G.F.; Cao, X.W.

    2015-01-01

    Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

  1. Investigation of modeling and simulation on a PWR power conversion system with RELAP5

    International Nuclear Information System (INIS)

    Rui Gao; Yang Yanhua; Lin Meng; Yuan Minghao; Xie Zhengrui

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Dayabay nuclear power station, this paper models the thermal-hydraulic systems for PWR by using the best-estimate program, RELAP5. To simulate the full-scope power conversion system, not only the reactor coolant system (RCP) of nuclear island, but also the main steam system (VVP), turbine steam and drain system (GPV), bypass system (GCT), feedwater system (FW), condensate extraction system (CEX), moisture separator reheater system (GSS), turbine-driven feedwater pump (APP), low-pressure and high-pressure feedwater heater systems (ABP and AHP) of conventional island are considered and modeled. A comparison between the simulated results and the actual data of reactor under full-power demonstrates a fine match for Dayabay, and also manifests the feasibility in simulating full-scope power conversion system of PWR with RELAP5. (author)

  2. Simulation of Venus polar vortices with the non-hydrostatic general circulation model

    Science.gov (United States)

    Rodin, Alexander V.; Mingalev, Oleg; Orlov, Konstantin

    2012-07-01

    The dynamics of Venus atmosphere in the polar regions presents a challenge for general circulation models. Numerous images and hyperspectral data from Venus Express mission shows that above 60 degrees latitude atmospheric motion is substantially different from that of the tropical and extratropical atmosphere. In particular, extended polar hoods composed presumably of fine haze particles, as well as polar vortices revealing mesoscale wave perturbations with variable zonal wavenumbers, imply the significance of vertical motion in these circulation elements. On these scales, however, hydrostatic balance commonly used in the general circulation models is no longer valid, and vertical forces have to be taken into account to obtain correct wind field. We present the first non-hydrostatic general circulation model of the Venus atmosphere based on the full set of gas dynamics equations. The model uses uniform grid with the resolution of 1.2 degrees in horizontal and 200 m in the vertical direction. Thermal forcing is simulated by means of relaxation approximation with specified thermal profile and time scale. The model takes advantage of hybrid calculations on graphical processors using CUDA technology in order to increase performance. Simulations show that vorticity is concentrated at high latitudes within planetary scale, off-axis vortices, precessing with a period of 30 to 40 days. The scale and position of these vortices coincides with polar hoods observed in the UV images. The regions characterized with high vorticity are surrounded by series of small vortices which may be caused by shear instability of the zonal flow. Vertical velocity component implies that in the central part of high vorticity areas atmospheric flow is downwelling and perturbed by mesoscale waves with zonal wavenumbers 1-4, resembling observed wave structures in the polar vortices. Simulations also show the existence of areas with strong vertical flow, concentrated in spiral branches extending

  3. PWR station blackout transient simulation in the INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

    2004-01-01

    Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

  4. DOMPAC dosimetry experiment. Neutronic simulation of the thickness of a PWR pressure vessel. Irradiation damages

    International Nuclear Information System (INIS)

    Alberman, A.; Faure, M.; Thierry, M.; Hoclet, O.; Le Dieu de Ville, A.; Nimal, J.C.; Soulat, P.

    1979-01-01

    For suitable extrapolation of irradiated PWR ferritic steel results, proper irradiation of the pressure vessel has been 'simulated' in test reactor. For this purpose, a huge steel block (20 cm in depth) was loaded with Saclay's graphite (GAMIN) and tungsten damage detectors. Core-block water gap was optimized through spectrum indexes method, by ANISN and SABINE codes so that spectrum in 1/4 thickness matches with ANISN computations for PWR Fessenheim 1. A good experimental agreement is found with calculated dpa damage gradient. 3D Monte Carlo computation (TRIPOLI), was performed on the DOMPAC device, and spectrum indexes evolution was found consistent with experimental results. Surveillance rigs behind a 'thermal shield' were also simulated, including damage and activation monitors. Dosimetry results give an order of magnitude of accuracies involved in projecting steel sample embrittlement to the pressure vessel [fr

  5. A new model for simulation of pressurizers in PWR power plants

    International Nuclear Information System (INIS)

    Madeira, A.A.

    1981-02-01

    The pressurizer of a PWR type reactor was simulated as a thermodynamical system made up of three regions with movable boundaries. The mechanisms of normal condensation, condensation induced by spray, flashing and heat exchange across the water - steam interface, were studied. Various tests have been carried out and satisfactory results were obtained when compared with those from other models and also with some available experimental data. (E.G.) [pt

  6. Comparison of thermal behavior of different PWR fuel rod simulators for LOCA experiments

    International Nuclear Information System (INIS)

    Casal, V.; Malang, S.; Rust, K.

    1982-10-01

    For experimental investigations of a loss-of-coolant accident (LOCA) of a PWR electrical heater rods are applied as thermal fuel rod simulators. To substitute heater rods from the SEMISCALE program by INTERATOM-KfK heater rods in a current experimental program at the Instituut for Energiteknikk-(OECD-Halden), the thermodynamic behavior of different heater rods during a LOCA were compared. The results show, that SEMISCALE-heater rods can be replaced by those fabricated by INTERATOM. (orig.) [de

  7. Principle simulator for a PWR nuclear power station

    International Nuclear Information System (INIS)

    Wahlstroem, B.

    1975-05-01

    A report is given on a simulator developed for the training of operational and planning staff for the Lovisa nuclear power station in Finland. All main components of the power station are illustrated and trainees can operate the simulator in the power range 3-100 %. The model was originally developed for planning the control system of Lovisa I, for which reason the simulator project could be carried out on a relatively limited budget. (author)

  8. Mathematical model for the simulation of PWR power plant

    International Nuclear Information System (INIS)

    Delfosse, C.

    1979-01-01

    A reactor simulator representing the principal characteristics of a nuclear power plant and their regulation has been developed. Special attention has been devoted to the simulation of the pressurizer, the steam turbine and the valves. Numerical tests have been realised in order to verify the speed of the calculations. (MDC)

  9. Optimization of PWR fuel assembly radial enrichment and burnable poison location based on adaptive simulated annealing

    International Nuclear Information System (INIS)

    Rogers, Timothy; Ragusa, Jean; Schultz, Stephen; St Clair, Robert

    2009-01-01

    The focus of this paper is to present a concurrent optimization scheme for the radial pin enrichment and burnable poison location in PWR fuel assemblies. The methodology is based on the Adaptive Simulated Annealing (ASA) technique, coupled with a neutron lattice physics code to update the cost function values. In this work, the variations in the pin U-235 enrichment are variables to be optimized radially, i.e., pin by pin. We consider the optimization of two categories of fuel assemblies, with and without Gadolinium burnable poison pins. When burnable poisons are present, both the radial distribution of enrichment and the poison locations are variables in the optimization process. Results for 15 x 15 PWR fuel assembly designs are provided.

  10. An immersed body method for coupled neutron transport and thermal hydraulic simulations of PWR assemblies

    International Nuclear Information System (INIS)

    Jewer, S.; Buchan, A.G.; Pain, C.C.; Cacuci, D.G.

    2014-01-01

    Highlights: • A new method of coupled radiation transport, heat and momentum exchanges on fluids, and heat transfer simulations. • Simulation of the thermal hydraulics and radiative properties within whole PWR assemblies. • An immersed body method for modelling complex solid domains on practical computational meshes. - Abstract: A recently developed immersed body method is adapted and used to model a typical pressurised water reactor (PWR) fuel assembly. The approach is implemented with the numerical framework of the finite element, transient criticality code, FETCH which is composed of the neutron transport code, EVENT, and the CFD code, FLUIDITY. Within this framework the neutron transport equation, Navier–Stokes equations and a fluid energy conservation equation are solved in a coupled manner on a coincident structured or unstructured mesh. The immersed body method has been used to model the solid fuel pins. The key feature of this method is that the fluid/neutronic domain and the solid domain are represented by overlapping and non-conforming meshes. The main difficulty of this approach, for which a solution is proposed in this work, is the conservative mapping of the energy and momentum exchange between the fluid/neutronic mesh and the solid fuel pin mesh. Three numerical examples are presented which include a validation of the fuel pin submodel against an analytical solution; an uncoupled (no neutron transport solution) PWR fuel assembly model with a specified power distribution which was validated against the COBRA-EN subchannel analysis code; and finally a coupled model of a PWR fuel assembly with reflective neutron boundary conditions. Coupling between the fluid and neutron transport solutions is through the nuclear cross sections dependence on Doppler fuel temperature, coolant density and temperature, which was taken into account by using pre-calculated cross-section lookup tables generated using WIMS9a. The method was found to show good agreement

  11. Modeling and simulation of pressurizer dynamic process in PWR nuclear power plant

    International Nuclear Information System (INIS)

    Ma Jin; Liu Changliang; Li Shu'na

    2010-01-01

    By analysis of the actual operating characteristics of pressurizer in pressurized water reactor (PWR) nuclear power plant and based on some reasonable simplification and basic assumptions, the quality and energy conservation equations about pressurizer' s steam zone and the liquid zone are set up. The purpose of this paper is to build a pressurizer model of two imbalance districts. Water level and pressure control system of pressurizer is formed though model encapsulation. Dynamic simulation curves of main parameters are also shown. At last, comparisons between the theoretical analysis and simulation results show that the pressurizer model of two imbalance districts is reasonable. (authors)

  12. Nonlinear punctual dynamic applied to simulation of PWR type reactors

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-01-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

  13. Non-linear punctual kinetics applied to PWR reactors simulation

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-11-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model for the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and CSMP using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (Author) [pt

  14. N2O and CO production by electric discharge - Atmospheric implications. [Venus atmosphere simulation

    Science.gov (United States)

    Levine, J. S.; Howell, W. E.; Hughes, R. E.; Chameides, W. L.

    1979-01-01

    Enhanced levels of N2O and CO were measured in tropospheric air samples exposed to a 17,500-J laboratory discharge. These enhanced levels correspond to an N2O production rate of about 4 trillion molecules/J and a CO production rate of about 10 to the 14th molecules/J. The CO measurements suggest that the primary region of chemical production in the discharge is the shocked air surrounding the lightning channel, as opposed to the slower-cooling inner core. Additional experiments in a simulated Venus atmosphere (CO2 - 95%, N2 - 5%, at one atmosphere) indicate an enhancement of CO from less than 0.1 ppm prior to the laboratory discharge to more than 2000 ppm after the discharge. Comparison with theoretical calculations appears to confirm the ability of a shock-wave/thermochemical model to predict the rate of production of trace species by an electrical discharge.

  15. Neutron spectrometry around the VENUS reactor using Monte Carlo simulations and Bonner spheres measurements

    International Nuclear Information System (INIS)

    Coeck, M.; Lacoste, V.; Muller, H.

    2005-01-01

    Full text: Reliable determination of neutron doses in workplaces is still an issue in the field of radiation protection. The EVIDOS project ('evaluation of individual dosimetry in mixed neutron and photon radiation fields', 5FP supported by the EC) aims to evaluate different methods for individual dosimetry in mixed neutron-photon workplaces in nuclear industry, and focuses on the neutron component. This objective cannot be reached on the basis of investigations in calibration fields only, but requires studies in representative workplaces of the nuclear industry. The VENUS reactor, a zero-power research reactor established by the SCK·CEN, was chosen as one of these workplaces. This paper presents the assessment of the neutron field near the VENUS reactor, particularly in areas near the reactor shielding and in the control room where operators are frequently present during a reactor run. From the neutron spectrum, an evaluation of H*(10) can be made. MCNPX simulations were performed to obtain a reference spectrum at the two areas of interest. Using a k eff calculation the source term was acquired which was subsequently used in a fixed source MCNPX model of the complete shielding geometry of the reactor hall. Reference spectrometry was also performed using a Bonner spheres system. The unfolding spectra were obtained using the NUBAY and GRAVEL codes. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterized spectrum and provides posterior probability distributions for both the set of parameters and a set of integral quantities. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. Bonner spheres data GRAVEL unfolding was also performed using the MCNPX spectra as an initial guess. In this paper the outcome of both calculations and measurements is compared. (author)

  16. A model to simulate the dynamic of a PWR pressurizer using the CSMP program

    International Nuclear Information System (INIS)

    Woiski, E.R.

    1981-01-01

    A mathematical model has been developed to simulate the dynamic behavior of a PWR pressurizer using the CSMP program. A two-control-volume formulation non-equilibrium model has been used for this purpose. Thermodynamic states are obtained after each integration cycle. The code was tested against experimental results of Shippingport and NPD (Nuclear Power Demonstration Plant) pressurizers. It was also tested against available data from Angra I and Angra II/III safety analysis report. Despite the model simplicity, the lack of important data and the low reliability or the experimental curves, the calculated and experimental results compared well. (Author) [pt

  17. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author) [pt

  18. SARDAN- A program for the transients simulation in a typical PWR plant

    International Nuclear Information System (INIS)

    Mattos Santos, R.L.P. de.

    1979-10-01

    A program in FORTRAN-IV language was developed that simulates the behaviour of the primary circuit in a typical PWR plant during condition II transients, in particular uncontrolled withdrawal of a control rod set, control rod set drops and uncontrolled boron dilution. It the mathematical model adopted the reactor core, the hot piping to which a pressurizer is coupled, the steam generator and the cold piping are considered. The results obtained in the analysis of the mentioned accidents are compared to those present at the Final Safety Analysis Report (FSAR) of the Angra-1 reactor and are considered satisfactory. (F.E.) [pt

  19. Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-01-01

    Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

  20. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.

    1987-01-01

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author) [pt

  1. Study of the formation and transport of corrosion products in PWR primary circuit simulators

    International Nuclear Information System (INIS)

    Noe, M.; Frejaville, G.; Camp, J.J.

    1983-01-01

    The formation, migration and deposition of corrosion products in PWR primary circuits are studied in out-of-reactor loops. The aim of these studies is to limit the build-up of the radiation fields impinging on out-of-flux walls and to reduce the danger of rapid corrosion of fuel cans, taking into account the tougher conditions imposed on current trends in the operation of such industrial plants. Four simulator loops and their respective possibilities and research methods are described. (author)

  2. Design and simulation experimental study of bracket plates in steam generator for AC600 PWR

    International Nuclear Information System (INIS)

    Zhang Fuyuan; Zhang Wenqi; Ji Quankai; Zeng Xi; Xie Yongyao

    1998-01-01

    Seven-holes type bracket plate at the inlet nozzle and three-holes taper bracket plate at outlet nozzle are designed. According to 'local form and structure change' simulation theory, hydraulic models and simulators for the simulative experiments are designed. Taking water as the medium, the simulative experiments have been completed at the room temperature. The ζ-Re curves (here, ζ is the local pressure loss coefficient at the nozzles after the bracket plates are installed and Re is Reynolds number) have been got. Based on the experimental results, the computation and the analysis have been shown that. If the bracket plates are used in the steam generator (SG) of AC600 PWR, the pressure drop of primary side in the SG is about 14 percent higher than that of the 55/19 B style SG

  3. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  4. General model for Pc-based simulation of PWR and BWR plant components

    Energy Technology Data Exchange (ETDEWEB)

    Ratemi, W M; Abomustafa, A M [Faculty of enginnering, alfateh univerity Tripoli, (Libyan Arab Jamahiriya)

    1995-10-01

    In this paper, we present a basic mathematical model derived from physical principles to suit the simulation of PWR-components such as pressurizer, intact steam generator, ruptured steam generator, and the reactor component of a BWR-plant. In our development, we produced an NMMS-package for nuclear modular modelling simulation. Such package is installed on a personal computer and it is designed to be user friendly through color graphics windows interfacing. The package works under three environments, namely, pre-processor, simulation, and post-processor. Our analysis of results using cross graphing technique for steam generator tube rupture (SGTR) accident, yielded a new proposal for on-line monitoring of control strategy of SGTR-accident for nuclear or conventional power plant. 4 figs.

  5. PKL/K9, Refill and Reflood Experiment in a Simulated PWR Primary System (PKL)

    International Nuclear Information System (INIS)

    1981-01-01

    1 - Description of test facility: PKL-facility simulates the essential primary system components of a typical West German 1300 PWR with regard to their thermohydraulic behaviour. The facility essentially consists of the pressure vessel with the heated bundle, the downcomer simulator, the primary loops with the components steam generator and pump simulator, the injection devices, the break geometry simulator, as well as the separators connected thereto, and the test containment to maintain a back-pressure at the location of break which is expected to be typical for emergency conditions. The number of heater rods and the cross-sections of the testing plant are on a reduced scale 1:134 in comparison with a typical German PWR. The elevations and locations are essentially full scale. Pressure vessel: The space between the pressure vessel and the inner core casing is sealed from the core region and the upper and lower plenum and connected with the upper plenum only by a pressure equalization line. The rod bundle surrounded by the inner core casing consists of 340 rods, 337 of which are indirect electrically heated. The test bundle cross-section as well as a heater element with the measuring elevations, the original-KWU-spacers and the axial power profile (7 power steps) are described. Downcomer: The downcomer is simulated by the downcomer nozzle region and the downcomer U-tube. The cold leg injection takes place both directly in the downcomer nozzle region and in the lines of t he intact single and double loop near to the downcomer nozzle region. A cylindrical insertion and repulsing metal sheets are installed in the downcomer nozzle region in order to avoid the emergency injection points into the broken loop. 2 - Description of test: Test K 9 out of a series PKL-IB was conducted on May 30, 1979 by Kraftwerk Union (KWU) at Erlangen (Germany). The objective of the integral cold leg injection test K 9 (double-ended 200%-break) was to investigate after a LOCA the refill and

  6. VOF Simulations of Countercurrent Gas-Liquid Flow in a PWR Hot Leg

    Directory of Open Access Journals (Sweden)

    Michio Murase

    2012-12-01

    Full Text Available In order to evaluate flow patterns and CCFL (countercurrent flow limitation characteristics in a PWR hot leg under reflux condensation, numerical simulations have been done using a two-fluid model and a VOF (volume of fluid method implemented in the CFD software, FLUENT6.3.26. The two-fluid model gave good agreement with CCFL data under low pressure conditions but did not give good results under high pressure steam-water conditions. On the other hand, the VOF method gave good agreement with CCFL data for tests with a rectangular channel but did not give good results for calculations in a circular channel. Therefore, in this paper, the computational grid and schemes were improved in the VOF method, numerical simulations were done for steam-water flows at 1.5 MPa under PWR full-scale conditions with the diameter of 0.75 m, and the calculated results were compared with the UPTF data at 1.5 MPa. As a result, the calculated flow pattern was found to be similar to the flow pattern observed in small-scale air-water tests, and the calculated CCFL characteristics agreed well with the UPTF data at 1.5 MPa except in the region of a large steam volumetric flux.

  7. The determination of magnesium in simulated PWR coolant by graphite furnace atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Gatford, C.; Torrance, K.

    1988-06-01

    The determination of magnesium in simulated PWR coolant has been investigated by graphite furnace atomic absorption spectrometry with atomization from a L'vov platform. The presence of boric acid in the coolant suppresses the magnesium absorption to such an extent that removal of the boron is necessary and three variations of a methyl borate volatilization technique for the in situ removal of boron from the sample platform were investigated. This work has shown that dilution of the sample with an equal volume of acidified methanol and volatilization of the methyl borate was adequate for the determination of magnesium in coolant samples containing up to 2000 mg 1 -1 of boron. In simulated coolant samples containing 25 and 4 μg 1 -1 of magnesium, positive biases of about 2 and 0.5 μg 1 -1 were measured and these errors were considered to be due to contamination. The limit of detection in the presence of 100 and 2000 mg 1 -1 boron were 0.14 and 0.93 μg 1 -1 respectively. These performance characteristics suggest the method is completely acceptable for monitoring the chemical purity of PWR coolant and associated waters containing boric acid. If, however, more precise analyses were to be required for research purposes then any significant improvement in the above figures would require increased purity of reagents, clean-room conditions to reduce contamination and a more versatile atomic absorption spectrophotometer. (author)

  8. Investigating the cooling ability of reactor vessel head injection in the Maanshan PWR using CFD simulation

    International Nuclear Information System (INIS)

    Tseng Yungshin; Lin Chihhung; Wan Jongrong; Shih Chunkuan; Tsai, F. Peter

    2011-01-01

    In order to reduce the crack growth rate on the welding of penetration pipe, Pressurized Water Reactor (PWR) of Maanshan nuclear power plant (NPP) uses vessel head injection to cool vessel lid and control rod driving components. The injection flow from the cold leg is drained by the pressure difference between cold leg and upper internal components. In this study, 10 million meshes model with 4 sub-models have been developed to simulate the thermal-hydraulic behavior by commercial CFD program FLUENT. The results indicate that the injection nozzles can provide good cooling ability to reduce the maximum temperature for lid on the vessel head. The maximum temperature of vessel lid is about 293.81degC. Based on the simulated temperature, ASME CODE N-729-1 was further used to recount the effective degradation years (EDY) and reinspection years (RIY) factors. It demonstrates that the EDY and RIY factors are still less than 1.0. Therefore, the re-inspection period for Maanshan PWR would not be significantly affected by the miner temperature difference. (author)

  9. A digital simulation of a pressurizer in a PWR nuclear power plant

    International Nuclear Information System (INIS)

    Sato, E.F.

    1980-11-01

    A model for pressurizer digital simulation of a PWR nuclear power plant during transients, considering all pressurizer control features, is presented. The pressurizer is divided into two regions separated by a water-vapor interface and non-equilibrium conditions are considered. The particular thermodynamic process followed during insurge and outsurges is determined at each instant of analysis without any previous assumption. The pressure behavior is defined by an explicit equation in any of four possible pressurizer thermodynamic conditions. Thermodynamic properties of steam and water are computed by ASME subroutines and the mathematical formulation presented in this study was programed in FORTRAN IV for a Burroughs-6700 digital computer system. This program was employed to simulate the Shippingport Atomic Power Station and Almirante Alvaro Alberto Nuclear Power Plant - Unit 1 pressurizers. The test results compared with experimental or vendor data show the validity of this analysis method. (Author) [pt

  10. Simulating the steam generator and the pressurizer of a PWR nuclear power plant

    International Nuclear Information System (INIS)

    De Greef, J.F.

    1985-01-01

    In a PWR nuclear power plant, considered as a power generating device, the steam generator as a subset plays an important role in the generation process, whereas the pressurizer rather acts as a control device for security purposes. Nevertheless, from a thermodynamical point of view, the two subsets behave basically in the same way, so that a common set of basic equations may be suggested to develop for each the proper mathematical simulation model. In this paper the generation of this common set of basic equations is described, from which a specific model for each device is derived. A numerical illustration of the behaviour of the two devices for typical inputs to the derived simulation model is pictured. (author)

  11. Laboratory results gained from cold worked type 316Ti under simulated PWR primary environment

    International Nuclear Information System (INIS)

    Devrient, B.; Kilian, R.; Koenig, G.; Widera, M.; Wermelinger, T.

    2015-01-01

    Beginning in 2005, intergranular stress corrosion cracking (IGSCC) of barrel bolts made from cold worked type 316Ti (German Material No. 1.4571 K) was observed in several S/KWU type PWRs. This mechanism was so far less understood for PWR primary conditions. Therefore an extended joint research program was launched by AREVA GmbH and VGB e.V. to clarify the specific conditions which contributed to the observed findings on barrel bolts. In the frame of this research program beneath the evaluation of the operational experience also laboratory tests on the general cracking behavior of cold worked type 316Ti material, which followed the same production line as for barrel bolt manufacturing in the eighties, with different cold work levels covering up to 30 % were performed to determine whether there is a specific susceptibility of cold worked austenitic stainless steel specimens to suffer IGSCC under simulated PWR primary conditions. All these slow strain rate tests on tapered specimens and component specimens came to the results that first, much higher cold work levels than used for the existing barrel bolts are needed for IGSCC initiation. Secondly, additional high active plastic deformation is needed to generate and propagate intergranular cracking. And thirdly, all specimens finally showed ductile fracture at the applied strain rates. (authors)

  12. An homogeneous model of steam generator to simulate operational transiento and accidents in PWR nuclear power plants

    International Nuclear Information System (INIS)

    Souza, A.L. de.

    1981-07-01

    GEVAP - A digital computer code was developed to simulate the thermodynamic transient behaviour of steam generators. The steam generator is divided in heating sections. In each section, the conservation equations of mass and energy are integrated numerically, using a predictor-corrector method. As good reslts where obtained, as compared to transients simulated using more detainled codes, it is concluded that GEVAP can be included as the steam generator module of a more complete systems simulation code for PWR's. (E.G.) [pt

  13. Simulation of nonlinear dynamics of a PWR core by an improved lumped formulation for fuel heat transfer

    International Nuclear Information System (INIS)

    Su, Jian; Cotta, Renato M.

    2000-01-01

    In this work, thermohydraulic behaviour of PWR, during reactivity insertion and partial loss-of-flow, is simulated by using a simplified mathematical model of reactor core and primary coolant. An improved lumped parameter formulation for transient heat conduction in fuel rod is used for core heat transfer modelling. Transient temperature response of fuel, cladding and coolant is analysed. (author)

  14. Low cycle fatigue behavior of hot-bent 347 stainless steel in a simulated PWR water environment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jun Ho; Seo, Myung Gyu; Jang, Chang Heui [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Hong, Jong Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Tae Soon [Central Research InstituteKorea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

    2016-11-15

    The effect of hot bending on the Low cycle fatigue (LCF) behavior of 347 SS was evaluated in Room temperature (RT) air and simulated Pressurized water reactor (PWR) water environments. The LCF life of 347 SS in PWR water was shorter than that in RT air for the as-received and hot-bent conditions. The LCF life of hot-bent 347 SS was relatively longer than that of the as-received condition in both RT air and PWR water. Microstructure analysis indicated development of dislocation structure near niobium carbide particles and increase in dislocation density for the hot-bent 347 SS. Such microstructure acted as barriers to dislocation movement during the LCF test, resulting in minimal hardening for the hot-bent 347 SS in RT air.

  15. Simulation of a Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos.

    1980-01-01

    The following work intends to perform the digital simulation, of the Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant for control systems design and analysis purposes. There are mathematical models for the reactor, the steam generator, the pressurizer and for transport lags of the coolant in the primary circuit. Nevertheless no one control system has been considered to permit any user the inclusion in the more convenient way of the desired control systems' models. The characteristics of the system in consideration are fundamentally equal to the ones of Almirante Alvaro Alberto Nuclear Power Plant, Unit I (Angra I) obtained in the Final Safety Analysis Report at Comissao Nacional de Energia Nuclear. (author)

  16. Design and static simulation of secondary loop of small PWR nuclear power plants

    International Nuclear Information System (INIS)

    Martin Lopez, L.A.N.

    1989-01-01

    A computer program that has been developed with the purpose of making easier the decisions concerning the design of the secondary loop of small PWR nuclear power plants through numerical experiments of low running costs and short time is presented. Initially, the first part of the computer program is described. It aims to preliminarily design several major components of the secondary circuit from user-defined design conditions. Next, the second part of the computer program is presented. It simulates the steady state operation at part-load conditions of the preliminary design of the plant by generating and solving systems of simultaneous nonlinear algebraic equations, their number varying from 17 to 107. The computer program has been tested for several application cases. The program results are discussed in the last part of the work, along with several aspects to be added to the program in future works. (author)

  17. HEXBU-3D, a three-dimensional PWR-simulator program for hexagonal fuel assemblies

    International Nuclear Information System (INIS)

    Karvinen, E.

    1981-06-01

    HEXBU-3D is a three-dimensional nodal simulator program for PWR reactors. It is designed for a reactor core that consists of hexagonal fuel assemblies and of big follower-type control assemblies. The program solves two-group diffusion equations in homogenized fuel assembly geometry by a sophisticated nodal method. The treatment of feedback effects from xenon-poisoning, fuel temperature, moderator temperature and density and soluble boron concentration are included in the program. The nodal equations are solved by a fast two-level iteration technique and the eigenvalue can be either the effective multiplication factor or the boron concentration of the moderator. Burnup calculations are performed by tabulated sets of burnup-dependent cross sections evaluated by a cell burnup program. HEXBY-3D has been originally programmed in FORTRAN V for the UNIVAC 1108 computer, but there is also another version which is operable on the CDC CYBER 170 computer. (author)

  18. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

    International Nuclear Information System (INIS)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

    2011-01-01

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  19. An axial calculation method for accurate two-dimensional PWR core simulation

    International Nuclear Information System (INIS)

    Grimm, P.

    1985-02-01

    An axial calculation method, which improves the agreement of the multiplication factors determined by two- and three-dimensional PWR neutronic calculations, is presented. The axial buckling is determined at each time point so as to reproduce the increase of the leakage due to the flattening of the axial power distribution and the effect of the axial variation of the group constants of the fuel on the reactivity is taken into account. The results of a test example show that the differences of k-eff and cycle length between two- and three-dimensional calculations, which are unsatisfactorily large if a constant buckling is used, become negligible if the results of the axial calculation are used in the two-dimensional core simulation. (Auth.)

  20. TRSM-a thermal-hydraulic real-time simulation model for PWR

    International Nuclear Information System (INIS)

    Zhou Weichang

    1997-01-01

    TRSM (a Thermal-hydraulic Real-time Simulation Model) has been developed for PWR real-time simulation and best-estimate prediction of normal operating and abnormal accident conditions. It is a non-equilibrium two phase flow thermal-hydraulic model based on five basic conservation equations. A drift flux model is used to account for the unequal velocities of liquid and gaseous mixture, with or without the presence of the noncondensibles. Critical flow models are applied for break flow and valve flow calculations. A 5-regime two phase heat convection model is applied for clad-to-coolant as well as fluid-to-tubing heat transfer. A rigorous reactor coolant pump model is used to calculate the pressure drop and rise for the suction and discharge ends with complete pump characteristics curves included. The TRSM model has been adapted in the full-scale training simulator of Qinshan Nuclear Power Plant 300 MW unit to simulate the thermal-hydraulic performance of the NSSS. The simulation results of a cold leg LOCA and a steam generator tube rupture (SGTR) accident are presented

  1. Test requirements for the integral effect test to simulate Korean PWR plants

    Energy Technology Data Exchange (ETDEWEB)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  2. Test requirements for the integral effect test to simulate Korean PWR plants

    International Nuclear Information System (INIS)

    Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K.

    2001-02-01

    In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

  3. A numerical integration approach suitable for simulating PWR dynamics using a microcomputer system

    International Nuclear Information System (INIS)

    Zhiwei, L.; Kerlin, T.W.

    1983-01-01

    It is attractive to use microcomputer systems to simulate nuclear power plant dynamics for the purpose of teaching and/or control system design. An analysis and a comparison of feasibility of existing numerical integration methods have been made. The criteria for choosing the integration step using various numerical integration methods including the matrix exponential method are derived. In order to speed up the simulation, an approach is presented using the Newton recursion calculus which can avoid convergence limitations in choosing the integration step size. The accuracy consideration will dominate the integration step limited. The advantages of this method have been demonstrated through a case study using CBM model 8032 microcomputer to simulate a reduced order linear PWR model under various perturbations. It has been proven theoretically and practically that the Runge-Kutta method and Adams-Moulton method are not feasible. The matrix exponential method is good at accuracy and fairly good at speed. The Newton recursion method can save 3/4 to 4/5 time compared to the matrix exponential method with reasonable accuracy. Vertical Barhis method can be expanded to deal with nonlinear nuclear power plant models and higher order models as well

  4. Vertical profiles for SO2 and SO on Venus from different one-dimensional simulations

    Science.gov (United States)

    Mills, Franklin P.; Jessup, Kandis-Lea; Yung, Yuk

    2017-10-01

    Sulfur dioxide (SO2) plays many roles in Venus’ atmosphere. It is a precursor for the sulfuric acid that condenses to form the global cloud layers and is likely a precursor for the unidentified UV absorber, which, along with CO2 near the tops of the clouds, appears to be responsible for absorbing about half of the energy deposited in Venus’ atmosphere [1]. Most published simulations of Venus’ mesospheric chemistry have used one-dimensional numerical models intended to represent global-average or diurnal-average conditions [eg, 2, 3, 4]. Observations, however, have found significant variations of SO and SO2 with latitude and local time throughout the mesosphere [eg, 5, 6]. Some recent simulations have examined local time variations of SO and SO2 using analytical models [5], one-dimensional steady-state solar-zenith-angle-dependent numerical models [6], and three-dimensional general circulation models (GCMs) [7]. As an initial step towards a quantitative comparison among these different types of models, this poster compares simulated SO, SO2, and SO/SO2 from global-average, diurnal-average, and solar-zenith-angle (SZA) dependent steady-state models for the mesosphere.The Caltech/JPL photochemical model [8] was used with vertical transport via eddy diffusion set based on observations and observationally-defined lower boundary conditions for HCl, CO, and OCS. Solar fluxes are based on SORCE SOLSTICE and SORCE SIM measurements from 26 December 2010 [9, 10]. The results indicate global-average and diurnal-average models may have significant limitations when used to interpret latitude- and local-time-dependent observations of SO2 and SO.[1] Titov D et al (2007) in Exploring Venus as a Terrestrial Planet, 121-138. [2] Zhang X et al (2012) Icarus, 217, 714-739. [3] Krasnopolsky V A (2012) Icarus, 218, 230-246. [4] Parkinson C D et al (2015) Planet Space Sci, 113-114, 226-236. [5] Sandor B J et al (2010) Icarus, 208, 49-60. [6] Jessup K-L et al (2015) Icarus, 258, 309

  5. Venus magnetosphere

    International Nuclear Information System (INIS)

    Podgornyj, I.M.

    1983-01-01

    Some peculiarities of the structure of the Venus magnetosphere are considered. A Swedish scientist H. Alfven supposes that nebular bodies with ionospheric shelles of the type of Venus atmosphere possess induced magnetospheres with dragged magnetic tails. In the Institute of Space Research of the USSR Academy of Sciences experiments on the modelling of such magnetosphere are performed. The possibility of formation of the shock wave in the body with plasma shell in the absence of the proper magnetic shell is proved. The cosmic ''Pioneer-Venus'' equipment is used to obtain such a distribution of the magnetic field depending on the distance to Venus as it was predicted by the laboratory model

  6. Simulation of a Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant. Simulacao do sistema nuclear de geracao de vapor de uma central PWR

    Energy Technology Data Exchange (ETDEWEB)

    Reis Martins Junior, L.L. dos.

    1980-01-01

    The following work intends to perform the digital simulation, of the Nuclear Steam Supply System (NSSS) of a PWR nuclear power plant for control systems design and analysis purposes. There are mathematical models for the reactor, the steam generator, the pressurizer and for transport lags of the coolant in the primary circuit. Nevertheless no one control system has been considered to permit any user the inclusion in the more convenient way of the desired control systems' models. The characteristics of the system in consideration are fundamentally equal to the ones of Almirante Alvaro Alberto Nuclear Power Plant, Unit I (Angra I) obtained in the Final Safety Analysis Report at Comissao Nacional de Energia Nuclear. (author).

  7. The VENUS detector at TRISTAN

    International Nuclear Information System (INIS)

    Sugimoto, Shojiro

    1983-01-01

    The design of the VENUS detector is described. In this paper, emphasis is placed on the central tracking chamber and the electromagnetic shower calorimeters. Referring to computer simulations and test measurements with prototypes, the expected performance of our detector system is discussed. The contents are, for the most part, taken from the VENUS proposal /2/. (author)

  8. Simulation of steam generator plugging tubes in a PWR to analyze the operating impact

    Energy Technology Data Exchange (ETDEWEB)

    Pla, Patricia, E-mail: patricia.pla-freixa@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands); Reventos, Francesc, E-mail: francesc.reventos@upc.edu [Technical University of Catalonia (UPC), Barcelona (Spain); Martin Ramos, Manuel, E-mail: manuel.martin-ramos@ec.europa.eu [Nuclear Safety and Security Coordination Unit, Policy Support Coordination, Joint Research Centre of the European Commission, Brussels (Belgium); Sol, Ismael, E-mail: isol@anacnv.com [Asociación Nuclear Ascó-Vandellós-II (ANAV), Tarragona (Spain); Strucic, Miodrag, E-mail: miodrag.strucic@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands)

    2016-08-15

    Highlights: • Plugging a fraction of the SG tubes does not affect power output of the plant. • There is a limit to SG plugging in the range of 10–15%. • The rupture of a SG tube in a 12% plugged SG has shown no significant differences in operator actions. • A SBLOCA in a 12% plugged SG has shown no significant differences in operator actions. - Abstract: A number of nuclear power plants (NPPs) with pressurized water reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems. Several methods were attempted to correct the defects of the tubes, but eventually the only permanent solution was to plug them. The consequences of plugging the tubes are the decrease of heat transfer surface, the reduction of the flow area and subsequent reduction of the primary system mass flow and for a fraction of plugged tubes higher than a given value, the reduction of reactor output and economic losses. The objective of this paper is to analyze whether steam generator tube plugging has an impact in the effectiveness of accident management actions. An analysis with Relap5 Mod 3.3 patch03 for the Spanish reactor Ascó-2, a 3-loop 2940.6 MWth Westinghouse PWR, in which plugging of steam generator tubes are simulated, is presented in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. To complete the study two events, in which the steam generators are used to cooldown the plant, were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the emergency operating

  9. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water

    Science.gov (United States)

    Terachi, T.; Yamada, T.; Miyamoto, T.; Arioka, K.

    2012-07-01

    The rates of SCC growth were measured under simulated PWR primary water conditions (500 ppm B + 2 ppm Li + 30 cm3/kg-H2O-STP DH2) using cold worked 316SS and 304SS. The direct current potential drop method was applied to measure the crack growth rates for 53 specimens. Dependence of the major engineering factors, such as yield strength, temperature and stress intensity was systematically examined. The rates of crack growth were proportional to the 2.9 power of yield strength, and directly proportional to the apparent yield strength. The estimated apparent activation energy was 84 kJ/mol. No significant differences in the SCC growth rates and behaviors were identified between 316SS and 304SS. Based on the measured results, an empirical equation for crack growth rate was proposed for engineering applications. Although there were deviations, 92.8% of the measured crack growth rates did not exceed twice the value calculated by the empirical equation.

  10. Precursor evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water

    Energy Technology Data Exchange (ETDEWEB)

    Zhai, Ziqing; Kruska, Karen; Toloczko, Mychailo B.; Bruemmer, Stephen M.

    2017-03-27

    Stress corrosion crack initiation of two thermally-treated, cold-worked (CW) alloy 690 materials was investigated in 360oC simulated PWR primary water using constant load tensile (CLT) tests and blunt notch compact tension (BNCT) tests equipped with direct current potential drop (DCPD) for in-situ detection of cracking. SCC initiation was not detected by DCPD for the 21% and 31%CW CLT specimens loaded at their yield stress after ~9,220 h, however intergranular (IG) precursor damage and isolated surface cracks were observed on the specimens. The two 31%CW BNCT specimens loaded at moderate stress intensity after several cyclic loading ramps showed DCPD-indicated crack initiation after 10,400h exposure at constant stress intensity, which resulted from significant growth of IG cracks. The 21%CW BNCT specimens only exhibited isolated small IG surface cracks and showed no apparent DCPD change throughout the test. Interestingly, post-test cross-section examinations revealed many grain boundary (GB) nano-cavities in the bulk of all the CLT and BNCT specimens particularly for the 31%CW materials. Cavities were also found along GBs extending to the surface suggesting an important role in crack nucleation. This paper provides an overview of the evolution of GB cavities and will discuss their effects on crack initiation in CW alloy 690.

  11. Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

    International Nuclear Information System (INIS)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

    2012-01-01

    The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

  12. Monte Carlo Simulation of Quantitative Electron Probe Microanalysis of the PWR Spent Fuel with a Pt Coating

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The PWR spent fuel sample should be coated with conducting material in order to provide a path for electrons and to prevent charging. Generally, the ZAF method has been used for quantitative electron probe microanalysis of conducting samples. However, the coated samples are not applicable for the ZAF method. Probe current, primary electron energy and x-ray produced by the primary beam are attenuated within the coating films. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program [2] to evaluate the x-ray attenuation within the Pt coating films. The target samples are the PWR spent fuels with 50 GWd/tU of burnup , 6 years of cooling time and a Pt coating film (3, 5, 7, 10 and 15 nm thickness)

  13. Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

    International Nuclear Information System (INIS)

    Mansouri, S.

    1985-06-01

    This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

  14. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    International Nuclear Information System (INIS)

    Mehboob, Khurram; Xinrong, Cao; Ahmed, Raheel; Ali, Majid

    2013-01-01

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value

  15. Fatigue crack growth threshold of austenitic stainless steels in simulated PWR primary water

    International Nuclear Information System (INIS)

    Tsutsumi, Kazuya; Yamamoto, Kenji; Nitta, Yoshikazu

    2007-01-01

    Many studies have revealed that fatigue crack growth (FCG) rate of austenitic stainless steels is accelerated in light water reactor environment compared to that in air at room temperature. Major driving factors in the acceleration of FCG rate are stress ratio, temperature and stress rise time. Based on this knowledge, FCG curves have been developed considering these factors as parameters. However, there are few data of FCG threshold ΔK th in light water reactor environment. Hence it is necessary to clarify FCG rate under near-threshold condition for more accurate evaluation of fatigue crack growth behavior under cyclic stress with relatively low ΔK. In the present study, therefore, ΔK th was determined for austenitic stainless steels in simulated PWR primary water, and FCG behavior under near-threshold condition was revealed by collecting fatigue crack propagation data. The results are summarized as follows: No propagation of fatigue crack was found in high temperature water, and there was a definite ΔK th . Average ΔK eff,th was 4.3 MPa·m 0.5 at 325degC, 3.3 MPa·m 0.5 at 100degC, and there was no considerable reduction compared to currently known ΔK eff,th in air. Thus, it was revealed tha ambient conditions had minimal effect, on ΔK eff,th , ΔK th increases with increasing temperature and decreasing frequency. As a result of fracture surface observation, oxide-induced-crack-closure was considered to be a cause of the dependency described above. In addition, it was suggested that changes in material properties also had influence on ΔK th, since ΔK eff,th itself increased at elevated temperature. (author)

  16. Best-estimate LOCA simulation in a PWR-W containment building with a detailed 3D GOTHIC model

    International Nuclear Information System (INIS)

    Jimenez, G.; Fernandez-Cosials, K.; Bocanegra, R.; Lopez-Alonso, E.

    2015-01-01

    The design-basis accidents in a PWR-W containment building are usually simulated with a lumped parameter model, normally used for license analysis. Nevertheless, some phenomenology is difficult to be simulated with a lumped model: the condensation rate in each structure, stagnant water areas, temperature in different compartments, sumps and recirculation pumps disabled because of lack of water, etc. Therefore, for the detailed study of the thermal-hydraulic (TH) behaviour in every room of the containment building could be more appropriate to do it with a detailed 3D representation of the containment building geometry. The main objective of this project has been to build a 3D PWR-W containment model with the GOTHIC code to analyze the detailed behavior during a design basis accident. In the process of the 3D GOTHIC model development some previous steps were necessary: a detailed CAD model of the containment, followed by a simplified model adapted to the GOTHIC geometric capabilities. Once the geometry has been adapted to the GOTHIC requirements, the 3D model is created with this information. A design-basis accident has been simulated with the 3D model (LBLOCA), and the local TH behaviour is analysed. The results show that in comparison with a lumped parameter model, high temperatures are reached locally. Nevertheless the average pressure behaviour is found to be similar to that given by a lumped parameter model. The present paper demonstrates that is possible to build a 3D PWR-W model with the GOTHIC code with enough resolution to analyse the TH behaviour in each one of the containment rooms but at the same time with reasonable computing time. Once the GOTHIC model has been created a new road is opened enabling the simulation of other accidents such as MSLB, a SBLOCA or even a long-term SBO sequence. This document is made up of an abstract and the slides of the presentation. (authors)

  17. A simple analytical scaling method for a scaled-down test facility simulating SB-LOCAs in a passive PWR

    International Nuclear Information System (INIS)

    Lee, Sang Il

    1992-02-01

    A Simple analytical scaling method is developed for a scaled-down test facility simulating SB-LOCAs in a passive PWR. The whole scenario of a SB-LOCA is divided into two phases on the basis of the pressure trend ; depressurization phase and pot-boiling phase. The pressure and the core mixture level are selected as the most critical parameters to be preserved between the prototype and the scaled-down model. In each phase the high important phenomena having the influence on the critical parameters are identified and the scaling parameters governing the high important phenomena are generated by the present method. To validate the model used, Marviken CFT and 336 rod bundle experiment are simulated. The models overpredict both the pressure and two phase mixture level, but it shows agreement at least qualitatively with experimental results. In order to validate whether the scaled-down model well represents the important phenomena, we simulate the nondimensional pressure response of a cold-leg 4-inch break transient for AP-600 and the scaled-down model. The results of the present method are in excellent agreement with those of AP-600. It can be concluded that the present method is suitable for scaling the test facility simulating SB-LOCAs in a passive PWR

  18. A preliminary evaluation of the simulation requirements of accident situations in a PWR training simulator

    International Nuclear Information System (INIS)

    Bhattacharya, A.S.

    1978-07-01

    An attempt was made to establish here the need to incorporate ''Accident Situations'' in Training Simulators, and, to indicate that the modelling of such otherwise complex situations could admit a great deal of simplifications. In most cases, a ''semidynamic'' model to display time dependent flows or Tank levels would do while ESF actuated start/stop or open/close of equipment could be modelled merely ''Live''. Transport delays and Time Constants in most cases are of no significance as are the accuracy considerations of most variables. The systems which need complete Dynamic Simulation are the ones which are always running, such as Pressuriser, RCS, Reactor, and the Steam Generator with the added advantages of the modest accuracy requirements and the plant being shutdown. (author)

  19. An investigation into the efficiency of ion-exchange membranes in simulated PWR coolants

    International Nuclear Information System (INIS)

    Clune, T.

    1980-11-01

    This report describes an investigation of the retention efficiency of cation-exchange membranes for magnesium, calcium and nickel ions in PWR-coolant type solutions containing 2 ppm lithium (as lithium hydroxide) and 1000 ppm boron (as boric acid). By analysis of the membranes themselves or of the effluent, the retention characteristics of the membranes in various experimental conditions have been examined. (author)

  20. Measurement of local flow pattern in boiling R12 simulating PWR conditions with multiple optical probes

    International Nuclear Information System (INIS)

    Garnier, J.

    1998-01-01

    center of particles is presented. Moreover, some numerical simulations gives the uncertainty induced by this treatment method. We quantify the uncertainty on gas velocity and on granulometric properties when, at the measurement point, there is a velocity gradient, a gradient of particles number and a gradient or particles diameter. Then, we detail the data acquisition system which measures the frequency of particles intercepted by the probe, the histogram of time of flight between the two fibers and the histogram of vapor time of the first P.I.F. Moreover, a numerical scope acquires the raw signal to check that the probe works properly. Finally, we present some of the results we obtained. A wide range of thermal hydraulic parameters has been covered (more than 10000 measurements have been performed). We first show that we can get a criteria to check that our particles have a spherical shape. We also find that rise and fall time of the electrical signal are strongly correlated with the gas velocity and we expect it will be possible to perform velocity measurements with a single fiber probe (after a specific calibration which is under definition). The concluding remarks deals with the future developments. A first work concern probes able to measure in the real conditions of a PWR (16 MPa and 340 deg. C). Another development concerns the size of the sensitive part for probes with classical optical fibres. We intend to get nanometric tips to minimize flow disturbance, increase accuracy for small particles measurements and get multiple probes (four fibres) with low dimensions for local measurements of interfacial area density. We also intend to extend the capability of our data acquisition system in order to keep more information on the P.I.F. (author)

  1. Meeting Venus

    Science.gov (United States)

    Sterken, Christiaan; Aspaas, Per Pippin

    2013-06-01

    On 2-3 June 2012, the University of Tromsoe hosted a conference about the cultural and scientific history of the transits of Venus. The conference took place in Tromsoe for two very specific reasons. First and foremost, the last transit of Venus of this century lent itself to be observed on the disc of the Midnight Sun in this part of Europe during the night of 5 to 6 June 2012. Second, several Venus transit expeditions in this region were central in the global enterprise of measuring the scale of the solar system in the eighteenth century. The site of the conference was the Nordnorsk Vitensenter (Science Centre of Northern Norway), which is located at the campus of the University of Tromsoe. After the conference, participants were invited to either stay in Tromsoe until the midnight of 5-6 June, or take part in a Venus transit voyage in Finnmark, during which the historical sites Vardoe, Hammerfest, and the North Cape were to be visited. The post-conference program culminated with the participants observing the transit of Venus in or near Tromsoe, Vardoe and even from a plane near Alta. These Proceedings contain a selection of the lectures delivered on 2-3 June 2012, and also a narrative description of the transit viewing from Tromsoe, Vardoe and Alta. The title of the book, Meeting Venus, refers the title of a play by the Hungarian film director, screenwriter and opera director Istvan Szabo (1938-). The autobiographical movie Meeting Venus (1991) directed by him is based on his experience directing Tannhauser at the Paris Opera in 1984. The movie brings the story of an imaginary international opera company that encounters a never ending series of difficulties and pitfalls that symbolise the challenges of any multicultural and international endeavour. As is evident from the many papers presented in this book, Meeting Venus not only contains the epic tales of the transits of the seventeenth, eighteenth and nineteenth centuries, it also covers the conference

  2. Three-Dimensional Structures of Thermal Tides Simulated by a Venus GCM

    Science.gov (United States)

    Takagi, Masahiro; Sugimoto, Norihiko; Ando, Hiroki; Matsuda, Yoshihisa

    2018-02-01

    Thermal tides in the Venus atmosphere are investigated by using a GCM named as AFES-Venus. The three-dimensional structures of wind and temperature associated with the thermal tides obtained in our model are fully examined and compared with observations. The result shows that the wind and temperature distributions of the thermal tides depend complexly on latitude and altitude in the cloud layer, mainly because they consist of vertically propagating and trapped modes with zonal wave numbers of 1-4, each of which predominates in different latitudes and altitudes under the influence of mid- and high-latitude jets. A strong circulation between the subsolar and antisolar (SS-AS) points, which is equivalent to a diurnal component of the thermal tides, is superposed on the superrotation. The vertical velocity of SS-AS circulation is about 10 times larger than that of the zonal-mean meridional circulation (ZMMC) in 60-70 km altitudes. It is suggested that the SS-AS circulation could contribute to the material transport, and its upward motion might be related to the UV dark region observed in the subsolar and early afternoon regions in low latitudes. The terdiurnal and quaterdiurnal tides, which may be excited by the nonlinear interactions among the diurnal and semidiurnal tides in middle and high latitudes, are detected in the solar-fixed Y-shape structure formed in the vertical wind field in the upper cloud layer. The ZMMC is weak and has a complex structure in the cloud layer; the Hadley circulation is confined to latitudes equatorward of 30°, and the Ferrel-like one appears in middle and high latitudes.

  3. Operating function tests of the PWR type RHR pump for engineering safety system under simulated strong ground excitation

    International Nuclear Information System (INIS)

    Uga, Takeo; Shiraki, Kazuhiro; Homma, Toshiaki; Inazuka, Hisashi; Nakajima, Norifumi.

    1979-08-01

    Results are described of operating function verification tests of a PWR RHR pump during an earthquake. Of the active reactor components, the PWR residual heat removal pump was chosen from view points of aseismic classification, safety function, structural complexity and past aseismic tests. Through survey of the service conditions and structure of this pump, seismic test conditions such as acceleration level, simulated seismic wave form and earthquake duration were decided for seismicity of the operating pump. Then, plans were prepared to evaluate vibration chracteristics of the pump and to estimate its aseismic design margins. Subsequently, test facility and instrumentation system were designed and constructed. Experimental results could thus be acquired on vibration characteristics of the pump and its dynamic behavior during different kinds and levels of simulated earthquake. In conclusion: (1) Stiffeners attached to the auxiliary system piping do improve aseismic performance of the pump. (2) The rotor-shaft-bearing system is secure unless it is subjected to transient disturbunces having high frequency content. (3) The motor and pump casing having resonance frequencies much higher than frequency content of the seismic wave show only small amplifications. (4) The RHR pump possesses an aseismic design margin more than 2.6 times the expected ultimate earthquake on design basis. (author)

  4. Simulation of corrosion product activity in ion- exchanger of PWR under acceleration of corrosion and flow rate perturbations

    International Nuclear Information System (INIS)

    Mirza, N.M.; Mirza, S.M.; Rafique, M.

    2005-01-01

    In this paper computer code developed earlier by the authors (CPAIR-P) has been employed to compute corrosion product activity in PWRs for flow rate perturbations. The values of radioactivity in ion exchanger of Pressurized Water Reactor (PWR) under normal and flow rate perturbation conditions have been calculated. For linearly accelerating corrosion rates, activity saturates for removal rate of 600 cm/sup 3// s in primary coolant of PWR. A higher removal rate of 750 cm/sup 3// s was selected for which the saturation value is sufficiently low (0. 28 micro Ci/cm/sup 3/). Simulation results shows that the Fe/sup 59/ Na/sup 24/, Mo/sup 99/, Mn/sup 56/ reaches saturation values with in about 700 hours of reactor operation. However, Co/sup 58/ and Co/sup 60/ keep on accumulating and do not saturate with in 2000 hours of these simulation time. When flow rate is decreased by 10% of rated flow rate after 500 hours of reactor operation, a dip in activity is seen, which reaches to the value of 0.00138 micro Ci cm/sup -3/ then again it begins to rise and reaches saturation value of 0.00147 cm/sup 3//s. (author)

  5. Helped positioning by using a simulation tool for qualification of PWR vessel examination technique

    International Nuclear Information System (INIS)

    Lasserre, Frederic; Pasquier, Thierry; Haiat, Guillaume; Calmon, Pierre; Leberre, Stephane; Lutsen, Mickael

    2006-01-01

    INTERCONTROLE have been performing the examination of all PWR vessels in France from the inside, using UT techniques since 1975. The in-service inspection machine (MIS) features several tools equipped with focussed transducers; each tool is dedicated to one specific area of the vessel. In the core region, the very first millimeters from the cladding-base metal interface has to be inspected with accuracy because of the under-cladding cracks type defects (perpendicular to the inner surface) likely to be found. The technique used up to now was qualified according to the RSE-M code in 1998. It is based on a set of 63 angle L-waves transducers specifically designed for the detection of defect tip diffraction echoes in the 25 first millimeters in through-wall thickness. The analysis methods for defect characterization are based on a global integration of various cladding induced phenomena. The technique, the procedure and the analysis methods were qualified for a given limited volume. The new qualification in process in France, requires that INTERCONTROLE find solutions for increasing the accuracy of the analysis, in a larger qualification volume than before, while remaining in close compliance with the RSE-M code. A new computer assisted analysis tool for the characterization, the sizing and the positioning of defects is part of the improvements currently in progress or already completed. This tool is the result of a thesis commissioned to the CEA (Atomic Energy Commission), now implemented in the CIVAMIS software (developed on a CIVA based system). The updated version of CIVAMIS including this characterization tool and the RSE-M qualification of the new analysis method (with validation on mock-ups) is now qualified. Despite of a larger qualification volume, the results obtained (mentioned in the present paper) fulfill the customer's requirements thanks to the amount of data, of information and of knowledge, available today. The ability to simulate the cladding in terms

  6. Simulation of the fuel rod thermal hydraulic performance during the blow down phase in a PWR

    International Nuclear Information System (INIS)

    Gadelha, J.A.M.

    1982-10-01

    A digital computer code to predict the fuel rod thermalhydraulic performance during a postulated loss-of-coolant accident (LOCA) in the primary circuit of a PWR nuclear power plant is developed. The fuel rod corresponds to that in an average channel in the core. Only the blowdown phase is considered during the accident. The conservation equations of mass, momentum, and energy, and the heat conduction equation are solved to determine the fuel rod conditions during the accident. Finite differences are applied as a numerical method in the solution of the equations modelling the rod and coolant conditions. (Author) [pt

  7. Design of a steam generator for PWR power plants and steady state simulation

    International Nuclear Information System (INIS)

    Ferreira, W.J.

    1982-01-01

    A procedure and a computer code for the thermal design of a steam generator for PWR power plants is developed. A vertical integral steam generator with inverted U-tubes and natural circulation of the secondary side is selected for modelling. Primary fluid velocity and recirculation ratio are varied to obtain the preliminary dimensions. Further, adjustments are made through iteractive solution of the equations of conservation of mass, energy and momentum. An agreement is found between design calculations for steam generators of different capacities and existing designs. (Author) [pt

  8. Uncertainty and sensitivity analysis in the neutronic parameters generation for BWR and PWR coupled thermal-hydraulic–neutronic simulations

    International Nuclear Information System (INIS)

    Ánchel, F.; Barrachina, T.; Miró, R.; Verdú, G.; Juanas, J.; Macián-Juan, R.

    2012-01-01

    Highlights: ► Best-estimate codes are affected by the uncertainty in the methods and the models. ► Influence of the uncertainty in the macroscopic cross-sections in a BWR and PWR RIA accidents analysis. ► The fast diffusion coefficient, the scattering cross section and both fission cross sections are the most influential factors. ► The absorption cross sections very little influence. ► Using a normal pdf the results are more “conservative” comparing the power peak reached with uncertainty quantified with a uniform pdf. - Abstract: The Best Estimate analysis consists of a coupled thermal-hydraulic and neutronic description of the nuclear system's behavior; uncertainties from both aspects should be included and jointly propagated. This paper presents a study of the influence of the uncertainty in the macroscopic neutronic information that describes a three-dimensional core model on the most relevant results of the simulation of a Reactivity Induced Accident (RIA). The analyses of a BWR-RIA and a PWR-RIA have been carried out with a three-dimensional thermal-hydraulic and neutronic model for the coupled system TRACE-PARCS and RELAP-PARCS. The cross section information has been generated by the SIMTAB methodology based on the joint use of CASMO-SIMULATE. The statistically based methodology performs a Monte-Carlo kind of sampling of the uncertainty in the macroscopic cross sections. The size of the sampling is determined by the characteristics of the tolerance intervals by applying the Noether–Wilks formulas. A number of simulations equal to the sample size have been carried out in which the cross sections used by PARCS are directly modified with uncertainty, and non-parametric statistical methods are applied to the resulting sample of the values of the output variables to determine their intervals of tolerance.

  9. Crack growth testing of cold worked stainless steel in a simulated PWR primary water environment to assess susceptibility to stress corrosion cracking

    International Nuclear Information System (INIS)

    Tice, D.R.; Stairmand, J.W.; Fairbrother, H.J.; Stock, A.

    2007-01-01

    Although austenitic stainless steels do not show a high degree of susceptibility to stress corrosion cracking (SCC) in PWR primary environments, there is limited evidence from laboratory testing that crack propagation may occur under some conditions for materials in a cold-worked condition. A test program is therefore underway to examine the factors influencing SCC propagation in good quality PWR primary coolant. Type 304 stainless steel was subjected to cold working by either rolling (at ambient or elevated temperature) or fatigue cycling, to produce a range of yield strengths. Compact tension specimens were fabricated from these materials and tested in simulated high temperature (250-300 o C) PWR primary coolant. It was observed that the degree of crack propagation was influenced by the degree of cold work, the crack growth orientation relative to the rolling direction and the method of working. (author)

  10. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1983-08-01

    This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 2000 0 C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW) [de

  11. Horizontal loading test by whole model specimen simulating inner concrete structure of PWR type nuclear power plant

    International Nuclear Information System (INIS)

    Furuya, Noriyuki; Sekine, Masataka; Kimura, Kozo; Yamaguchi, Yoshihiro; Yamaguchi, Tsuneo; Takeda, Toshikazu

    1985-01-01

    The Nuclear Power Engineering Test Center has performed a horizontal loading test by a whole model specimen simulating the inner concrete structure of a PWR type nuclear power plant in order to investigate restoring force characteristics of reactor buildings. This report describes the results of examination of applicability to the test results of analysis methods based on elastic theory. The analysis results of elastic stiffness, concrete cracking load, rebar yielding loads and ultimate strength were compared with the test results. According to this examination, it is recognized that even these analysis methods based on elastic theory are comparatively effective for analysis of an inner concrete structure of fairly complex configuration, although there are limits of the scope of applicability. (author)

  12. Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

    International Nuclear Information System (INIS)

    Soares, Alexandre de Souza

    2014-01-01

    The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

  13. Greenhouse effects on Venus

    Science.gov (United States)

    Bell, Peter M.

    Calculations that used Pioneer-Venus measurements of atmosphere composition, temperature profiles, and radiative heating predicted Venus' surface temperature ‘very precisely,’ says the Ames Research Center. The calculations predict not only Venus' surface temperature but agree with temperatures measured at various altitudes above the surface by the four Pioneer Venus atmosphere probe craft.Using Pioneer-Venus spacecraft data, a research team has virtually proved that the searing 482° C surface temperature of Venus is due to an atmospheric greenhouse effect. Until now the Venus greenhouse effect has been largely a theory.

  14. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Min; Sabundjian, Gaianê, E-mail: smlee@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  15. Simulation of a severe accident at a typical PWR due to break of a hot leg ECCS line using MELCOR code

    International Nuclear Information System (INIS)

    Lee, Seung Min; Sabundjian, Gaianê

    2017-01-01

    The aim of this work was to simulate a severe accident at a typical PWR caused by break in Emergency Core Cooling System (ECCS) line of a hot leg using the MELCOR code. The nodalization of this typical PWR was elaborated by the Global Research for Safety (GRS) and provided to the CNEN for analysis of the severe accidents at the Angra 2, which is similar to that PWR. Although both of them are not identical the results obtained for that typical PWR may be valuable because of the lack of officially published calculation for Angra 2. Relevant parameters such as pressure, temperature and water level in various control volumes after the break in the hot leg were calculated as well as degree of core degradation and hydrogen concentration in containment. The result obtained in this work could be considered satisfactory in the sense that the physical phenomena reproduced by the simulation were in general very reasonable, and most of the events occurred within acceptable time intervals. However, the uncertainty analysis was not carried out in this work. Furthermore, this scenario could be used as a base for the study of the effectiveness of some preventive or/and mitigating measures of Severe Accident Management (SAMG) by adding associated conditions for each measure in its input. (author)

  16. Characterization of interfacial reactions and oxide films on 316L stainless steel in various simulated PWR primary water environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie; Xiao, Qian [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shanghai Key Laboratory of Advanced Ferrometallurgy, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Ru, Xiangkun; Peng, Hao; Xiong, Qi; Li, Hongjuan [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China)

    2017-06-15

    The effect of water chemistry on the electrochemical and oxidizing behaviors of 316L SS was investigated in hydrogenated, deaerated and oxygenated PWR primary water at 310 °C. Water chemistry significantly influenced the electrochemical impedance spectroscopy parameters. The highest charge-transfer resistance and oxide-film resistance occurred in oxygenated water. The highest electric double-layer capacitance and constant phase element of the oxide film were in hydrogenated water. The oxide films formed in deaerated and hydrogenated environments were similar in composition but different in morphology. An oxide film with spinel outer particles and a compact and Cr-rich inner layer was formed in both hydrogenated and deaerated water. Larger and more loosely distributed outer oxide particles were formed in deaerated water. In oxygenated water, an oxide film with hematite outer particles and a porous and Ni-rich inner layer was formed. The reaction kinetics parameters obtained by electrochemical impedance spectroscopy measurements and oxidation film properties relating to the steady or quasi-steady state conditions in the time-period of measurements could provide fundamental information for understanding stress corrosion cracking processes and controlling parameters. - Highlights: •Long-term EIS measurements of 316L SS in simulated PWR primary water. •Highest charge-transfer resistance and oxide film resistance in oxygenated water. •Highest electric double-layer capacitance and oxide film CPE in hydrogenated water. •Similar compositions, different shapes of oxides in deaerated/hydrogenated water. •Inner layer Cr-rich in hydrogenated/deaerated water, Ni-rich in oxygenated water.

  17. Venus gravity fields

    Science.gov (United States)

    Sjogren, W. L.; Ananda, M.; Williams, B. G.; Birkeland, P. W.; Esposito, P. S.; Wimberly, R. N.; Ritke, S. J.

    1981-01-01

    Results of Pioneer Venus Orbiter observations concerning the gravity field of Venus are presented. The gravitational data was obtained from reductions of Doppler radio tracking data for the Orbiter, which is in a highly eccentric orbit with periapsis altitude varying from 145 to 180 km and nearly fixed periapsis latitude of 15 deg N. The global gravity field was obtained through the simultaneous estimation of the orbit state parameters and gravity coefficients from long-period variations in orbital element rates. The global field has been described with sixth degree and order spherical harmonic coefficients, which are capable of resolving the three major topographical features on Venus. Local anomalies have been mapped using line-of-sight accelerations derived from the Doppler residuals between 40 deg N and 10 deg S latitude at approximately 300 km spatial resolution. Gravitational data is observed to correspond to topographical data obtained by radar altimeter, with most of the gravitational anomalies about 20-30 milligals. Simulations evaluating the isostatic states of two topographic features indicate that at least partial isostasy prevails, with the possibility of complete compensation.

  18. Computational simulation of natural convection of a molten core in lower head of a PWR pressure vessel

    International Nuclear Information System (INIS)

    Vieira, Camila Braga; Romero, Gabriel Alves; Jian Su

    2010-01-01

    Computational simulation of natural convection in a molten core during a hypothetical severe accident in the lower head of a typical PWR pressure vessel was performed for two-dimensional semi-circular geometry with isothermal walls. Transient turbulent natural convection heat transfer of a fluid with uniformly distributed volumetric heat generation rate was simulated by using a commercial computational fluid dynamics software ANSYS CFX 12.0. The Boussinesq model was used for the buoyancy effect generated by the internal heat source in the flow field. The two-equation k-ω based SST (Shear Stress Transport) turbulence model was used to mould the turbulent stresses in the Reynolds-Average Navier-Stokes equations (RANS). Two Prandtl numbers, 6:13 and 7:0, were considered. Five Rayleigh numbers were simulated for each Prandtl number used (109, 1010, 1011, 1012, and 1013). The average Nusselt numbers on the bottom surface of the semicircular cavity were in excellent agreement with Mayinger et al. (1976) correlation and only at Ra = 109 the average Nusselt number on the top flat surface was in agreement with Mayinger et al. (1976) and Kulacki and Emara (1975) correlations. (author)

  19. Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

    International Nuclear Information System (INIS)

    Mur, J.; Larrauri, D.

    1998-07-01

    Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

  20. Simulation of a PWR power plant for process control and diagnosis

    International Nuclear Information System (INIS)

    Ravnsbjerg Nielsen, F.

    1991-12-01

    A computer model of a simplified pressurized nuclear power plant is developed with aim at studies concerning process control, diagnosis and decision making. The model includes the traditional PWR plant components, primary circuit with reactor, pressurizer and steam generator, steam circuit with steam line, turbine and condenser, interconnected with pumps, valves and controllers. The model can be used for calculation of transients for both normal operation and incidents such as turbine trip, loss of feedwater, run down of pumps or various valve failures. The computer model is not directed to any specific existing plant. For convenience and alleviation in implementation the physical description of many components are simplified to an extent where the qualitative behavior of the system is not violated. For computer memory economy a variety of thermodynamical functions for water and steam have been approximated with analytical expressions based on table values. The model is implemented in the C language and has been run on both the IBM PC and the SUN workstation. (au) 8 tabs., 25 ills., 10 refs

  1. Fatigue-crack growth behavior of Type 347 stainless steels under simulated PWR water conditions

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seokmin; Min, Ki-Deuk; Yoon, Ji-Hyun; Kim, Min-Chul; Lee, Bong-Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Fatigue crack growth rate (FCGR) curve of stainless steel exists in ASME code section XI, but it is still not considering the environmental effects. The longer time nuclear power plant is operated, the more the environmental degradation issues of materials pop up. There are some researches on fatigue crack growth rate of S304 and S316, but researches of FCGR of S347 used in Korea nuclear power plant are insufficient. In this study, the FCGR of S347 stainless steel was evaluated in the PWR high temperature water conditions. The FCGRs of S347 stainless steel under pressurized-water conditions were measured by using compact-tension (CT) specimens at different levels of dissolved oxygen (DO) and frequency. 1. FCGRs of SS347 were slower than that in ASME XI and environmental effect did not occur when frequency was higher than 1Hz. 2. Fatigue crack growth is accelerated by corrosion fatigue and it is more severe when frequency is slower than 0.1Hz. 3. Increase of crack tip opening time increased corrosion fatigue and it deteriorated environmental fatigue properties.

  2. Low cycle fatigue of Alloy 690 and welds in a simulated PWR primary water environment

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jongdae; Cho, Pyungyeon; Jang, Changheui [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Cho, Pyungyeon [Khalifa Univ., Abu Dhabi (United Arab Emirates); Kim, Tae Soon; Lee, Yong Sung [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2013-05-15

    In this study, environmental fatigue tests for these materials were performed and the new prediction model of fatigue life of Alloy 690 and weld in primary water condition was proposed. To evaluate the fatigue life of Alloy 690 and 52M in a PWR environment, low cycle fatigue tests were performed and revised fatigue life prediction models and environmental factor were proposed. With the revised Fen model for Alloy 690 and 52M, the reliability of the fatigue life prediction has been improved. The reduction of low cycle fatigue life of metallic materials in the primary coolant water environments has been the subject of debate between the utility and regulator since 1980s. It became the significant licensing problem since the issue of RG-1.207 by U. S. NRC. The statistical model for the environmental factor, Fen, specified in RG-1.207 was based on the extensive test results accumulated by the ANL and Japanese national program. Of the materials, the limited fatigue life data of Ni-Cr-Fe alloys were used to develop the Fen for the alloys. Furthermore, test data for Alloy 690 and its weld are limited. Considering that Alloy 690 will be extensively used in the new nuclear power plants, additional effort to validate or improve current Fen model is required.

  3. Criteria for safety-related nuclear-power-plant operator actions: 1982 pressurized-water-reactor (PWR) simulator exercises

    International Nuclear Information System (INIS)

    Crowe, D.S.; Beare, A.N.; Kozinsky, E.J.; Haas, P.M.

    1983-06-01

    The primary objective of the Safety-Related Operator Action (SROA) Program at Oak Ridge National Laboratory is to provide a data base to support development of criteria for safety-related actions by nuclear power plant operators. When compared to field data collected on similar events, a base of operator performance data developed from the simulator experiments can then be used to establish safety-related operator action design evaluation criteria, evaluate the effects of performance shaping factors, and support safety/risk assessment analyses. This report presents data obtained from refresher training exercises conducted in a pressurized water reactor (PWR) power plant control room simulator. The 14 exercises were performed by 24 teams of licensed operators from one utility, and operator performance was recorded by an automatic Performance Measurement System. Data tapes were analyzed to extract operator response times (RTs) and error rate information. Demographic and subjective data were collected by means of brief questionnaires and analyzed in an attempt to evaluate the effects of selected performance shaping factors on operator performance

  4. The influence of simultaneous or sequential test conditions in the properties of industrial polymers, submitted to PWR accident simulations

    International Nuclear Information System (INIS)

    Carlin, F.; Alba, C.; Chenion, J.; Gaussens, G.; Henry, J.Y.

    1986-10-01

    The effect of PWR plant normal and accident operating conditions on polymers forms the basis of nuclear qualification of safety-related containment equipment. This study was carried out on the request of safety organizations. Its purpose was to check whether accident simulations carried out sequentially during equipment qualification tests would lead to the same deterioration as that caused by an accident involving simultaneous irradiation and thermodynamic effects. The IPSN, DAS and the United States NRC have collaborated in preparing this study. The work carried out by ORIS Company as well as the results obtained from measurement of the mechanical properties of 8 industrial polymers are described in this report. The results are given in the conclusion. They tend to show that, overall, the most suitable test cycle for simulating accident operating conditions would be one which included irradiation and consecutive thermodynamic shock. The results of this study and the results obtained in a previous study, which included the same test cycles, except for more severe thermo-ageing, have been compared. This comparison, which was made on three elastomers, shows that ageing after the accident has a different effect on each material [fr

  5. RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7

    International Nuclear Information System (INIS)

    Andrs, David; Berry, Ray; Gaston, Derek; Martineau, Richard; Peterson, John; Zhang, Hongbin; Zhao, Haihua; Zou, Ling

    2012-01-01

    The document contains the simulation results of a steady state model PWR problem with the RELAP-7 code. The RELAP-7 code is the next generation nuclear reactor system safety analysis code being developed at Idaho National Laboratory (INL). The code is based on INL's modern scientific software development framework - MOOSE (Multi-Physics Object-Oriented Simulation Environment). This report summarizes the initial results of simulating a model steady-state single phase PWR problem using the current version of the RELAP-7 code. The major purpose of this demonstration simulation is to show that RELAP-7 code can be rapidly developed to simulate single-phase reactor problems. RELAP-7 is a new project started on October 1st, 2011. It will become the main reactor systems simulation toolkit for RISMC (Risk Informed Safety Margin Characterization) and the next generation tool in the RELAP reactor safety/systems analysis application series (the replacement for RELAP5). The key to the success of RELAP-7 is the simultaneous advancement of physical models, numerical methods, and software design while maintaining a solid user perspective. Physical models include both PDEs (Partial Differential Equations) and ODEs (Ordinary Differential Equations) and experimental based closure models. RELAP-7 will eventually utilize well posed governing equations for multiphase flow, which can be strictly verified. Closure models used in RELAP5 and newly developed models will be reviewed and selected to reflect the progress made during the past three decades. RELAP-7 uses modern numerical methods, which allow implicit time integration, higher order schemes in both time and space, and strongly coupled multi-physics simulations. RELAP-7 is written with object oriented programming language C++. Its development follows modern software design paradigms. The code is easy to read, develop, maintain, and couple with other codes. Most importantly, the modern software design allows the RELAP-7 code to

  6. Gravity field of Venus - A preliminary analysis

    Science.gov (United States)

    Phillips, R. J.; Sjogren, W. L.; Abbott, E. A.; Smith, J. C.; Wimberly, R. N.; Wagner, C. A.

    1979-01-01

    The gravitational field of Venus obtained by tracking the Pioneer Venus Orbiter is examined. For each spacecraft orbit, two hours of Doppler data centered around periapsis were used to estimate spacecraft position and velocity and the velocity residuals obtained were spline fit and differentiated to produce line of sight gravitational accelerations. Consistent variations in line of sight accelerations from orbit to orbit reveal the presence of gravitational anomalies. A simulation of isostatic compensation for an elevated region on the surface of Venus indicates that the mean depth of compensation is no greater than about 100 km. Gravitational spectra obtained from a Fourier analysis of line of sight accelerations from selected Venus orbits are compared to the earth's gravitational spectrum and spherical harmonic gravitational potential power spectra of the earth, the moon and Mars. The Venus power spectrum is found to be remarkably similar to that of the earth, however systematic variations in the harmonics suggest differences in dynamic processes or lithospheric behavior.

  7. Magnetite solubility studies under simulated PWR primary-side conditions, using lithiated, hydrogenated water

    International Nuclear Information System (INIS)

    Hewett, John; Morrison, Jonathan; Cooper, Christopher; Ponton, Clive; Connolly, Brian; Dickinson, Shirley; Henshaw, Jim

    2014-01-01

    As software for modelling dissolution, precipitation, and transport of metallic species and subsequent CRUD deposition within nuclear plant becomes more advanced, there is an increasing need for accurate and reliable thermodynamic data. The solubility behaviour of magnetite is an example of such data, and is central to any treatment of CRUD solubility due to the prevalence of magnetite and nickel ferrites in CRUD. Several workers have shown the most consistent solubility data comes from once-through flowing systems. However, despite a strong consensus between the results in acidic to mildly alkaline solutions, there is disagreement between the results at approximately pH 25C 9 and higher. A programme of experimental work is on-going at the University of Birmingham, focusing on solubility of metal oxides (e.g., magnetite) in conditions relevant to PWR primary coolant. One objective of this programme is to calculate thermodynamic constants from the data obtained. Magnetite solubility from 200 to 300°C, in lithiated, hydrogenated water of pH 25C 9–11 is being studied using a once-through rig constructed of 316L stainless steel. The feedwater is pumped at 100 bar pressure through a heated bed of magnetite granules, and the output solution is collected and analysed for iron and several other metals by ICP-MS. This paper presents results from preliminary tests without magnetite granules, in which the corroding surface of the rig itself was used as the sole source of soluble iron and of dissolved hydrogen. Levels of iron were generally within an order of magnitude of literature solubility values. Comparison of results at different flow rates and temperatures, in conjunction with conclusions drawn from the published literature, suggests that this is likely due to the presence of particulate matter in a greatly under-saturated solution, compensating for the low surface area of oxide in contact with the solution. (author)

  8. Corrosion behaviour of E110- and E635- type zirconium alloys under PWR irradiation simulating conditions

    International Nuclear Information System (INIS)

    Markelov, V.A.; Novikov, V.V.; Kon'kov, V.F.; Tselishchev, A.V.; Dologov, A.B.; Zmitko, M.; Maserik, V.; Kocik, J.

    2008-01-01

    As structural materials for VVER 1000 fuel rod claddings and FA components use is made of zirconium alloys E110 (Zr 1Nb) and E635 (Zr 1.2Sn 1Nb 0.35Fe) that meet the design parameters of operation. Nonetheless, the work is in progress to perfect those alloys to reach higher corrosion and shape change resistance. At VNIINM updated E110M and E635M alloys have been developed on E110 and E635 bases. To assess the corrosion behaviour of the updated alloys in comparison to the base alloys their cladding samples were tested in RVS 3 loop of LWR 15 reactor (NRI, Rez) in PWR water chemistry with coolant surface and volume boiling. The data are presented on the influence effected by in pile irradiation for up to 324 days on oxide coat thickness and microstructure of fuel claddings produced from the four tested alloys. It has been revealed that E110 alloy its updated version E110M and E635M alloy compared to E635 have higher corrosion resistances. The paper discusses th+e results on the in pile corrosion of cladding samples from the alloys under study in comparison to the results acquired for similar samples tested in LWR 15 inactive channel and under autoclave conditions. Using methods of TEM, EDX analyses of extraction replicas dislocation structure and phase composition changes were studied in samples of all four alloy claddings LWR 15 reactor irradiated to the material damage dose of 1.5 dpa. The interrelation is discussed between irradiation effected strengthening and corrosion of fuel claddings made of E110 and E635 type zirconium alloys and the evolution of their structure and phase states

  9. Methodology to evaluate the crack growth rate by stress corrosion cracking in dissimilar metals weld in simulated environment of PWR nuclear reactor

    International Nuclear Information System (INIS)

    Paula, Raphael G.; Figueiredo, Celia A.; Rabelo, Emerson G.

    2013-01-01

    Inconel alloys weld metal is widely used to join dissimilar metals in nuclear reactors applications. It was recently observed failures of weld components in plants, which have triggered an international effort to determine reliable data on the stress corrosion cracking behavior of this material in reactor environment. The objective of this work is to develop a methodology to determine the crack growth rate caused by stress corrosion in Inconel alloy 182, using the specimen (Compact Tensile) in simulated PWR environment. (author)

  10. Effects of PbO on the oxide films of incoloy 800HT in simulated primary circuit of PWR

    International Nuclear Information System (INIS)

    Tan, Yu; Yang, Junhan; Wang, Wanwan; Shi, Rongxue; Liang, Kexin; Zhang, Shenghan

    2016-01-01

    Effects of trace PbO on oxide films of Incoloy 800HT were investigated in simulated primary circuit water chemistry of PWR, also with proper Co addition. The trace PbO addition in high temperature water blocked the protective spinel oxides formation of the oxide films of Incoloy 800HT. XPS results indicated that the lead, added as PbO into the high temperature water, shows not only +2 valance but also +4 and 0 valances in the oxide film of 800HT co-operated with Fe, Cr and Ni to form oxides films. Potentiodynamic polarization results indicated that as PbO concentration increased, the current densities of the less protective oxide films of Incoloy 800HT decreased in a buffer solution tested at room temperature. The capacitance results indicated that the donor densities of oxidation film of Incoloy 800HT decreased as trace PbO addition into the high temperature water. - Highlights: • Trace PbO addition into the high temperature water block the formation of spinel oxides on Incoloy 800HT. • The donor density of oxide film decreases with trace PbO addition. • The current density of potentiodynamic polarization decreases of oxide film with trace PbO addition.

  11. Effect of surface state on the oxidation behavior of welded 308L in simulated nominal primary water of PWR

    Science.gov (United States)

    Ming, Hongliang; Zhang, Zhiming; Wang, Jiazhen; Zhu, Ruolin; Ding, Jie; Wang, Jianqiu; Han, En-Hou; Ke, Wei

    2015-05-01

    The oxidation behavior of 308L weld metal (WM) with different surface state in the simulated nominal primary water of pressurized water reactor (PWR) was studied by scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), X-ray diffraction (XRD) analyzer and X-ray photoelectron spectroscopy (XPS). After 480 h immersion, a duplex oxide film composed of a Fe-rich outer layer (Fe3O4, Fe2O3 and a small amount of NiFe2O4, Ni(OH)2, Cr(OH)3 and (Ni, Fe)Cr2O4) and a Cr-rich inner layer (FeCr2O4 and NiCr2O4) can be formed on the 308L WM samples with different surface state. The surface state has no influence on the phase composition of the oxide films but obviously affects the thickness of the oxide films and the morphology of the oxides (number & size). With increasing the density of dislocations and subgrain boundaries in the cold-worked superficial layer, the thickness of the oxide film, the number and size of the oxides decrease.

  12. Conception of a PWR simulator as a tool for safety analysis

    International Nuclear Information System (INIS)

    Lanore, J.M.; Bernard, P.; Romeyer Dherbey, J.; Bonnet, C.; Quilchini, P.

    1982-09-01

    A simulator can be a very useful tool for safety analysis to study accident sequences involving malfunctions of the systems and operator interventions. The main characteristics of the simulator SALAMANDRE (description of the systems, physical models, programming organization, control desk) have then been selected according tot he objectives of safety analysis

  13. Concept of scaled test facility for simulating the PWR thermalhydraulic behaviour

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1990-01-01

    This work deals with the design of a scaled test facility of a typical pressurized water reactor plant, to simulation of small break Loss-of-Coolant Accident. The computer code RELAP 5/ MOD1 has been utilized to simulate the accident and to compare the test facility behaviour with the reactor plant one. The results demonstrate similar thermal-hydraulic behaviours of the two sistema. (author)

  14. Verification and validation of a numeric procedure for flow simulation of a 2x2 PWR rod bundle

    International Nuclear Information System (INIS)

    Santos, Andre A.C.; Barros Filho, Jose Afonso; Navarro, Moyses A.

    2011-01-01

    Before Computational Fluid Dynamics (CFD) can be considered as a reliable tool for the analysis of flow through rod bundles there is a need to establish the credibility of the numerical results. Procedures must be defined to evaluate the error and uncertainty due to aspects such as mesh refinement, turbulence model, wall treatment and appropriate definition of boundary conditions. These procedures are referred to as Verification and Validation (V and V) processes. In 2009 a standard was published by the American Society of Mechanical Engineers (ASME) establishing detailed procedures for V and V of CFD simulations. This paper presents a V and V evaluation of a numerical methodology applied to the simulation of a PWR rod bundle segment with a split vane spacer grid based on ASMEs standard. In this study six progressively refined meshes were generated to evaluate the numerical uncertainty through the verification procedure. Experimental and analytical results available in the literature were used in this study for validation purpose. The results show that the ASME verification procedure can give highly variable predictions of uncertainty depending on the mesh triplet used for the evaluation. However, the procedure can give good insight towards optimization of the mesh size and overall result quality. Although the experimental results used for the validation were not ideal, through the validation procedure the deficiencies and strengths of the presented modeling could be detected and reasonably evaluated. Even though it is difficult to obtain reliable estimates of the uncertainty of flow quantities in the turbulent flow, this study shows that the V and V process is a necessary step in a CFD analysis of a spacer grid design. (author)

  15. NCS--a software for visual modeling and simulation of PWR nuclear power plant control system

    International Nuclear Information System (INIS)

    Cui Zhenhua

    1998-12-01

    The modeling and simulation of nuclear power plant control system has been investigated. Some mathematical models for rapid and accurate simulation are derived, including core models, pressurizer model, steam generator model, etc. Several numerical methods such as Runge-Kutta Method and Treanor Method are adopted to solve the above system models. In order to model the control system conveniently, a block diagram-oriented visual modeling platform is designed. And the Discrete Similarity Method is used to calculate the control system models. A corresponding simulating software, NCS, is developed for researching on the control systems of commercial nuclear power plant. And some satisfactory results are obtained. The research works will be of referential and applying value to design and analysis of nuclear power plant control system

  16. Development of an engineering simulator for integral type PWR for nuclear ship

    International Nuclear Information System (INIS)

    Takahashi, Teruo; Shimazaki, Junya; Nakazawa, Toshio

    2000-01-01

    JAERI has developed a real-time engineering simulator for the integral type reactor MRX (Marine Reactor X) of power 100 MWt to evaluate the design and operational performance and to study highly automatic operations of a reactor plant. Marine reactor is operated under the conditions of pitching and rolling and load change, in comparison with a reactor for a land-based generating plant. And the MRX has systems with structural features, such as water-filled containment vessel, once-through type steam generator and emergency decay heat removal system. Considerations are paid to take these operational conditions and structural features into the simulation model. It is shown that the simulated results are consistent with the planned design and operational performance, and on the other hand present us some technical issues to be investigated in the design specifications. (author)

  17. Stochastic simulation of PWR vessel integrity for pressurized thermal shock conditions

    International Nuclear Information System (INIS)

    Jackson, P.S.; Moelling, D.S.

    1984-01-01

    A stochastic simulation methodology is presented for performing probabilistic analyses of Pressurized Water Reactor vessel integrity. Application of the methodology to vessel-specific integrity analyses is described in the context of Pressurized Thermal Shock (PTS) conditions. A Bayesian method is described for developing vessel-specific models of the density of undetected volumetric flaws from ultrasonic inservice inspection results. Uncertainty limits on the probabilistic results due to sampling errors are determined from the results of the stochastic simulation. An example is provided to illustrate the methodology

  18. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  19. Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor; Ensaios de lixiviacao em produtos cimentados contendo rejeito simulado de concentrado do evaporador de reator PWR

    Energy Technology Data Exchange (ETDEWEB)

    Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes, E-mail: hauczmj@cdtn.b, E-mail: jaalmeida@cdtn.b, E-mail: tellocc@cdtn.b, E-mail: fdc@cdtn.b, E-mail: seless@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-10-26

    This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

  20. Chemical Weathering on Venus

    Science.gov (United States)

    Zolotov, Mikhail

    2018-01-01

    Chemical and phase compositions of Venus's surface could reflect history of gas- and fluid-rock interactions, recent and past climate changes, and a loss of water from the Earth's sister planet. The concept of chemical weathering on Venus through gas-solid type reactions has been established in 1960s after the discovery of hot and dense CO2-rich atmosphere inferred from Earth-based and Mariner 2 radio emission data. Initial works suggested carbonation, hydration, and oxidation of exposed igneous rocks and a control (buffering) of atmospheric gases by solid-gas type chemical equilibria in the near-surface lithosphere. Calcite, quartz, wollastonite, amphiboles, and Fe oxides were considered likely secondary minerals. Since the late 1970s, measurements of trace gases in the sub-cloud atmosphere by Pioneer Venus and Venera entry probes and Earth-based infrared spectroscopy doubted the likelihood of hydration and carbonation. The H2O gas content appeared to be low to allow a stable existence of hydrated and a majority of OH-bearing minerals. The concentration of SO2 was too high to allow the stability of calcite and Ca-rich silicates with respect to sulfatization to CaSO4. In 1980s, the supposed ongoing consumption of atmospheric SO2 to sulfates gained support by the detection of an elevated bulk S content at Venera and Vega landing sites. The induced composition of the near-surface atmosphere implied oxidation of ferrous minerals to magnetite and hematite, consistent with the infrared reflectance of surface materials. The likelihood of sulfatization and oxidation has been illustrated in modeling experiments at simulated Venus conditions. Venus's surface morphology suggests that hot surface rocks and fines of mainly mafic composition contacted atmospheric gases during several hundreds of millions years since a global volcanic resurfacing. Some exposed materials could have reacted at higher and lower temperatures in a presence of diverse gases at different altitudinal

  1. Computer code to simulate transients in a steam generator of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Silva, J.M. da.

    1979-01-01

    A digital computer code KIBE was developed to simulate the transient behavior of a Steam Generator used in Pressurized Water Reactor Power PLants. The equations of Conservation of mass, energy and momentum were numerically integrated by an implicit method progressively in the several axial sections into which the Steam Generator was divided. Forced convection heat transfer was assumed on the primary side, while on the secondary side all the different modes of heat transfer were permitted and deternined from the various correlations. The stability of the stationary state was verified by its reproducibility during the integration of the conservation equation without any pertubation. Transient behavior resulting from pertubations in the flow and the internal energy (temperature) at the inlet of the primary side were simulated. The results obtained exhibited satisfactory behaviour. (author) [pt

  2. Computer simulation of black out followed by multiple failures in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1989-01-01

    The computer code RELAP 5/MOD 1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 MWe pressurized water reactor plant of the KWU design during a station blackout following a inadequate performance of the pressurizer and steam generator safety valves. During the simulation the reactor scram system the emergency coolant system of the primary loop and the emergency Feedwater system of the secondary loop are considered inactive. (author) [pt

  3. Modelling of core protection and monitoring system for PWR nuclear power plant simulator

    International Nuclear Information System (INIS)

    Jung Kun Lee; Byoung Sung Han

    1997-01-01

    A nuclear power plant simulator was developed for Younggwang units 3 and 4 nuclear power plant (YGN Nos 3 and 4) in Korea; it has been in operation on training center since November 1996. The core protection calculator (CPC) and the core operating limit supervisory system (COLSS) for the simulator were also developed. The CPC is a digital computer-based core protection system, which performs on-line calculation of departure from nucleate boiling ratio (DNBR) and local power density (LPD). It initiates reactor trip when the core conditions exceed designated DNBR or LPD limitations. The COLSS is designed to assist operators by implementing the limiting conditions for operations in the technical specifications. With these systems, it is possible to increase capacity factor and safety of nuclear power plants, because the COLSS data can show accurate operation margin to plant operators and the CPC can protect reactor core. In this study, the function of CPC/COLSS is analyzed in detail, and then simulation model for CPC/COLSS is presented based on the function. Compared with the YGN Nos 3 and 4 plant operation data and CEDIPS/COLSS FORTRAN code test results, the predictions with the model show reasonable results. (Author)

  4. JMCT Monte Carlo simulation analysis of full core PWR Pin-By-Pin and shielding

    International Nuclear Information System (INIS)

    Deng, L.; Li, G.; Zhang, B.; Shangguan, D.; Ma, Y.; Hu, Z.; Fu, Y.; Li, R.; Hu, X.; Cheng, T.; Shi, D.

    2015-01-01

    This paper describes the application of the JMCT Monte Carlo code to the simulation of Kord Smith Challenge H-M model, BEAVRS model and Chinese SG-III model. For H-M model, the 6.3624 millions tally regions and the 98.3 billion neutron histories do. The detailed pin flux and energy deposition densities obtain. 95% regions have less 1% standard deviation. For BEAVRS model, firstly, we performed the neutron transport calculation of 398 axial planes in the Hot Zero Power (HZP) status. Almost the same results with MC21 and OpenMC results are achieved. The detailed pin-power density distribution and standard deviation are shown. Then, we performed the calculation of ten depletion steps in 30 axial plane cases. The depletion regions exceed 1.5 million and 12,000 processors uses. Finally, the Chinese SG-III laser model is simulated. The neutron and photon flux distributions are given, respectively. The results show that the JMCT code well suits for extremely large reactor and shielding simulation. (author)

  5. Experimentation, modelling and simulation of water droplets impact on ballooned sheath of PWR core fuel assemblies in a LOCA situation

    International Nuclear Information System (INIS)

    Lelong, Franck

    2010-01-01

    In a pressurized water reactor (PWR), during a Loss Of Coolant Accident (LOCA), liquid water evaporates and the fuel assemblies are not cooled anymore; as a consequence, the temperature rises to such an extent that some parts of the fuel assemblies can be deformed resulting in 'ballooned regions'. When reflooding occurs, the cooling of these partially blocked parts of the fuel assemblies will depend on the coolant flow that is a mixture of overheated vapour and under-saturated droplets. The aim of this thesis is to study the heat transfer between droplets and hot walls of the fuel rods. In this purpose, an experimental device has been designed in accordance with droplets and wall features (droplet velocity and diameter, wall temperature) representative of LOCA conditions. The cooling of a hot Nickel disk, previously heated by induction, is cooled down by a stream of monodispersed droplet. The rear face temperature profiles are measured by infrared thermography. Then, the estimation of wall heat flux is performed by an inverse conduction technique from these infrared images. The effect of droplet dynamical properties (diameter, velocity) on the heat flux is studied. These experimental data allow us to validate an analytical model of heat exchange between droplet and hot slab. This model is based on combined dynamical and thermal considerations. On the one hand, the droplet dynamics is considered through a spring analogy in order to evaluate the evolution of droplet features such as the spreading diameter when the droplet is squeezed over the hot surface. On the other hand, thermal parameters, such as the thickness of the vapour cushion beneath the droplet, are determined from an energy balance. In the short term, this model will be integrated in a CFD code (named NEPTUNE-CFD) to simulate the cooling of a reactor core during a LOCA, taking into account the droplet/wall heat exchange. (author)

  6. Numerical simulation of radioisotope's dependency on containment performance for large dry PWR containment under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Mehboob, Khurram, E-mail: khurramhrbeu@gmail.com [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Xinrong, Cao [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ahmed, Raheel [College of Automation, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China); Ali, Majid [College of Nuclear Science and Technology, Harbin Engineering University, 145-31 Nantong Street, Nangang District, Harbin, Heilongjiang 150001 (China)

    2013-09-15

    Highlights: • Calculation and comparison of activity of BURN-UP code with ORIGEN2 code. • Development of SASTC computer code. • Radioisotopes dependency on containment ESFs. • Mitigation in atmospheric release with ESFs operation. • Variation in radioisotopes source term with spray flow and pH value. -- Abstract: During the core melt accidents large amount of fission products can be released into the containment building. These fission products escape into the environment to contribute in accident source term. The mitigation in environmental release is demanded for such radiological consequences. Thus, countermeasures to source term, mitigations of release of radioactivity have been studied for 1000 MWe PWR reactor. The procedure of study is divided into five steps: (1) calculation and verification of core inventory, evaluated by BURN-UP code, (2) containment modeling based on radioactivity removal factors, (3) selection of potential accidents initiates the severe accident, (4) calculation of release of radioactivity, (5) study the dependency of release of radioactivity on containment engineering safety features (ESFs) inducing mitigation. Loss of coolant accident (LOCA), small break LOCA and flow blockage accidents (FBA) are selected as initiating accidents. The mitigation effect of ESFs on source term has been studied against ESFs performance. Parametric study of release of radioactivity has been carried out by modeling and simulating the containment parameters in MATLAB, which takes BURN-UP outcomes as input along with the probabilistic data. The dependency of iodine and aerosol source term on boric and caustic acid spray has been determined. The variation in source term mitigation with the variation of containment spray flow rate and pH values have been studied. The variation in containment retention factor (CRF) has also been studied with the ESF performance. A rapid decrease in source term is observed with the increase in pH value.

  7. The KINA neutronic module of the LEGO code for steady-state and transient PWR plant simulations

    International Nuclear Information System (INIS)

    Nicolopoulos, D.; Pollacchini, L.; Vimercati, G.; Spelta, S.

    1989-01-01

    The Automation Research Center (CRA) of ENEl has implemented some models for analyzing both incidental and operational transients in PWR power plants. For such models an axial neutron kinetics module characterized by high computational efficency with adequate results accuracy was called for. CISE has been entrusted with the task of implementing such a module named KINA and based on IQS (Improved Quasi Static) method, to be included in the library of LEGO modular code used by CRA to set up PWR power models. Moreover, The KINA module has been adapted to the neutron constants computing model developed by the EdF-SEPTEN, which has been using and improving the LEGO code for a long time in cooperation with ENEL-CRA. In this paper, after some remarks on the LEGO code, a general description of KINA neutronic module is given. The resylts of a preliminary validation activity of KINA for an EdF 1300 MWe PWR plant are also presented

  8. Implementation of high fidelity models for the conditions of operation in stop in PWR simulators

    International Nuclear Information System (INIS)

    Gonzalez Sevillano, I.; Jimenez Bogarin, R.; Ortega Pascual, F.

    2014-01-01

    The operation in stop cold conditions and in particular the States of operation with reduced inventory, the call of half loop or half nozzle, is becoming increasingly more important. These States of operation are characterized by having the coolant level approximately on the generatrix of the branches, so that any deviation in the level or malfunction of the system for the disposal of waste heat could lead to compromising situations. The importance of this type of situation is reflected in the APS in other modes (APSOM), which show that the risk in these conditions may be comparable to the power. Hence the importance that the simulator training programmes include scenarios that cover these States of operation. The article describes on the one hand, the difficulties encountered in the simulation of situations characterized by low pressure and presence of Non-Condensable and, on the other hand, its implementation, not only in the field of training of plant personnel, but also in the field of review/validation of operating procedures. (Author)

  9. Ratchetting in PWR: on use of numerical simulations to reduce the conservatism of design rules

    International Nuclear Information System (INIS)

    Geyer, P.; Meziere, Y.; Taheri, S.

    1996-01-01

    The French design Code (RCC-M Code) for pressurised water nuclear power plants includes criteria with regard to the risk of progressive deformation. With the new operating conditions, some components no longer check this criteria. Therefore, R and D actions were initiated in 1991 by EDF, CEA and FRAMATOME with the aim of putting forward more suitable design criteria that the present rules, which seem to be too conservative. In the two first parts, this paper presents the RCC-M design rules with regard to the risk progressive deformation, the components that don't check the criteria, and the alternative solutions proposed by FRAMATOME. In the two last parts, we focus on one R and D study to show the theoretical and experimental ways used to solve the problem. Numerical simulations become necessary to model experimental results. But constitutive equations for cyclic plasticity derived from Chaboche model appear inadequate to get accurate predictions. (authors)

  10. Scaling studies - PWR

    International Nuclear Information System (INIS)

    Sonneck, G.

    1983-05-01

    A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

  11. 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)

  12. SIPA, a PWR simulator for post-accident training and studies

    International Nuclear Information System (INIS)

    Peltier, J.; Poizat, F.

    1990-01-01

    SIPA (Simulator for Post-Accident conditions) which is now under development will be operated by EDF and CEA. Each organization will have its own version, SIPA 1 for EDF and SIPA 2 for CEA. The three main purposes will meet the needs of EDF and CEA as described below: - training of the EDF's ISR (Ingenieurs de Surete et Radioprotection = Shift Safety Advisors) which needs physical relevance, real time during accidental transients and visualisation of two-phase flow phenomena to well understand what could physically happen, - studies for EDF's designs which require calculation of a lot of points or scenarios. Quality Assurance of the models and data package, interactivity for procedure finalisation, availability of resources to all engineers, and possibility of creation of new models, - safety analysis requirements for CEA/IPSN (technical support of the French safety authority, the Central Service for the Safety of Nuclear Installations) which includes the actual safety analysis (analysis of procedures, design basis accidents, probabilistic safety analysis, real incidents studies, reactor tests...), the preparation and the execution of safety drills and training of engineer analysts

  13. Missions to Venus

    Science.gov (United States)

    Titov, D. V.; Baines, K. H.; Basilevsky, A. T.; Chassefiere, E.; Chin, G.; Crisp, D.; Esposito, L. W.; Lebreton, J.-P.; Lellouch, E.; Moroz, V. I.; Nagy, A. F.; Owen, T. C.; Oyama, K.-I.; Russell, C. T.; Taylor, F. W.; Young, R. E.

    2002-10-01

    Venus has always been a fascinating objective for planetary studies. At the beginning of the space era Venus became one of the first targets for spacecraft missions. Our neighbour in the solar system and, in size, the twin sister of Earth, Venus was expected to be very similar to our planet. However, the first phase of Venus spacecraft exploration in 1962-1992 by the family of Soviet Venera and Vega spacecraft and US Mariner, Pioneer Venus, and Magellan missions discovered an entirely different, exotic world hidden behind a curtain of dense clouds. These studies gave us a basic knowledge of the conditions on the planet, but generated many more questions concerning the atmospheric composition, chemistry, structure, dynamics, surface-atmosphere interactions, atmospheric and geological evolution, and the plasma environment. Despite all of this exploration by more than 20 spacecraft, the "morning star" still remains a mysterious world. But for more than a decade Venus has been a "forgotten" planet with no new missions featuring in the plans of the world space agencies. Now we are witnessing the revival of interest in this planet: the Venus Orbiter mission is approved in Japan, Venus Express - a European orbiter mission - has successfully passed the selection procedure in ESA, and several Venus Discovery proposals are knocking at the doors of NASA. The paper presents an exciting story of Venus spacecraft exploration, summarizes open scientific problems, and builds a bridge to the future missions.

  14. Simulation study on insoluble granular corrosion products deposited in PWR core

    International Nuclear Information System (INIS)

    Yang Xu; Zhou Tao; Ru Xiaolong; Lin Daping; Fang Xiaolu

    2014-01-01

    In the operation of reactor, such as fuel rods, reactor vessel internals etc. will be affected by corrosion erosion of high pressure coolant. It will produce many insoluble corrosion products. The FLUENT software is adopted to simulate insoluble granular corrosion products deposit distribution in the reactor core. The fluid phase uses the standard model to predict the flow field in the channel and forecast turbulence variation in the near-wall region. The insoluble granular corrosion products use DPM (Discrete Phase Model) to track the trajectory of the particles. The discrete phase model in FLUENT follows the Euler-Lagrange approach. The fluid phase is treated as a continuum by solving the Navier-Stokes equations, while the dispersed phase is solved by tracking a large number of particles through the calculated flow field. Through the study found, Corrosion products particles form high concentration area near the symmetry, and the entrance section of the corrosion products particles concentration is higher than export section. Corrosion products particles deposition attached on large area for the entrance of the cladding, this will change the core neutron flux distribution and the thermal conductivity of cladding material, and cause core axial offset anomaly (AOA). Corrosion products particles dot deposit in the outlet of cladding, which can lead to pitting phenomenon in a sheath. Pitting area will cause deterioration of heat transfer, destroy the cladding integrity. In view of the law of corrosion products deposition and corrosion characteristics of components in the reactor core. this paper proposes regular targeted local cleanup and other mitigation measures. (authors)

  15. BI/TRI-dimensional effects observed in PWR fuel during transient conditions and their numerical simulation

    Energy Technology Data Exchange (ETDEWEB)

    Linet, B; Hourdequin, N [Departement de Mecanique et Technologie, CEA Centre d` Etudes Nucleaires de Saclay, Gif-sur-Yvette (France)

    1997-08-01

    TOUTATIS is the modular program (both modules 2D and 3D are included) from the METERO project developed by the French Atomic Energy Commission ``CEA``. The model allows the user to calculate the deformations connected to the pellet-clad systems, and hence the Pellet-Cladding Interactions ``PCI`` induced by unilateral contact. Furthermore TOUTATIS provides sufficient versatility to allow the simulation of almost any phenomena, from creep and plasticity to the stress corrosion (residual stresses, dish filling of the pellets from the center, thermo-mechanical feedback) or fuel cracking (3D). The general approach provides a unique capability for understanding different phenomena, some of which remain still unexplained. The first example is related to rod bending, since this phenomenon has been observed in some experimental reactors. Several possible explanations have been put forward, such as flux dipping, buckling or thermohydraulic perturbations. Indeed a spatial parabolic distribution of the flux induces a shift of the isopower area in the pellets, but its effect decreases progressively as the distance from the center of the pellet is increased. So the variations on the clad temperature are just a few degrees and cannot produce the stated rod bending. The second hypothesis was based on a thermohydraulic perturbation. Both chosen configurations (azymutal area/small spot), which induced a thermal perturbation (corroborated by shift of the bubble area), are nevertheless insufficient to bring about the recorded strains. Lastly the calculations performed with the 3D model showed clearly that this rod bending was caused by single buckling induced itself by the immobilization of the rod in experimental channel. 19 figs.

  16. Boron dilution transients during natural circulation flow in PWR-Experiments and CFD simulations

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)], E-mail: T.Hoehne@fzd.de; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter [Forschungszentrum Dresden-Rossendorf (FZD)-Institute of Safety Research, P.O. Box 510119, D-01314 Dresden (Germany)

    2008-08-15

    Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV). In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.

  17. BI/TRI-dimensional effects observed in PWR fuel during transient conditions and their numerical simulation

    International Nuclear Information System (INIS)

    Linet, B.; Hourdequin, N.

    1997-01-01

    TOUTATIS is the modular program (both modules 2D and 3D are included) from the METERO project developed by the French Atomic Energy Commission ''CEA''. The model allows the user to calculate the deformations connected to the pellet-clad systems, and hence the Pellet-Cladding Interactions ''PCI'' induced by unilateral contact. Furthermore TOUTATIS provides sufficient versatility to allow the simulation of almost any phenomena, from creep and plasticity to the stress corrosion (residual stresses, dish filling of the pellets from the center, thermo-mechanical feedback) or fuel cracking (3D). The general approach provides a unique capability for understanding different phenomena, some of which remain still unexplained. The first example is related to rod bending, since this phenomenon has been observed in some experimental reactors. Several possible explanations have been put forward, such as flux dipping, buckling or thermohydraulic perturbations. Indeed a spatial parabolic distribution of the flux induces a shift of the isopower area in the pellets, but its effect decreases progressively as the distance from the center of the pellet is increased. So the variations on the clad temperature are just a few degrees and cannot produce the stated rod bending. The second hypothesis was based on a thermohydraulic perturbation. Both chosen configurations (azymutal area/small spot), which induced a thermal perturbation (corroborated by shift of the bubble area), are nevertheless insufficient to bring about the recorded strains. Lastly the calculations performed with the 3D model showed clearly that this rod bending was caused by single buckling induced itself by the immobilization of the rod in experimental channel. 19 figs

  18. Development of neutron own codes for the simulation of PWR reactor core; Desarrollo de codigos neutronicos propios para la simulacion del nucleo de reactores PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ahnert, C.; Cabellos, O.; Garcia-Herranz, N.; Cuervo, D.; Herrero, J. J.; Jimenez, J.; Ochoa, R.

    2011-07-01

    The core physic simulation is enough complex to need computers and ad-hoc software, and its evolution is to best-estimate methodologies, in order to improve availability and safety margins in the power plant operation. the Nuclear Engineering Department (UPM) has developed the SEANAP System in use in several power plants in Spain, with simulation in 3D and at the pin level detail, of the nominal and actual core burnup, with the on-line surveillance, and operational maneuvers optimization. (Author) 8 refs.

  19. Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function {Phi}(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time

  20. Monte Carlo simulation of the electron and X-ray depth distribution for quantitative electron probe microanalysis of PWR spent fuels

    International Nuclear Information System (INIS)

    Kwon, Hyoung Mun; Lee, Hyung Kwon; Son, Young Zoon; Chun, Yong Bum

    2011-01-01

    Electron probe microanalysis requires several corrections to quantify an element of a specimen. The X-rays produced by the primary beam are created at some depth in the specimen. This distribution is usually represented as the function Φ(pz), and it is possible to calculate the correction factors for atomic number and absorption effects. The electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to quantify some elements of the PWR spent fuel with 50 GWd/tU of burnup and 2 years of cooling time

  1. ROX PWR

    International Nuclear Information System (INIS)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H.

    1999-01-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO 2 , ThO 2 and Er 2 O 3 , and a heterogeneous core with Zr-ROX and UO 2 assemblies. As a result, the additives UO 2 + Er 2 O 3 are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO 2 heterogeneous core, further reduction of power peaking seems necessary. (author)

  2. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

    International Nuclear Information System (INIS)

    Belem, J.A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

  3. Investigation of irradiation induced inter-granular stress corrosion cracking susceptibility on austenitic stainless steels for PWR by simulated radiation induced segregation materials

    Energy Technology Data Exchange (ETDEWEB)

    Yonezawa, Toshio; Fujimoto, Koji; Kanasaki, Hiroshi; Iwamura, Toshihiko [Mitsubishi Heavy Industries Ltd., Takasago R and D Center, Takasago, Hyogo (Japan); Nakada, Shizuo; Ajiki, Kazuhide [Mitsubishi Heavy Industries Ltd., Kobe Shipyard and Machinery Works, Kobe, Hyogo (Japan); Urata, Sigeru [General Office of Nuclear and Fossil Power Production, Kansai Electric Power Co., Inc., Osaka (Japan)

    2000-07-01

    An Irradiation Assisted Stress Corrosion Cracking (IASCC) has not been found in Pressurized Water Reactors (PWRs). However, the authors have investigated on the possibility of IASCC so as to be able to estimate the degradation of PWR plants up to the end of their lifetime. In this study, the authors melted the test alloys whose bulk compositions simulated the grain boundary compositions of irradiated Type 304 and Type 316 CW stainless steels. Low chromium, high nickel and silicon (12%Cr-28%Ni-3%Si) steel showed high susceptibility to PWSCC (Primary Water Stress Corrosion Cracking) by SSRT (Slow Strain Rate Tensile) test in simulated PWR primary water. PWSCC susceptibility of the test steels increases with a decrease of chromium content and a increase of nickel and silicon contents. The aged test steel included coherent M{sub 23}C{sub 6} carbides with matrices at the grain boundaries showed low PWSCC susceptibility. This tendency is in very good agreement with that of the PWSCC susceptibility of nickel based alloys X-750 and 690. From these results, if there is the possibility of IASCC for austenitic stainless steels in PWRs, in the future, the IASCC shall be caused by the PWSCC as a result of irradiation induced grain boundary segregation. (author)

  4. Some Lessons Learned From the SIPACT Simulations on the Design of PWR and Improvement of AM Measures

    International Nuclear Information System (INIS)

    Pochard, R.; Jedrzejewski, F.; Nilsuwankosit, S.

    2002-01-01

    In the general context of the nuclear activities, life extension of the existing plants is the interesting option for countries that are already well equipped with NPPs. As the working life of 60 years is now expected possible for some well maintained plants, their safety measures needs to be improved such that they should be comparable to the new or future designs, taken into account the results from the probabilistic and the deterministic accident analysis. To accomplish this aim, the Accident Management (AM) is the important part of the process that must be utilized including possible automation of some processes. At INSTN, the extensive sensitivity studies related to the feed and bleed process on the primary and the secondary side had been carried out with the SIPACT simulator, based on the Cathare code, for a 900 MWe pressurized water reactor. The simulations had been mainly conducted for the Beyond Design Basis Accident (BDBA) condition. This condition included the total loss of feed-water and a small break with the loss of the high pressure injection system (HPIS). From these studies, several interesting findings had been obtained. For AM purpose and with the bleeding process, the criterion called 'the safety time margin' for core uncover was introduced. By plotting the safety time margin against the bleeding time, the relation between them was established and used to optimize, when possible, the AM measures. For the scenario that involved the total loss of feed water, in case of full bleeding, a window was found for the bleeding time around the degradation of the heat exchange in SGs would be resulted. In this scenario, one of the solutions was to open only one relief valve at first in order to let through only the minimal mass. At the time of the injection by the accumulator, the other two relief valves were then opened. As a result, the flow through the relief valves could be effectively compensated by the flow from the accumulator, the mass balance in

  5. Venus - Ishtar gravity anomaly

    Science.gov (United States)

    Sjogren, W. L.; Bills, B. G.; Mottinger, N. A.

    1984-01-01

    The gravity anomaly associated with Ishtar Terra on Venus is characterized, comparing line-of-sight acceleration profiles derived by differentiating Pioneer Venus Orbiter Doppler residual profiles with an Airy-compensated topographic model. The results are presented in graphs and maps, confirming the preliminary findings of Phillips et al. (1979). The isostatic compensation depth is found to be 150 + or - 30 km.

  6. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    Energy Technology Data Exchange (ETDEWEB)

    Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

    2016-12-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  7. Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

    International Nuclear Information System (INIS)

    Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

    2016-01-01

    Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

  8. Simulation model and methodology for calculating the damage by internal radiation in a PWR reactor; Modelo de simulacion y metodologia para el calculo del dano por irradiacion en los internos de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barreira Pereira, P.

    2012-07-01

    This study involves the development of the methodology and three-dimensional models to estimate the damage to the vessel internals of a commercial PWR reactor from irradiation history of operating cycles.

  9. Bituminization of simulated PWR type reactor wastes, boric evaporator bottons and ion exchange resins, carried out in CNEN/SP using commercial bitumen available in the Brazilian market

    International Nuclear Information System (INIS)

    Grosche Filho, C.E.; Chandra, U.

    1986-01-01

    The first results of the study of bituminization of simulated PWR wastes, boric evaporator bottons and spent ion-exclange resins (OH - , H + ) and incinerated ash-wates are presented and discussed. The study consisted of characterization of the commercial bitumen, locally available and bitumen wastes products of varying whight compositions. The characterization was carried out using standard analysis methods of ABNT and ASTM, and included measurement of, penetration, softening point and flash point. In addition, the bitumen samples were analized for their resin and asphaltene contents. For leaching studies, wastes products of bitumen and resin loaded with 134 Cs were utilized. The method used was according to the ISO norms. The simulation of the industrial process was carried out using an extruder-evaporator typically used in the plastic industries offered by Industria de Maquinas Miotto Ltda., Sao Bernardo do Campo - SP. (Author) [pt

  10. Effects of cold working ratio and stress intensity factor on intergranular stress corrosion cracking susceptibility of non-sensitized austenitic stainless steels in simulated BWR and PWR primary water

    International Nuclear Information System (INIS)

    Yaguchi, Seiji; Yonezawa, Toshio

    2012-01-01

    To evaluate the effects of cold working ratio, stress intensity factor and water chemistry on an IGSCC susceptibility of non-sensitized austenitic stainless steel, constant displacement DCB specimens were applied to SCC tests in simulated BWR and PWR primary water for the three types of austenitic stainless steels, Types 316L, 347 and 321. IGSCC was observed on the test specimens in simulated BWR and PWR primary water. The observed IGSCC was categorized into the following two types. The one is that the IGSCC observed on the same plane of the pre-fatigue crack plane, and the other is that the IGSCC observed on a plane perpendicular to the pre-fatigue crack plane. The later IGSCC fractured plane is parallel to the rolling plane of a cold rolled material. Two types of IGSCC fractured planes were changed according to the combination of the testing conditions (cold working ratio, stress intensity factor and simulated water). It seems to suggest that the most susceptible plane due to fabrication process of materials might play a significant role of IGSCC for non-sensitized cold worked austenitic stainless steels, especially, in simulated PWR primary water. Based upon evaluating on the reference crack growth rate (R-CGR) of the test specimens, the R-CGR seems to be mainly affected by cold working ratio. In case of simulated PWR primary water, it seems that the effect of metallurgical aspects dominates IGSCC susceptibility. (author)

  11. ROX PWR

    Energy Technology Data Exchange (ETDEWEB)

    Akie, H.; Yamashita, T.; Shirasu, N.; Takano, H.; Anoda, Y.; Kimura, H. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-12-01

    For an efficient burnup of excess plutonium from nuclear reactors spent fuels and dismantled warheads, plutonium rock-like oxide(ROX) fuel has been investigated. The ROX fuel is expected to provide high Pu transmutation capability, irradiation stability and chemical and geological stability. While, a zirconia-based ROX(Zr-ROX)-fueled PWR core has some problems of Doppler reactivity coefficient and power peaking factor. For the improvement of these characteristics, two approaches were considered: the additives such as UO{sub 2}, ThO{sub 2} and Er{sub 2}O{sub 3}, and a heterogeneous core with Zr-ROX and UO{sub 2} assemblies. As a result, the additives UO{sub 2}+ Er{sub 2}O{sub 3} are found to sufficiently improve the reactivity coefficients and accident behavior, and to flatten power distribution. On the other hand, in the 1/3Zr-ROX + 2/3UO{sub 2} heterogeneous core, further reduction of power peaking seems necessary. (author)

  12. Study on concentrating treatment test of simulated radioactive wastewater containing boron by reverse osmosis membrane in PWR NPP

    International Nuclear Information System (INIS)

    Ye Xinnan; Jiang Baihua; Fan Wenwen; Zhang Zhiyin; Yang Cangsheng

    2015-01-01

    The reverse osmosis membrane equipment in PWR NPP was employed to investigate the application of pilot scale system in the radioactive wastewater treatment at the full recirculation operation. The removal performance of the equipment for the boron and the radioactivity nuclide were studied, respectively. The experimental results show that the removal efficiency of the aromatic polyamide composite reverse osmosis membrane for boron is over 83.3% and the concentration of boron in concentrate is over 10000 mg/L. The experimental results also show that the removal efficiency of two nuclides including cobalt and cesium is over 97.9%. (authors)

  13. High-resolution gravity model of Venus

    Science.gov (United States)

    Reasenberg, R. D.; Goldberg, Z. M.

    1992-01-01

    The anomalous gravity field of Venus shows high correlation with surface features revealed by radar. We extract gravity models from the Doppler tracking data from the Pioneer Venus Orbiter by means of a two-step process. In the first step, we solve the nonlinear spacecraft state estimation problem using a Kalman filter-smoother. The Kalman filter has been evaluated through simulations. This evaluation and some unusual features of the filter are discussed. In the second step, we perform a geophysical inversion using a linear Bayesian estimator. To allow an unbiased comparison between gravity and topography, we use a simulation technique to smooth and distort the radar topographic data so as to yield maps having the same characteristics as our gravity maps. The maps presented cover 2/3 of the surface of Venus and display the strong topography-gravity correlation previously reported. The topography-gravity scatter plots show two distinct trends.

  14. Behavior of four PWR rods subjected to a simulated loss-of-coolant accient in the power burst facility

    International Nuclear Information System (INIS)

    Cook, T.F.; Hagrman, D.L.; Sepold, L.K.

    1978-01-01

    Cladding deformation characteristics resulting from the first nuclear blowdown tests (LOC-11) conducted in the Power Burst Facility (PBF) are emphasized in this paper. The objective of the LOC-11 tests was to obtain data on the thermal, mechanical, and materials behavior of pressurized and unpressurized fuel rods when exposed to a blowdown similiar to that expected in a pressurized water reactor (PWR) during a hypothesized double-ended cold-leg break. The test hardware consisted of four separately shrouded fresh fuel rods of PWR 15 x 15 design. Initial plenum pressures ranged from atmospheric to 4.8 MPa (representative of end-of-life). During LOC-11C, the four fuel rods were subjected to 6.5 hours of nuclear operation at approximately 67 kW/m average rod power to cause decay heat build-up. Just before the start of blowdown, cladding surface temperatures were about 620 K and fuel centerline temperatures were in the 2500 to 2600 K range. During the 30-second blowdown transient, CHF occurred 2 seconds after initiation. Fuel centerline temperature dropped continuously, while cladding surface temperatures increased. Maximum cladding temperatures of 1030 to 1050 K occurred 15 seconds into the transient. Posttest destructive examination revealed cladding microstructures and oxide thicknesses consistent with the measured cladding temperatures. The cladding surface thermocouples did not appreciably affect cladding temperature distributuion (fin cooling effect) in the vicinity of the thermocouples

  15. Retarding effect of prior-overloading on stress corrosion cracking of cold rolled 316L SS in simulated PWR water environment

    Science.gov (United States)

    Chen, Junjie; Lu, Zhanpeng; Xiao, Qian; Ru, Xiangkun; Ma, Jiarong; Shoji, Tetsuo

    2017-12-01

    The effect of prior single tensile overloading on the stress corrosion cracking behavior of cold rolled 316L in a simulated PWR water environment at 310 °C was investigated. SCC growth retardation by overloading was observed in cold rolled 316L specimens in both the T-L and L-T orientations. The stretch zone observed on the fracture surfaces of the overloaded specimens affected SCC propagation. The compressive residual stress induced by overloading process reduced the effective driving force of SCC propagation. The negative dK/da effect ahead of the crack tip likely contributes to the retardation of SCC growth. The duration of overloading is dependent on water chemistry and the local stress conditions.

  16. Effect of temperature and dissolved hydrogen on oxide films formed on Ni and Alloy 182 in simulated PWR water

    International Nuclear Information System (INIS)

    Mendonça, R.; Bosch, R.-W.; Van Renterghem, W.; Vankeerberghen, M.; Araújo Figueiredo, C. de

    2016-01-01

    Alloy 182 is a nickel-based weld metal, which is susceptible to stress corrosion cracking in PWR primary water. It shows a peak in SCC susceptibility at a certain temperature and hydrogen concentration. This peak is related to the electrochemical condition where the Ni to NiO transition takes place. One hypothesis is that the oxide layer at this condition is not properly developed and so the material is not optimally protected against SCC. Therefore the oxide layer formed on Alloy 182 is investigated as a function of the dissolved hydrogen concentration and temperature around this Ni/NiO transition. Exposure tests were performed with Alloy 182 and Ni coupons in a PWR environment at temperatures between 300 °C and 345 °C and dissolved hydrogen concentration between 5 and 35 cc (STP)H 2 /kg. Post-test analysis of the formed oxide layers were carried out by SEM, EDS and XPS. The exposure tests with Ni coupons showed that the Ni/NiO transition curve is at a higher temperature than the curve based on thermodynamic calculations. The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures, but that the morphology changed from spinel crystals to needle like oxides when the Ni/NiO transition curve was approached. Oxide layers were present below the Ni/NiO transition curve i.e. when the Ni coupon was still free of oxides. In addition an evolved slip dissolution model was proposed that could explain the observed experimental results and the peak in SCC susceptibility for Ni-based alloys around the Ni/NiO transition. - Highlights: • Exposure tests with Ni-coupons showed that the Ni/NiO transition curve shifted to more oxidizing conditions. • The Ni specimens tested in PWR water were free of oxides at all temperatures. • The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures. • The Alloy 182 surface morphology changed from spinel crystals to needle like oxides when the Ni/NiO curve was approached

  17. Effect of temperature and dissolved hydrogen on oxide films formed on Ni and Alloy 182 in simulated PWR water

    Energy Technology Data Exchange (ETDEWEB)

    Mendonça, R. [CAPES Foundation, Ministry of Education, Brasilia (Brazil); Bosch, R.-W., E-mail: rbosch@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Van Renterghem, W.; Vankeerberghen, M. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Araújo Figueiredo, C. de [CDTN/CNEN, Av. Antônio Carlos 6627, 31270-901 Belo Horizonte, MG (Brazil)

    2016-08-15

    Alloy 182 is a nickel-based weld metal, which is susceptible to stress corrosion cracking in PWR primary water. It shows a peak in SCC susceptibility at a certain temperature and hydrogen concentration. This peak is related to the electrochemical condition where the Ni to NiO transition takes place. One hypothesis is that the oxide layer at this condition is not properly developed and so the material is not optimally protected against SCC. Therefore the oxide layer formed on Alloy 182 is investigated as a function of the dissolved hydrogen concentration and temperature around this Ni/NiO transition. Exposure tests were performed with Alloy 182 and Ni coupons in a PWR environment at temperatures between 300 °C and 345 °C and dissolved hydrogen concentration between 5 and 35 cc (STP)H{sub 2}/kg. Post-test analysis of the formed oxide layers were carried out by SEM, EDS and XPS. The exposure tests with Ni coupons showed that the Ni/NiO transition curve is at a higher temperature than the curve based on thermodynamic calculations. The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures, but that the morphology changed from spinel crystals to needle like oxides when the Ni/NiO transition curve was approached. Oxide layers were present below the Ni/NiO transition curve i.e. when the Ni coupon was still free of oxides. In addition an evolved slip dissolution model was proposed that could explain the observed experimental results and the peak in SCC susceptibility for Ni-based alloys around the Ni/NiO transition. - Highlights: • Exposure tests with Ni-coupons showed that the Ni/NiO transition curve shifted to more oxidizing conditions. • The Ni specimens tested in PWR water were free of oxides at all temperatures. • The exposure tests with Alloy 182 showed that oxide layers were present at all temperatures. • The Alloy 182 surface morphology changed from spinel crystals to needle like oxides when the Ni/NiO curve was

  18. Venus Landsailing Rover

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA Glenn has developed electronics and low-power photovoltaics that will continue to function even at the Venus temperature of 450°C. So the fundamental elements...

  19. Venus: Our Misunderstood Sister

    Science.gov (United States)

    Dyar, Darby; Smrekar, Suzanne E.

    2018-01-01

    Of all known bodies in the galaxy, Venus is the most Earth-like in size, composition, surface age, and incoming energy. As we search for habitable planets around other stars, learning how Venus works is critical to understanding how Earth evolved to host life, and whether rocky exoplanets in stars’ habitable zones are faraway Earths or Venuses. What caused Venus’ path to its present hostile environment, devoid of oceans, magnetic field, and plate tectonics? This talk reviews recent mission results, presents key unresolved science questions, and describes proposed missions to answer these questions.Despite its importance in understanding habitability, Venus is the least-explored rocky planet, last visited by NASA in 1994. Fundamental, unanswered questions for Venus include: 1. How did Venus evolve differently? 2. How have volatiles shaped its evolution? 3. Did Venus catastrophically resurface? 4. What geologic processes are active today? 5. Why does Venus lack plate tectonics?On Earth, plate tectonics supports long-term climate stability and habitability by cycling volatiles in and out of the mantle. New information on planetary volatiles disputes the long-held notion that Venus’ interior is dry; several lines of evidence indicate that planets start out wet, creating long-term atmospheres by outgassing. ESA’s Venus Express mission provided evidence for recent and ongoing volcanism and for Si-rich crust like Earth’s continents. New hypotheses suggest that lithospheric temperature can explain why Venus lacks tectonics, and are consistent with present-day initiation of subduction on Venus.New data are needed to answer these key questions of rocky planet evolution. Orbital IR data can be acquired through windows in Venus’ CO2-rich atmosphere, informing surface mineralogy, rock types, cloud variations, and active volcanism. High resolution gravity, radar, and topography data along with mineralogical constraints must be obtained. Mineralogy and geochemistry

  20. Neutron Fluence, Dosimetry and Damage Response Determination in In-Core/Ex-Core Components of the VENUS CEN/SCK LWR Using 3-D Monte Carlo Simulations: NEA's VENUS-3 Benchmark

    International Nuclear Information System (INIS)

    Perlado, J. Manuel; Marian, Jaime; Sanz, Jesus Garcia

    2000-01-01

    Validating state-of-the-art methods used to predict fluence exposure to reactor pressure vessels (RPVs) has become an important issue in identifying the sources of uncertainty in the estimated RPV fluence for pressurized water reactors. This is a very important aspect in evaluating irradiation damage leading to the hardening and embrittlement of such structural components. One of the major benchmark experiments carried out to test three-dimensional methodologies is the VENUS-3 Benchmark Experiment in which three-dimensional Monte Carlo and S n codes have proved more efficient than synthesis methods. At the Instituto de Fusion Nuclear (DENIM) at the Universidad Politecnica de Madrid, a detailed full three-dimensional model of the Venus Critical Facility has been developed making use of the Monte Carlo transport code MCNP4B. The problem geometry and source modeling are described, and results, including calculated versus experimental (C/E) ratios as well as additional studies, are presented. Evidence was found that the great majority of C/E values fell within the 10% tolerance and most within 5%. Tolerance limits are discussed on the basis of evaluated data library and fission spectra sensitivity, where a value ranging between 10 to 15% should be accepted. Also, a calculation of the atomic displacement rate has been carried out in various locations throughout the reactor, finding that values of 0.0001 displacements per atom in external components such as the core barrel are representative of this type of reactor during a 30-yr time span

  1. Venus Elongation Measurements for the Transit of Venus, using the ...

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 9; Issue 11. Venus Elongation Measurements for the Transit of Venus, using the Historical Jantar Mantar Observatory. N Rathnasree. Classroom Volume 9 Issue 11 November 2004 pp 46-55 ...

  2. Long-term in situ corrosion investigation of Zr alloys in simulated PWR environment by electrochemical measurements

    International Nuclear Information System (INIS)

    Goehr, H.; Schaller, J.; Ruhmann, H.; Garzarolli, F.

    1996-01-01

    The corrosion behavior of Zircaloy-type alloys with different tin contents of 1.55, 0.70, and 0.55 wt% was studied at 350 C and 17 MPa in an environment similar to PWR primary water. Impedance spectra were taken at time intervals and evaluated for thickness and morphology of the oxide layer as well as for its electrical resistance. The tests without any temperature and pressure cycling showed similar oxidation behavior with repeated transitions as in discontinuously performed standard autoclave tests. Early in the pre-transition range, a dense oxide layer is formed, and fast changes of corrosion potential and electrical resistance are observed. The dense layer increases in thickness and homogeneity up to the transition, where a sudden breakdown occurs. Abrupt changes of the corrosion potential and electrical resistance were observed also at those points. After transition, a new dense layer is built up. The corrosion potential changes are caused by a decrease of the electrical corrosion current with increasing oxide layer thickness, by the formation of a potential drop over the high-resistance dense oxide layer, and by structural changes at the transition points. In general, alloys with different tin contents show similar behavior. However, they show differences in the time to transition, the kinetic constants deduced from their impedance spectra, and in the ionic and electronic resistance of the dense inner layer controlling corrosion

  3. Maturity of the PWR

    International Nuclear Information System (INIS)

    Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

    1983-03-01

    Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

  4. Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

    1987-03-01

    This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO 2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL) [de

  5. Performance of core exit thermocouple for PWR accident management action in vessel top break LOCA simulation experiment at OECD/NEA ROSA project

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Takeda, Takeshi; Nakamura, Hideo

    2009-01-01

    Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary side in case of core temperature excursion. Test 6-1 is first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reason of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection. (author)

  6. Numerical simulations of counter-current two-phase flow experiments in a PWR hot leg model using an interfacial area density model

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, Thomas, E-mail: t.hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany); Deendarlianto,; Lucas, Dirk [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Safety Research, P.O. Box 510 119, D-01314 Dresden (Germany)

    2011-10-15

    In order to improve the understanding of counter-current two-phase flows and to validate new physical models, CFD simulations of 1/3rd scale model of the hot leg of a German Konvoi PWR with rectangular cross section was performed. Selected counter-current flow limitation (CCFL) experiments at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) were calculated with ANSYS CFX 12.1 using the multi-fluid Euler-Euler modeling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a k-{omega} turbulence model for each phase. In the simulation, the surface drag was approached by a new correlation inside the Algebraic Interfacial Area Density (AIAD) model. The AIAD model allows the detection of the morphological form of the two phase flow and the corresponding switching via a blending function of each correlation from one object pair to another. As a result this model can distinguish between bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicated that quantitative agreement of the CCFL characteristics between calculation and experimental data was obtained. The goal is to provide an easy usable AIAD framework for all Code users, with the possibility of the implementation of their own correlations.

  7. Numerical simulations of counter-current two-phase flow experiments in a PWR hot leg model using an interfacial area density model

    Energy Technology Data Exchange (ETDEWEB)

    Hohne, T.; Deendarlianto; Vallee, C.; Lucas, D.; Beyer, M., E-mail: t.hoehne@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Inst. of Safety Research, Dresden (Germany)

    2011-07-01

    In order to improve the understanding of counter-current two-phase flows and to validate new physical models, CFD simulations of 1/3rd scale model of the hot leg of a German Konvoi PWR with rectangular cross section was performed. Selected counter-current flow limitation (CCFL) experiments at the Helmholtz-Zentrum Dresden- Rossendorf (HZDR) were calculated with ANSYS CFX 12.1 using the multi-fluid Euler-Euler modeling approach. The transient calculations were carried out using a gas/liquid inhomogeneous multiphase flow model coupled with a SST turbulence model for each phase. In the simulation, the surface drag was approached by a new correlation inside the Algebraic Interfacial Area Density (AIAD) model. The AIAD model allows the detection of the morphological form of the two phase flow and the corresponding switching via a blending function of each correlation from one object pair to another. As a result this model can distinguish between bubbles, droplets and the free surface using the local liquid phase volume fraction value. A comparison with the high-speed video observations shows a good qualitative agreement. The results indicated that quantitative agreement of the CCFL characteristics between calculation and experimental data was obtained. The goal is to provide an easy usable AIAD framework for all ANSYS CFX users, with the possibility of the implementation of their own correlations. (author)

  8. Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor; Modelo simplificado para simulacao do comportamento termohidraulico do canal quente de reator nuclear do tipo PWR

    Energy Technology Data Exchange (ETDEWEB)

    Belem, J A.T.

    1993-09-01

    The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.

  9. Clouds of Venus

    Energy Technology Data Exchange (ETDEWEB)

    Knollenberg, R G [Particle Measuring Systems, Inc., 1855 South 57th Court, Boulder, Colorado 80301, U.S.A.; Hansen, J [National Aeronautics and Space Administration, New York (USA). Goddard Inst. for Space Studies; Ragent, B [National Aeronautics and Space Administration, Moffett Field, Calif. (USA). Ames Research Center; Martonchik, J [Jet Propulsion Lab., Pasadena, Calif. (USA); Tomasko, M [Arizona Univ., Tucson (USA)

    1977-05-01

    The current state of knowledge of the Venusian clouds is reviewed. The visible clouds of Venus are shown to be quite similar to low level terrestrial hazes of strong anthropogenic influence. Possible nucleation and particle growth mechanisms are presented. The Pioneer Venus experiments that emphasize cloud measurements are described and their expected findings are discussed in detail. The results of these experiments should define the cloud particle composition, microphysics, thermal and radiative heat budget, rough dynamical features and horizontal and vertical variations in these and other parameters. This information should be sufficient to initialize cloud models which can be used to explain the cloud formation, decay, and particle life cycle.

  10. Numerical analysis and simulation of behavior of high burn-up PWR fuel pulse-irradiated in reactivity-initiated accident conditions

    International Nuclear Information System (INIS)

    Suzuki, M.; Sugiyama, T.; Udagawa, Y.; Nagase, F.; Fuketa, T.

    2010-01-01

    The four cases of the NSRR experiments, consisting of two room temperature tests and two high temperature tests, using high burn-up PWR fuel rods are analyzed by using the RANNS code to discuss the fuel behavior in hypothetical pulse-irradiation conditions, and the results are compared with metallography observations of ruptured claddings. The cladding rupture occurred by a shear sliding which starts from the tip of incipient crack generated in the hydride dense layer. The analyses reveal that the onset of shear sliding leading to cladding rupture can be closely associated with the stress intensity factor KI at the crack tip and local plastic strain evolution around the tip as well, and that these two factors depend also on the temperature of cladding. Simulation calculations on the basis of experimental conditions reveals that the cladding stress is dependent on the height and half-width of pulse power, and for the same integral enthalpy of pulse a larger half-width mitigates the severity of transient and decreases KI to allow plastic strain by temperature rise, thus failure possibility would be markedly decreased

  11. Stress intensity factor at the tip of cladding incipient crack in RIA-simulating experiments for high-burnup PWR fuels

    International Nuclear Information System (INIS)

    Udagawa, Yutaka; Suzuki, Motoe; Sugiyama, Tomoyuki; Fuketa, Toyoshi

    2009-01-01

    RIA-simulating experiments for high-burnup PWR fuels have been performed in the NSRR, and the stress intensity factor K 1 at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold under RIA conditions. An incipient crack depth was determined by observation of metallographs. The maximum hydride-rim thickness in the cladding of the test fuel rod was regarded as the incipient crack depth in each test case. Hoop stress in the cladding periphery during the pulse power transient was calculated by the RANNS code. K 1 was calculated based on crack depth and hoop stress. According to the RANNS calculation, PCMI failure cases can be divided into two groups: failure in the elastic phase and failure in the plastic phase. In the former case, elastic deformation was predominant around the incipient crack at failure time. K 1 is available only in this case. In the latter, plastic deformation was predominant around the incipient crack at failure time. Failure in the elastic phase never occurred when K 1 was less than 17 MPa m 1/2 . For failure in the plastic phase, the plastic hoop strain of the cladding periphery at failure time clearly showed a tendency to decrease with incipient crack depth. The combination of K 1 , for failure in the elastic phase, and plastic hoop strain at failure, for failure in the plastic phase, can be an effective index of PCMI failure under RIA conditions. (author)

  12. Pre-test prediction and post-test analysis of PWR fuel rod ballooning in the MT-3 in-pile LOCA simulation experiment in the NRU reactor

    International Nuclear Information System (INIS)

    Donaldson, A.T.; Horwood, R.A.; Healey, T.

    1983-01-01

    The USNRC and the UKAEA have jointly funded a series of in-pile LOCA simulation experiments in the Canadian NRU reactor in order to secure further information on the thermal hydraulic and clad deformation response of PWR fuel rod bundles. Test MT-3 in the series was performed using reflood rate and rod internal pressure conditions specified by the UK nuclear industry. The parameters were selected to ensure the development of a near-isothermal clad temperature history during which zircaloy was required to balloon and rupture near the alpha-alpha/beta phase transition. Specification of the reflood rate conditions was assisted by the performance of a precursor test on an unpressurised rod bundle and by complementary application of appropriate thermal hydraulic analyses. Identification of the rod internal pressure needed to cause ballooning and rupture was achieved using a creep deformation model, BALLOON, in conjunction with the clad thermal history defined by the prior thermal hydraulic test. This paper presents the basis of the BALLOON analysis and describes its application in calculating the fill gas pressure for rods MT-3, their axial ballooning profile and the clad temperature at peak radial strain elevations. (author)

  13. Geology of Venus

    International Nuclear Information System (INIS)

    Basilevsky, A.T.; Head, J.W. III.

    1988-01-01

    This paper summarizes the emerging picture of the surface of Venus provided by high-resolution earth-based radar telescopes and orbital radar altimetry and imaging systems. The nature and significance of the geological processes operating there are considered. The types of information needed to complete the picture are addressed. 71 references

  14. The effects of cold rolling orientation and water chemistry on stress corrosion cracking behavior of 316L stainless steel in simulated PWR water environments

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Junjie [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Lu, Zhanpeng, E-mail: zplu@t.shu.edu.cn [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Xiao, Qian; Ru, Xiangkun; Han, Guangdong; Chen, Zhen [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); Zhou, Bangxin [Institute of Materials Science, School of Materials Science and Engineering, Shanghai University, Mailbox 269, 149 Yanchang Road, Shanghai, 200072 (China); State Key Laboratory of Advanced Special Steels, Shanghai University, 149 Yanchang Road, Shanghai, 200072 (China); Shoji, Tetsuo [New Industry Creation Hatchery Center, Tohoku University, Sendai 980-8579 (Japan)

    2016-04-15

    Stress corrosion cracking behaviors of one-directionally cold rolled 316L stainless steel specimens in T–L and L–T orientations were investigated in hydrogenated and deaerated PWR primary water environments at 310 °C. Transgranular cracking was observed during the in situ pre-cracking procedure and the crack growth rate was almost not affected by the specimen orientation. Locally intergranular stress corrosion cracks were found on the fracture surfaces of specimens in the hydrogenated PWR water. Extensive intergranular stress corrosion cracks were found on the fracture surfaces of specimens in deaerated PWR water. More extensive cracks were found in specimen T–L orientation with a higher crack growth rate than that in the specimen L–T orientation with a lower crack growth rate. Crack branching phenomenon found in specimen L–T orientation in deaerated PWR water was synergistically affected by the applied stress direction as well as the preferential oxidation path along the elongated grain boundaries, and the latter was dominant. - Highlights: • Transgranular fatigue crack growth rate was not affected by the cold rolling orientation. • Locally intergranular SCC was found in the hydrogenated PWR water. • Extensive intergranular SCC cracks were found in deaerated PWR water. • T–L specimen showed more extensive SCC cracks and a higher crack growth rate. • Crack branching related to the applied stress and the preferential oxidation path.

  15. Development of BWR [boiling water reactor] and PWR [pressurized water reactor] event descriptions for nuclear facility simulator training

    International Nuclear Information System (INIS)

    Carter, R.J.; Bovell, C.R.

    1987-01-01

    A number of tools that can aid nuclear facility training developers in designing realistic simulator scenarios have been developed. This paper describes each of the tools, i.e., event lists, events-by-competencies matrices, and event descriptions, and illustrates how the tools can be used to construct scenarios

  16. 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)

  17. Lightning on Venus

    Science.gov (United States)

    Scarf, F. L.

    1985-01-01

    On the night side of Venus, the plasma wave instrument on the Pioneer-Venus Orbiter frequently detects strong and impulsive low-frequency noise bursts when the local magnetic field is strong and steady and when the field is oriented to point down to the ionosphere. The signals have characteristics of lightning whistlers, and an attempt was made to identify the sources by tracing rays along the B-field from the Orbiter down toward the surface. An extensive data set strongly indicates a clustering of lightning sources near the Beta and Phoebe Regios, with additional significant clustering near the Atla Regio at the eastern edge of Aphrodite Terra. These results suggest that there are localized lightning sources at or near the planetary surface.

  18. Development of nuclear power plant monitoring system with neutral network using on-line PWR plant simulator

    International Nuclear Information System (INIS)

    Nabeshima Kunihiko; Suzuki Katsuo; Nose, Shoichi; Kudo, Kazuhiko

    1996-01-01

    The purpose of this paper is to demonstrate a nuclear power plant monitoring system using artificial neural network (ANN). The major advantages of the monitoring system are that a multi-output process system can be modelled using measurement information without establishing any mathematical expressions. The dynamics model of reactor plant was constructed by using three layered auto-associative neural network with backpropagation learning algorithm. The basic idea of anomaly detection method is to monitor the deviation between process signals measured from actual plant and corresponding output signals from the ANN plant model. The simulator used is a self contained system designed for training. Four kinds of simulated malfunction caused by equipment failure during steady state operation were used to evaluate the capability of the neural network monitoring system. The results showed that this monitoring system detected the symptom of small anomaly earlier than the prevailing alarm system. (author). 7 refs, 7 figs, 2 tabs

  19. Development of nuclear power plant monitoring system with neutral network using on-line PWR plant simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kunihiko, Nabeshima; Katsuo, Suzuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Nose, Shoichi; Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-12-31

    The purpose of this paper is to demonstrate a nuclear power plant monitoring system using artificial neural network (ANN). The major advantages of the monitoring system are that a multi-output process system can be modelled using measurement information without establishing any mathematical expressions. The dynamics model of reactor plant was constructed by using three layered auto-associative neural network with backpropagation learning algorithm. The basic idea of anomaly detection method is to monitor the deviation between process signals measured from actual plant and corresponding output signals from the ANN plant model. The simulator used is a self contained system designed for training. Four kinds of simulated malfunction caused by equipment failure during steady state operation were used to evaluate the capability of the neural network monitoring system. The results showed that this monitoring system detected the symptom of small anomaly earlier than the prevailing alarm system. (author). 7 refs, 7 figs, 2 tabs.

  20. PWR simplified fuel element simulation using calculation trailer ANSYS CFX and PARCS including pressure drop and turbulence in the spacer

    International Nuclear Information System (INIS)

    Pena-Monferrer, C.; Chiva, S.; Miro, R.; Barrachina, T.; Pellacani, F.; Macian-Juan, R.

    2012-01-01

    With the recent development of a new computational tool for calculations of nuclear reactors based on the coupling between the PARCS neutron transport code and computational fluid dynamics commercial code (CFD) ANSYS CFX opens new possibilities in the fuel element design that contributes to a better understanding and a better simulation of the processes of heat transfer and specific phenomena of fluid dynamics as the c rossflow .

  1. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  2. Uncertainty and Sensitivity of Neutron Kinetic Parameters in the Dynamic Response of a PWR Rod Ejection Accident Coupled Simulation

    Directory of Open Access Journals (Sweden)

    C. Mesado

    2012-01-01

    Full Text Available In nuclear safety analysis, it is very important to be able to simulate the different transients that can occur in a nuclear power plant with a very high accuracy. Although the best estimate codes can simulate the transients and provide realistic system responses, the use of nonexact models, together with assumptions and estimations, is a source of uncertainties which must be properly evaluated. This paper describes a Rod Ejection Accident (REA simulated using the coupled code RELAP5/PARCSv2.7 with a perturbation on the cross-sectional sets in order to determine the uncertainties in the macroscopic neutronic information. The procedure to perform the uncertainty and sensitivity (U&S analysis is a sampling-based method which is easy to implement and allows different procedures for the sensitivity analyses despite its high computational time. DAKOTA-Jaguar software package is the selected toolkit for the U&S analysis presented in this paper. The size of the sampling is determined by applying the Wilks’ formula for double tolerance limits with a 95% of uncertainty and with 95% of statistical confidence for the output variables. Each sample has a corresponding set of perturbations that will modify the cross-sectional sets used by PARCS. Finally, the intervals of tolerance of the output variables will be obtained by the use of nonparametric statistical methods.

  3. The various contributions in Venus rotation rate and LOD

    Science.gov (United States)

    Cottereau, L.; Rambaux, N.; Lebonnois, S.; Souchay, J.

    2011-07-01

    Context. Thanks to the Venus Express Mission, new data on the properties of Venus could be obtained, in particular concerning its rotation. Aims: In view of these upcoming results, the purpose of this paper is to determine and compare the major physical processes influencing the rotation of Venus and, more particularly, the angular rotation rate. Methods: Applying models already used for Earth, the effect of the triaxiality of a rigid Venus on its period of rotation are computed. Then the variations of Venus rotation caused by the elasticity, the atmosphere, and the core of the planet are evaluated. Results: Although the largest irregularities in the rotation rate of the Earth on short time scales are caused by its atmosphere and elastic deformations, we show that the irregularities for Venus are dominated by the tidal torque exerted by the Sun on its solid body. Indeed, as Venus has a slow rotation, these effects have a large amplitude of two minutes of time (mn). These variations in the rotation rate are greater than the one induced by atmospheric wind variations that can reach 25-50 s of time (s), depending on the simulation used. The variations due to the core effects that vary with its size between 3 and 20 s are smaller. Compared to these effects, the influence of the elastic deformation caused by the zonal tidal potential is negligible. Conclusions: As the variations in the rotation of Venus reported here are close to 3 mn peak to peak, they should influence past, present, and future observations, thereby providing further constraints on the planet's internal structure and atmosphere.

  4. Lunar and Planetary Science XXXV: Venus

    Science.gov (United States)

    2004-01-01

    The session "Venus" included the following reports:Preliminary Study of Laser-induced Breakdown Spectroscopy (LIBS) for a Venus Mission; Venus Surface Investigation Using VIRTIS Onboard the ESA/Venus Express Mission; Use of Magellan Images for Venus Landing Safety Assessment; Volatile Element Geochemistry in the Lower Atmosphere of Venus; Resurfacing Styles and Rates on Venus: Assessment of 18 Venusian Quadrangles; Stereo Imaging of Impact Craters in the Beta-Atla-Themis (BAT) Region, Venus; Depths of Extended Crater-related Deposits on Venus ; Potential Pyroclastic Deposit in the Nemesis Tessera (V14) Quadrangle of Venus; Relationship Between Coronae, Regional Plains and Rift Zones on Venus, Preliminary Results; Coronae of Parga Chasma, Venus; The Evolution of Four Volcano/Corona Hybrids on Venus; Calderas on Venus and Earth: Comparison and Models of Formation; Venus Festoon Deposits: Analysis of Characteristics and Modes of Emplacement; Topographic and Structural Analysis of Devana Chasma, Venus: A Propagating Rift System; Anomalous Radial Structures at Irnini Mons, Venus: A Parametric Study of Stresses on a Pressurized Hole; Analysis of Gravity and Topography Signals in Atalanta-Vinmara and Lavinia Planitiae Canali are Lava, Not River, Channels; and Formation of Venusian Channels in a Shield Paint Substrate.

  5. Simulation of accident and restrained transients in PWR nuclear power plant with RELAP 5/MOD 1 computer code

    International Nuclear Information System (INIS)

    Silva Filho, E.

    1986-01-01

    The computer code RELAP5/MOD1 has been utilized to investigate the thermal-hydraulic behaviour of a standard 1300 Mwe pressurized water reactor plant of the KWU design during a station blackout and during a loss-of-coolant accident involving 2% break in the cross-sectional area the cold leg in one of the four loops and located between the pump and the reactor pressure vessel. During the simulations the reactor scram system and the emergency coolant system were considered inactive. (Author) [pt

  6. Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction ESSI-4 ESSI-11

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauscheck, H.; Wallenfels, K.P.; Buescher, B.J.

    1985-03-01

    The tests had the initial heatup rate as main parameter. The experimental arrangement consisted of a fuel rod simulator (central tungsten heater, UO 2 ring pellets and zircaloy cladding), a zircaloy shroud and the fiber ceramic insulation. A steam flow of ca. 20 g/min was introduced at the lower end of the bundle. A temperature escalation was observed in every test. The maximum cladding surface temperature in the single rod tests never exceeded 2200 0 C. The escalation began in the upper region of the rods and moved down the rods, opposite to the direction of steam flow. For fast initial heatup rates, the runoff of molten zircaloy was a limiting process for the escalation. For slow heatup rates, the formation of a protective oxide layer reduced the reaction rate. The test with less insulation thickness showed a reduction of the escalation. A stronger influence was found for the gap between shroud and insulation. This is caused by convection heat losses to the steam circulating in this gap by natural convection. Removal of the gap between shroud and insulation in essentially the same experimental arrangement produced a faster escalation. The posttest appearance of the fuel rod simulators showed that, at slow heatup rates oxidation of the cladding was complete, and the fuel rod was relatively intact. Conversely, at fast heatup rates, relatively little cladding oxidation with extensive dissolution of the UO 2 pellets and runoff of molten cladding was observed. (orig./HP) [de

  7. Thermohydraulic calculations of PWR primary circuits

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1984-01-01

    Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

  8. The Plains of Venus

    Science.gov (United States)

    Sharpton, V. L.

    2013-12-01

    Volcanic plains units of various types comprise at least 80% of the surface of Venus. Though devoid of topographic splendor and, therefore often overlooked, these plains units house a spectacular array of volcanic, tectonic, and impact features. Here I propose that the plains hold the keys to understanding the resurfacing history of Venus and resolving the global stratigraphy debate. The quasi-random distribution of impact craters and the small number that have been conspicuously modified from the outside by plains-forming volcanism have led some to propose that Venus was catastrophically resurfaced around 725×375 Ma with little volcanism since. Challenges, however, hinge on interpretations of certain morphological characteristics of impact craters: For instance, Venusian impact craters exhibit either radar dark (smooth) floor deposits or bright, blocky floors. Bright floor craters (BFC) are typically 100-400 m deeper than dark floor craters (DFC). Furthermore, all 58 impact craters with ephemeral bright ejecta rays and/or distal parabolic ejecta patterns have bright floor deposits. This suggests that BFCs are younger, on average, than DFCs. These observations suggest that DFCs could be partially filled with lava during plains emplacement and, therefore, are not strictly younger than the plains units as widely held. Because the DFC group comprises ~80% of the total crater population on Venus the recalculated emplacement age of the plains would be ~145 Ma if DFCs are indeed volcanically modified during plains formation. Improved image and topographic data are required to measure stratigraphic and morphometric relationships and resolve this issue. Plains units are also home to an abundant and diverse set of volcanic features including steep-sided domes, shield fields, isolated volcanoes, collapse features and lava channels, some of which extend for 1000s of kilometers. The inferred viscosity range of plains-forming lavas, therefore, is immense, ranging from the

  9. Sensitivity analysis for thermo-hydraulics model of a Westinghouse type PWR. Verification of the simulation results

    Energy Technology Data Exchange (ETDEWEB)

    Farahani, Aref Zarnooshe [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Yousefpour, Faramarz [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Hoseyni, Seyed Mohsen [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Basic Sciences; Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Young Researchers and Elite Club

    2017-07-15

    Development of a steady-state model is the first step in nuclear safety analysis. The developed model should be qualitatively analyzed first, then a sensitivity analysis is required on the number of nodes for models of different systems to ensure the reliability of the obtained results. This contribution aims to show through sensitivity analysis, the independence of modeling results to the number of nodes in a qualified MELCOR model for a Westinghouse type pressurized power plant. For this purpose, and to minimize user error, the nuclear analysis software, SNAP, is employed. Different sensitivity cases were developed by modification of the existing model and refinement of the nodes for the simulated systems including steam generators, reactor coolant system and also reactor core and its connecting flow paths. By comparing the obtained results to those of the original model no significant difference is observed which is indicative of the model independence to the finer nodes.

  10. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Peck, S.O.; Wallenfels, K.P.

    1983-12-01

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO 2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO 2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 2 0 C/s at 1100 0 C increased to approximately 6 0 C/s. The maximum temperature reached was 2250 0 C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO 2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO 2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.) [de

  11. Temperature escalation in PWR fuel rod simulator bundles due to the Zircaloy/steam reaction: Test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1984-07-01

    This report describes the test conduct and results of the bundle test ESBU-2A, which was run to investigate the temperature escalation of zircaloy clad fuel rods. This investigation of temperature escalation is part of a series of out-of-pile experiments, performed within the framework of the PNS Severe Fuel Damage Program. The test bundle was of a 3 x 3 array of fuel rod simulators with a 0.4 m heated length. The fuel rod simulators were electrically heated and consisted of tungsten heaters, UO 2 annular pellets, and zircaloy cladding. A nominal steam flow of 0.7 g/s was inlet to the bundle. The bundle was surrounded by a zircaloy shroud which was insulated with ZrO 2 fiber ceramic wrap. The initial heatup rate of the bundle was 0.4 0 C/s. The temperature escalation began at the 255 mm elevation after 1200 0 C had been reached. At this elevation, the measured peak temperature was limited to 1500 0 C. It was concluded from different thermocouple results, that induced by this first escalation melt was formed in the lower part of the bundle. Consequently, the escalation in the lower part must be much higher, at least up to the melting temperature of zircaloy. Due to the failure in the steam production system, steam starvation in the upper region may explain the beginning of the escalation at the 255 mm elevation. The maximum temperature reached was 2175 0 C on the center rod at the end of the test. The unregularities in the steam supply may be the reason for less oxidation than expected. (orig./GL) [de

  12. Parallel GPU implementation of PWR reactor burnup

    International Nuclear Information System (INIS)

    Heimlich, A.; Silva, F.C.; Martinez, A.S.

    2016-01-01

    Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

  13. Stress corrosion cracking growth rate of TT alloy 690 and its weld joint in simulated PWR primary water

    International Nuclear Information System (INIS)

    Yonezawa, T.

    2015-01-01

    Recently, some researchers reported that the SCC growth rate (SCCGR) of cold worked thermally treated (TT) Alloy 690 was significantly different in heat by heat. But, author has hypothesized that these high SCCGRs in cold worked TT Alloy 690 could be due to the metallurgical characteristics of these heats. In order to confirm this hypothesis, this study has been started in the author's laboratory, and the following 4 new evidences were obtained. First, microcracks of carbides and voids were observed in eutectic M 23 C 6 GB carbides (primary carbides) for cold rolled laboratory heat after as cast or lightly forged condition or for chemical composition simulated Bettis'TT Alloy 690 heat, after cold rolling, before SCC test. However, microcracks in primary carbides along grain boundaries and voids were rarely detected in the cold rolled commercial heat of TT Alloy 690 used for CRDM penetrations. Secondly, the SCCGR observed in TT Alloy 690 was different in each hot working process and each heat. Comparing the SCCGRs for all heats of cold worked TT Alloy 690, the SCCGR decreased with increasing of Vickers hardness. However, in same heats of cold worked TT Alloy 690, the SCCGR increased with increasing of Vickers hardness. Thirdly, the SCCGR in cold rolled TT Alloy 690 should be integrated by the effect of hardness or cold working ratio and by the effect of existing ratio of primary M23C6 carbides with cracks and Voids due to chemical composition and the fabrication process of TT Alloy 690. Fourthly, it is argued that the high SCCGRs in highly cold rolled TT Alloy 690 are not representative of the practical situation with TT Alloy 690 in service for CRDM adapter nozzles etc. The high SCCGR of highly cold rolled TT Alloy 690 is not thought to be an accurate tool in predicting the possibility of cracking of TT Alloy 690 for CRDM adapter nozzles. (author)

  14. VENUS Ranging Study

    Science.gov (United States)

    2014-12-01

    Majesté la Reine (en droit du Canada), telle que réprésentée par le ministre de la Défense nationale, 2014 Abstract The underwater acoustic propagation...50 km des capteurs sous-marins situés aux nœuds du réseau VENUS dont les données acoustiques et sismiques sont accessibles au public sur Internet...Southwest British Columbia, Geophysical Journal International , 170(2), 800–812. [15] Hamilton, E. L. (1979), Vp/Vs and Poisson’s ratios in marine

  15. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  16. Next generation PWR

    International Nuclear Information System (INIS)

    Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

    2001-01-01

    Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

  17. Venus - Phoebe Region

    Science.gov (United States)

    1990-01-01

    This Magellan radar image is of part of the Phoebe region of Venus. It is a mosaic of parts of revolutions 146 and 147 acquired in the first radar test on Aug. 16, 1990. The area in the image is located at 291 degrees east longitude, 19 degrees south latitude. The image shows an area 30 kilometers (19.6 miles) wide and 76 km (47 miles) long. On the basis of Pioneer Venus and Arecibo data, it is known that two major rift zones occur in southern Phoebe Regio and that they terminate at about 20 to 25 degrees south latitude, about 2,000 km (1,240 miles) apart. This image is of an area just north of the southern end of the western rift zone. The region is characterized by a complex geologic history involving both volcanism and faulting. Several of the geologic units show distinctive overlapping or cross cutting relationships that permit identification and separation of geologic events and construction of the geologic history of the region. The oldest rocks in this image form the complexly deformed and faulted, radar bright, hilly terrain in the northern half. Faults of a variety of orientations are observed. A narrow fault trough (about one-half to one km (three tenths to six tenths of a mile) wide is seen crossing the bright hills near the lower part in the middle of the image. This is one of the youngest faults in the faulted, hilly unit as it is seen to cut across many other structures. The fault trough in turn appears to be embayed and flooded by the darker plains that appear in the south half of the image. These plains are interpreted to be of volcanic origin. The dark plains may be formed of a complex of overlapping volcanic flows. For example, the somewhat darker region of plains in the lower left (southwest) corner of the image may be a different age series of plains forming volcanic lava flows. Finally, the narrow bright line crossing the image in its lower part is interpreted to be a fault which cross cuts both plains units and is thus the youngest event in

  18. Laying bare Venus' dark secrets

    International Nuclear Information System (INIS)

    Allen, D.A.

    1987-01-01

    Ground-based IR observations of the dark side of Venus obtained in 1983 and 1985 with the Anglo-Australian Telescope are studied. An IR spectrum of Venus' dark side is analyzed. It is observed that the Venus atmosphere is composed of CO and radiation escapes only at 1.74 microns and 2.2 to 2.4 microns. The possible origin of the radiation, either due to absorbed sunlight or escaping thermal radiation, was investigated. These two hypotheses were eliminated, and it is proposed that the clouds of Venus are transparent and the radiation originates from the same stratum as the brighter portions but is weakened by the passage through the upper layer. The significance of the observed dark side markings is discussed

  19. Venus Suface Sampling and Analysis

    Data.gov (United States)

    National Aeronautics and Space Administration — This effort is developing the technology to transfer particulate samples from a Venus drill (being developed by Honeybee Robotics in a Phase 2 Small Business...

  20. Late Veneer consequences on Venus' long term evolution

    Science.gov (United States)

    Gillmann, C.; Golabek, G.; Tackley, P. J.; Raymond, S. N.

    2017-12-01

    Modelling of Venus' evolution is able to produce scenarios consistent with present-day observation. These results are however heavily dependent on atmosphere escape and initial volatile inventory. This primordial history (the first 500 Myr) is heavily influenced by collisions. We investigate how Late Veneer impacts change the initial state of Venus and their consequences on its coupled mantle/atmosphere evolution. We focus on volatile fluxes: atmospheric escape and mantle degassing. Mantle dynamics is simulated using the StagYY code. Atmosphere escape covers both thermal and non-thermal processes. Surface conditions are calculated with a radiative-convective model. Feedback of the atmosphere on the mantle through surface temperature is included. Large impacts are capable of contributing to atmospheric escape, volatile replenishment and energy transfer. We use the SOVA hydrocode to take into account volatile loss and deposition during a collision. Large impacts are not numerous enough to substantially erode Venus' atmosphere. Single impacts don't have enough eroding power. Swarms of small bodies (history of the planet and leads to lower present-day surface temperatures. Total depletion of the mantle seems unlikely, meaning either few large impacts (1 to 4) or low energy (slow, grazing…) collisions. Combined with the lack of plate tectonics and volatile recycling in the interior of Venus, Late Veneer collisions could help explain why Venus seems dry today.

  1. Venus: The First Habitable World of Our Solar System?

    Science.gov (United States)

    Way, Michael Joseph; Del Genio, Anthony; Kiang, Nancy; Sohl, Linda; Clune, Tom; Aleinov, Igor; Kelley, Maxwell

    2015-01-01

    A great deal of effort in the search for life off-Earth in the past 20+ years has focused on Mars via a plethora of space and ground based missions. While there is good evidence that surface liquid water existed on Mars in substantial quantities, it is not clear how long such water existed. Most studies point to this water existing billions of years ago. However,those familiar with the Faint Young Sun hypothesis for Earth will quickly realize that this problem is even more pronounced for Mars. In this context recent simulations have been completed with the GISS 3-D GCM (1) of paleo Venus (approx. 3 billion years ago) when the sun was approx. 25 less luminous than today. A combination of a less luminous Sun and a slow rotation rate reveal that Venus could have had conditions on its surface amenable to surface liquid water. Previous work has also provided bounds on how much water Venus could have had using measured DH ratios. It is possible that less assumptions have to be made to make Venus an early habitable world than have to be made for Mars, even thoughVenus is a much tougher world on which to confirm this hypothesis.

  2. Advancing Venus Geophysics with the NF4 VOX Gravity Investigation.

    Science.gov (United States)

    Iess, L.; Mazarico, E.; Andrews-Hanna, J. C.; De Marchi, F.; Di Achille, G.; Di Benedetto, M.; Smrekar, S. E.

    2017-12-01

    The Venus Origins Explorer is a JPL-led New Frontiers 4 mission proposal to Venus to answer critical questions about the origin and evolution of Venus. Venus stands out among other planets as Earth's twin planet, and is a natural target to better understand our own planet's place, in our own Solar System but also among the ever-increasing number of exoplanetary systems. The VOX radio science investigation will make use of an innovative Ka-band transponder provided by the Italian Space Agency (ASI) to map the global gravity field of Venus to much finer resolution and accuracy than the current knowledge, based on the NASA Magellan mission. We will present the results of comprehensive simulations performed with the NASA GSFC orbit determination and geodetic parameter estimation software `GEODYN', based on a realistic mission scenario, tracking schedule, and high-fidelity Doppler tracking noise model. We will show how the achieved resolution and accuracy help fulfill the geophysical goals of the VOX mission, in particular through the mapping of subsurface crustal density or thickness variations that will inform the composition and origin of the tesserae and help ascertain the heat loss and importance of tectonism and subduction.

  3. Rate of volcanism on Venus

    International Nuclear Information System (INIS)

    Fegley, B. Jr.; Prinn, R.G.

    1988-07-01

    The maintenance of the global H 2 SO 4 clouds on Venus requires volcanism to replenish the atmospheric SO 2 which is continually being removed from the atmosphere by reaction with calcium minerals on the surface of Venus. The first laboratory measurements of the rate of one such reaction, between SO 2 and calcite (CaCO 3 ) to form anhydrite (CaSO 4 ), are reported. If the rate of this reaction is representative of the SO 2 reaction rate at the Venus surface, then we estimate that all SO 2 in the Venus atmosphere (and thus the H 2 SO 4 clouds) will be removed in 1.9 million years unless the lost SO 2 is replenished by volcanism. The required rate of volcanism ranges from about 0.4 to about 11 cu km of magma erupted per year, depending on the assumed sulfur content of the erupted material. If this material has the same composition as the Venus surface at the Venera 13, 14 and Vega 2 landing sites, then the required rate of volcanism is about 1 cu km per year. This independent geochemically estimated rate can be used to determine if either (or neither) of the two discordant (2 cu km/year vs. 200 to 300 cu km/year) geophysically estimated rates is correct. The geochemically estimated rate also suggests that Venus is less volcanically active than the Earth

  4. Return to Venus of AKATSUKI, the Japanese Venus Orbiter

    Science.gov (United States)

    Nakamura, M.; Iwagami, N.; Satoh, T.; Taguchi, M.; Watanabe, S.; Takahashi, Y.; Imamura, T.; Suzuki, M.; Ueno, M.; Yamazaki, A.; Fukuhara, T.; Yamada, M.; Ishii, N.; Ogohara, K.

    2011-12-01

    Japanese Venus Climate Orbiter 'AKATSUKI' (PLANET-C) was proposed in 2001 with strong support by international Venus science community and approved as an ISAS mission soon after the proposal. AKATSUKI and ESA's Venus Express complement each other in Venus climate study. Various coordinated observations using the two spacecraft have been planned. Also participating scientists from US have been selected. Its science target is to understand the climate of Venus. The mission life we expected was more than 2 Earth years in Venus orbit. AKATSUKI was successfully launched at 06:58:22JST on May 21, by H-IIA F17. After the separation from H-IIA, the telemetry from AKATSUKI was normally detected by DSN Goldstone station (10:00JST) and the solar cell paddles' expansion was confirmed. AKATSUKI was put into the 3-axis stabilized mode in the initial operation from Uchinoura station and the critical operation was finished at 20:00JST on the same day. The malfunction, which happened during the Venus Orbit Insertion (VOI) on7 Dec, 2010 is as follows. We set all commands on Dec. 5. Attitude control for Venus orbit insertion (VOI) was automatically done on Dec. 6. Orbital maneuver engine (OME) was fired 08:49 JST on Dec. 7. 1min. after firing the spacecraft went into the occultation region and we had no telemetry, but we expected to continuous firing for 12min. Recording on the spacecraft told us later that, unfortunately the firing continued just 152sec. and stopped. The reason of the malfunction of the OME was the blocking of check valve of the gas pressure line to push the fuel to the engine. We failed to make the spacecraft the Venus orbiter, and it is rotating the sun with the orbital period of 203 days. As the Venus orbit the sun with the period of 225 days, AKATSUKI has a chance to meet Venus again in 5 or 6 years depending on the orbit correction plan. Let us summarize the present situation of AKATSUKI. Most of the fuel still remains. But the condition of the propulsion

  5. Minor actinide transmutation on PWR burnable poison rods

    International Nuclear Information System (INIS)

    Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

    2015-01-01

    Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

  6. Laboratory simulation of rod-to-rod mechanical interactions during postulated loss-of-coolant accidents in a PWR involving cladding oxidation

    International Nuclear Information System (INIS)

    Hindle, E.D.; Haste, T.J.; Harrison, W.R.

    1987-01-01

    Creep deformation of Zircaloy cladding in postulated PWR loss-of-coolant accidents may lead to rod-to-rod mechanical interactions. Tests have been performed in the electrically heated FOURSQUARE rig at 750 0 C and 850 0 C in steam to investigate this effect. Conservatisms inherent in a simple 'square with rounded corners' coolant channel blockage model have been quantified; about 5-10% flow area may remain even at strains which in ideal circumstances would give total blockage. Reduction of average burst strains produced by an oxide layer (up to 13 μm) has been demonstrated, resulting from strain concentration at oxide cracks. (author)

  7. Plutonium recycling in PWR

    International Nuclear Information System (INIS)

    Youinou, G.; Girieud, R.; Guigon, B.

    2000-01-01

    Two concepts of 100% MOX PWR cores are presented. They are designed such as to minimize the consequences of the introduction of Pu on the core control. The first one has a high moderation ratio and the second one utilizes an enriched uranium support. The important design parameters as well as their capabilities to multi recycle Pu are discussed. We conclude with the potential interest of the two concepts. (author)

  8. The integrated PWR

    International Nuclear Information System (INIS)

    Gautier, G.M.

    2002-01-01

    This document presents the integrated reactors concepts by a presentation of four reactors: PIUS, SIR, IRIS and CAREM. The core conception, the operating, the safety, the economical aspects and the possible users are detailed. From the performance of the classical integrated PWR, the necessity of new innovative fuels utilization, the research of a simplified design to make easier the safety and the KWh cost decrease, a new integrated reactor is presented: SCAR 600. (A.L.B.)

  9. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

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

  10. AGR v PWR

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

  11. Water chemistry in PWR

    International Nuclear Information System (INIS)

    Abe, Kenji

    1987-01-01

    This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

  12. PWR: 10 years after and perspectives

    International Nuclear Information System (INIS)

    1990-01-01

    These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

  13. Results from VENUS

    International Nuclear Information System (INIS)

    Ogawa, K.

    1990-01-01

    Recent results from VENUS experiments on e + e - reactions at energies between 52 and 60.8 GeV are presented. The R-values, the ratio of the total hadronic cross section to that of μ pair production, look slightly high within the present energy region. To understand this observation, a detailed study was carried out on the production of a heavy quark with |Q|=e/3. By using a next-to-leading log. approximation, the QCD cut-off parameter, Λ MS , was obtained as being Λ MS =208 MeV(+80MeV, -62MeV). The differential cross sections for e + e - → e + e - , γγ, μ + μ - , and τ + τ - were found to be consistent with predictions of the standard model. The average charge asymmetry for e + e - → qq-bar was also measured and found to be consistent with the prediction of the standard model. No evidence was observed indicating new particle production. No single photon production was observed and the upper limit of the number of light neutrino types was set to be N ν < 17.8 (90 % CL). (author)

  14. Venus project : experimentation at ENEA's pilot site

    International Nuclear Information System (INIS)

    Bargellini, M.L.; Fontana, F.; Niccolai, L.; Scavino, G.; Mancini, R.; Levialdi, S.

    1996-12-01

    The document describes the ENEA's (Italian Agency for New Technologies, Energy and the Environment) experience in the Venus Project (Esprit III 6398). Venus is an advanced visual interface based on icon representation that permits to end-user to inquiry databases. VENUS interfaces to ENEA's databases: cometa materials Module, Cometa Laboratories Module and European Programs. This report contents the results of the experimentation and of the validation carried out in ENEA's related to the Venus generations. Moreover, the description of the architecture, the user requirements syntesis and the validation methodology of the VENUS systems have been included

  15. Modelling activity transport behavior in PWR plant

    International Nuclear Information System (INIS)

    Henshaw, Jim; McGurk, John; Dickinson, Shirley; Burrows, Robert; Hinds, Kelvin; Hussey, Dennis; Deshon, Jeff; Barrios Figueras, Joan Pau; Maldonado Sanchez, Santiago; Fernandez Lillo, Enrique; Garbett, Keith

    2012-09-01

    The activation and transport of corrosion products around a PWR circuit is a major concern to PWR plant operators as these may give rise to high personnel doses. The understanding of what controls dose rates on ex-core surfaces and shutdown releases has improved over the years but still several questions remain unanswered. For example the relative importance of particle and soluble deposition in the core to activity levels in the plant is not clear. Wide plant to plant and cycle to cycle variations are noted with no apparent explanations why such variations are observed. Over the past few years this group have been developing models to simulate corrosion product transport around a PWR circuit. These models form the basis for the latest version of the BOA code and simulate the movement of Fe and Ni around the primary circuit. Part of this development is to include the activation and subsequent transport of radioactive species around the circuit and this paper describes some initial modelling work in this area. A simple model of activation, release and deposition is described and then applied to explain the plant behaviour at Sizewell B and Vandellos II. This model accounts for activation in the core, soluble and particulate activity movement around the circuit and for activity capture ex-core on both the inner and outer oxides. The model gives a reasonable comparison with plant observations and highlights what controls activity transport in these plants and importantly what factors can be ignored. (authors)

  16. Three dimensional calculations of the primary coolant flow in a 900 MW PWR vessel. Numerical simulation of the accurate RCP start-up flow rate

    International Nuclear Information System (INIS)

    Martin, A.; Alvarez, D.; Cases, F.; Stelletta, S.

    1997-06-01

    This report explains the last results about the mixing in the 900 MW PWR vessels. The accurate fluid flow transient, induced by the RCP starting-up, is represented. In a first time, we present the Thermalhydraulic Finite Element Code N3S used for the 3D numerical computations. After that, results obtained for one reactor operation case are given. This case is dealing with the transient mixing of a clear plug in the vessel when one primary pump starts-up. A comparison made between two injection modes; a steady state fluid flow conditions or the accurate RCP transient fluid flow conditions. The results giving the local minimum of concentration and the time response of the mean concentration at the core inlet are compared. The results show the real importance of the unsteadiness characteristics of the fluid flow transport of the clear water plug. (author)

  17. French PWR safety philosophy

    International Nuclear Information System (INIS)

    Conte, M.

    1986-05-01

    Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach, each of them having possibilities and limits. As a consequence of the global risk objective set in 1977 for nuclear reactors, safety analysis was extended to the evaluation of events more complex than the conventional ones, and later to the evaluation of the feasibility of the offsite emergency plans in case of severe accidents

  18. PWR decontamination feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Silliman, P.L.

    1978-12-18

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations.

  19. PWR decontamination feasibility study

    International Nuclear Information System (INIS)

    Silliman, P.L.

    1978-01-01

    The decontamination work which has been accomplished is reviewed and it is concluded that it is worthwhile to investigate further four methods for decontamination for future demonstration. These are: dilute chemical; single stage strong chemical; redox processes; and redox/chemical in combination. Laboratory work is recommended to define the agents and processes for demonstration and to determine the effect of the solvents on PWR materials. The feasibility of Indian Point 1 for decontamination demonstrations is discussed, and it is shown that the system components of Indian Point 1 are well suited for use in demonstrations

  20. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  1. Venus and Mercury as Planets

    Science.gov (United States)

    1974-01-01

    A general evolutionary history of the solar planetary system is given. The previously observed characteristics of Venus and Mercury (i.e. length of day, solar orbit, temperature) are discussed. The role of the Mariner 10 space probe in gathering scientific information on the two planets is briefly described.

  2. Surface and interior of Venus

    Energy Technology Data Exchange (ETDEWEB)

    Masursky, H [U.S. Geological Survey, Flagstaff, Arizona, USA; Kaula, W M [California Univ., Los Angeles (USA); McGill, G E [Massachusetts Univ., Amherst (USA); Pettengill, G H; Shapiro, I I [Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Earth and Planetary Sciences; Phillips, R J [Jet Propulsion Lab., Pasadena, Calif. (USA); Russell, C T [California Univ., Los Angeles (USA). Inst. of Geophysics and Planetary Physics; Schubert, G [California Univ., Los Angeles (USA)

    1977-06-01

    Present ideas about the surface and interior of Venus are based on data obtained from (1) Earth-based radio and radar: temperature, rotation, shape, and topography; (2) fly-by and orbiting spacecraft: gravity and magnetic fields; and (3) landers: winds, local structure, gamma radiation. Surface features, including large basins, crater-like depressions, and a linear valley, have been recognized from recent ground-based radar images. Pictures of the surface acquired by the USSR's Venera 9 and 10 show abundant boulders and apparent wind erosion. On the Pioneer Venus 1978 Orbiter mission, the radar mapper experiment will determine surface heights, dielectric constant values and small-scale slope values along the sub-orbital track between 50/sup 0/S and 75/sup 0/N. This experiment will also estimate the global shape and provide coarse radar images (40-80 km identification resolution) of part of the surface. Gravity data will be obtained by radio tracking. Maps combining radar altimetry with spacecraft and ground-based images will be made. A fluxgate magnetometer will measure the magnetic fields around Venus. The radar and gravity data will provide clues to the level of crustal differentiation and tectonic activity. The magnetometer will determine the field variations accurately. Data from the combined experiments may constrain the dynamo mechanism; if so, a deeper understanding of both Venus and Earth will be gained.

  3. Venus and Mercury as planets

    International Nuclear Information System (INIS)

    1974-01-01

    A general evolutionary history of the solar planetary system is given. The previously observed characteristics of Venus and Mercury (i.e. length of day, solar orbit, temperature) are discussed. The role of the Mariner 10 space probe in gathering scientific information on the two planets is briefly described

  4. 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

  5. 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

  6. Implications of /sup 36/A excess on Venus

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, M [Tokyo Univ. (Japan). Inst. of Space and Aeronautical Science

    1979-05-01

    The finding of /sup 36/A excess on Venus by the mass-spectroscopic measurement of the Venus Pioneer appears to endorse the more rapid accretion theory of Venus than the Earth and the secondary origin of the terrestrial atmosphere.

  7. Venus Lightning: What We Have Learned from the Venus Express Fluxgate Magnetometer

    Science.gov (United States)

    Russell, C. T.; Strangeway, R. J.; Wei, H. Y.; Zhang, T. L.

    2010-03-01

    The Venus Express magnetometer sees short (tens of milliseconds) pulses of EM waves in the Venus ionosphere as predicted by the lightning model for the PVO electric pulses. These waves are stronger than similar terrestrial signals produced by lightning.

  8. Modeling on a PWR power conversion system with system program

    International Nuclear Information System (INIS)

    Gao Rui; Yang Yanhua; Lin Meng

    2007-01-01

    Based on the power conversion system of nuclear and conventional islands of Daya Bay Power Station, this paper models the thermal-hydraulic systems of primary and secondary loops for PWR by using the PWR best-estimate program-RELAP5. To simulate the full-scope power conversion system, not only the traditional basic system models of nuclear island, but also the major system models of conventional island are all considered and modeled. A comparison between the calculated results and the actual data of reactor demonstrates a fine match for Daya Bay Nuclear Power Station, and manifests the feasibility in simulating full-scope power conversion system of PWR by RELAP5 at the same time. (authors)

  9. Safety considerations of PWR's

    International Nuclear Information System (INIS)

    Arnold, W.H. Jr.

    1977-01-01

    The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

  10. PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

    1982-10-01

    Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

  11. A Study on Structured Simulation Framework for Design and Evaluation of Human-Machine Interface System -Application for On-line Risk Monitoring for PWR Nuclear Power Plant-

    International Nuclear Information System (INIS)

    Zhan, J.; Yang, M.; Li, S.C.; Peng, M.J.; Yan, S.Y.; Zhang, Z.J.

    2006-01-01

    The operators in the main control room of Nuclear Power Plant (NPP) need to monitor plant condition through operation panels and understand the system problems by their experiences and skills. It is a very hard work because even a single fault will cause a large number of plant parameters abnormal and operators are required to perform trouble-shooting actions in a short time interval. It will bring potential risks if operators misunderstand the system problems or make a commission error to manipulate an irrelevant switch with their current operation. This study aims at developing an on-line risk monitoring technique based on Multilevel Flow Models (MFM) for monitoring and predicting potential risks in current plant condition by calculating plant reliability. The proposed technique can be also used for navigating operators by estimating the influence of their operations on plant condition before they take an action that will be necessary in plant operation, and therefore, can reduce human errors. This paper describes the risk monitoring technique and illustrates its application by a Steam Generator Tube Rupture (SGTR) accident in a 2-loop Pressurized Water Reactor (PWR) Marine Nuclear Power Plant (MNPP). (authors)

  12. New perspectives on the accretion and internal evolution of Venus

    Science.gov (United States)

    O'Rourke, J. G.

    2017-12-01

    Dichotomous conditions on Earth and Venus present one of the most compelling mysteries in our Solar System. Ongoing debate centers on how the internal dynamics of Venus have shaped its atmospheric composition, surface features, and even habitability over geologic time. In particular, Venus may have resembled Earth for billions of years before suffering catastrophic transformation, or perhaps some accretionary process set these twin planets on divergent paths from the beginning. Unfortunately, the limited quality of decades-old data—particularly the low resolution of radar imagery and global topography from NASA's Magellan mission—harms our ability to draw definite conclusions. But some progress is possible given recent advances in modeling techniques and improved topography derived from stereo images that are available for roughly twenty percent of the surface. Here I present simulations of the interior evolution of Venus consistent with all available constraints and, more importantly, identify future measurements that would dramatically narrow the range of acceptable scenarios. Obtaining high-resolution imagery and topography, along with any information about the temporal history of a magnetic field, is extremely important. Deformation of geologic features constrains the surface heat flow and lithospheric rheology during their formation. Determining whether craters with radar-dark floors (which comprise 80% of the population) are actually embayed by lava flows would finally settle the controversy over catastrophic versus equilibrium resurfacing. If the core of Venus has completely solidified, then the lack of an internally generated magnetic field today is unsurprising. We might expect dynamo action in the past since relatively high mantle temperatures may increase the rate of core cooling—unless a lack of giant impacts during accretion permitted chemical stratification that resists convection. In any case, uncertainty about our celestial cousin reveals a

  13. PWR burnable absorber evaluation

    International Nuclear Information System (INIS)

    Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

    1995-01-01

    The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

  14. 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

  15. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

    A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

  16. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  17. Nuclear instrumentation in VENUS-F

    Science.gov (United States)

    Wagemans, J.; Borms, L.; Kochetkov, A.; Krása, A.; Van Grieken, C.; Vittiglio, G.

    2018-01-01

    VENUS-F is a fast zero power reactor with 30 wt% U fuel and Pb/Bi as a coolant simulator. Depending on the experimental configuration, various neutron spectra (fast, epithermal, and thermal islands) are present. This paper gives a review of the nuclear instrumentation that is applied for reactor control and in a large variety of physics experiments. Activation foils and fission chambers are used to measure spatial neutron flux profiles, spectrum indices, reactivity effects (with positive period and compensation method or the MSM method) and kinetic parameters (with the Rossi-alpha method). Fission chamber calibrations are performed in the standard irradiation fields of the BR1 reactor (prompt fission neutron spectrum and Maxwellian thermal neutron spectrum).

  18. Venus tectonics: another Earth or another Mars

    International Nuclear Information System (INIS)

    McGill, G.E.

    1979-01-01

    The presence of presumably primordial large craters has led to the suggestion that Venus may have a thick lithosphere like that of Mars despite its similarities to Earth in size and density. However, crust and upper mantle temperatures on Venus are very likely higher than on Earth so that a dry Venus could have a lithosphere with a thickness similar to that of Earth. If a trace of volatiles is present in the mantle, the lithosphere of Venus could be thinner. Due to the absence of liquid water, erosion and deposition will be much slower on Venus than on Earth, favoring retention of primordial cratered surfaces on portions of the crust that have not been destroyed or buried by tectonic and volcanic activity. Geochemical models of solar system origin and petrological considerations suggest that K is about as abundant in Venus as in Earth. The abundance of 40 Ar in the atmosphere of Venus lies somewhere between the Earth value and one-tenth of the Earth value. Because erosional liberation of 40 Ar on Venus will be relatively inefficient, this range for 40 Ar abundance at least permits an active tectonic history, and if the 40 Ar abundance is towards the high end of the range, it may well require an active tectonic history. Thus we are not constrained to a Mars-like model of Venus tectonics by craters and possible mantle dryness; an Earth-like model is equally probable

  19. Venus transits - A French view

    Science.gov (United States)

    Débarbat, Suzanne

    2005-04-01

    After a careful study of Mars observations obtained by Tycho Brahé (1546-1601), Kepler (1571-1630) discovered the now-called Kepler's third law. In 1627 he published his famous Tabulae Rudolphinae, a homage to his protector Rudolph II (1552-1612), tables (Kepler 1609, 1627) from which he predicted Mercury and Venus transits over the Sun. In 1629 Kepler published his Admonitio ad Astronomos Advertisement to Astronomers (Kepler 1630), Avertissement aux Astronomes in French Au sujet de phénomènes rares et étonnants de l'an 1631: l'incursion de Vénus et de Mercure sur le Soleil. This was the beginning of the interest of French astronomers, among many others, in such transits, mostly for Venus, the subject of this paper in which dates are given in the Gregorian calendar.

  20. Signs of Life on Venus

    Science.gov (United States)

    Ksanfomality, L.

    2012-04-01

    The search for "habitable zones" in extrasolar planetary systems is based on the premise of "normal" physical conditions in a habitable zone, i.e. pressure, temperature range, and atmospheric composition similar to those on the Earth. However, one should not exclude completely the possibility of the existence of life at relatively high temperatures, despite the fact that at the first glance it seems impossible. The planet Venus with its dense, hot (735 K), oxigenless CO2 - atmosphere and high 92 bar-pressure at the surface could be the natural laboratory for the studies of this type. Amid exoplanets, celestial bodies with the physical conditions similar to the Venusian can be met. The only existing data of actual close-in observations of Venus' surface are the results of a series of missions of the soviet VENERA landers which took place the 1970's and 80's in the atmosphere and on the surface of Venus. For 36 and 29 years since these missions, respectively, I repeatedly returned to the obtained images of the Venus' surface in order to reveal on them any unusual objects observed in the real conditions of Venus. The new analysis of the Venus' panoramas was based on the search of unusual elements in two ways. Since the efficiency of the VENERA landers maintained for a long time they produced a large number of primary television panoramas during the lander's work. Thus, one can try to detect: (a) any differences in successive images (appearance or disappearance of parts of the image or change of their shape), and understand what these changes are related to (e.g., wind), and whether they are related to hypothetical habitability of a planet. Another sign (b) of the wanted object is their morphological peculiarities which distinguishes them from the ordinary surface details. The results of VENERA-9 (1975) and VENERA -13 (1982) are of the main interest. A few relatively large objects ranging from a decimeter to half meter and with unusual morphology were observed in some

  1. Transient study of a PWR pressurizer

    International Nuclear Information System (INIS)

    Sotoma, H.

    1973-01-01

    An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

  2. Two optimal control methods for PWR core control

    International Nuclear Information System (INIS)

    Karppinen, J.; Blomsnes, B.; Versluis, R.M.

    1976-01-01

    The Multistage Mathematical Programming (MMP) and State Variable Feedback (SVF) methods for PWR core control are presented in this paper. The MMP method is primarily intended for optimization of the core behaviour with respect to xenon induced power distribution effects in load cycle operation. The SVF method is most suited for xenon oscillation damping in situations where the core load is unpredictable or expected to stay constant. Results from simulation studies in which the two methods have been applied for control of simple PWR core models are presented. (orig./RW) [de

  3. French PWR Safety Philosophy

    International Nuclear Information System (INIS)

    Conte, M. M.

    1986-01-01

    The first 900 MWe units, built under the American Westinghouse licence and with reference to the U. S. regulation, were followed by 28 standardized units, C P1 and C P2 series. Increasing knowledge and lessons learned from starting and operating experience of French nuclear power plants, completed by the experience learned from the operation of foreign reactors, has contributed to the improvement of French PWR design and safety philosophy. As early as 1976, this experience was taken into account by French Safety organisms to discuss, with Electricite de France, the safety options for the planned 1300 MWe units, P4 and P4 series. In 1983, the new reactor scheduled, Ni4 series 1400 MWe, is a totally French design which satisfies the French regulations and other French standards and codes. Based on a deterministic approach, the French safety analysis was progressively completed by a probabilistic approach each of them having possibilities and limits. Increasing knowledge and lessons learned from operating experience have contributed to the French safety philosophy improvement. The methodology now applied to safety evaluation develops a new facet of the in depth defense concept by taking highly unlikely events into consideration, by developing the search of safety consistency of the design, and by completing the deterministic approach by the probabilistic one

  4. Stochastic optimization of loading pattern for PWR

    International Nuclear Information System (INIS)

    Smuc, T.; Pevec, D.

    1994-01-01

    The application of stochastic optimization methods in solving in-core fuel management problems is restrained by the need for a large number of proposed solutions loading patterns, if a high quality final solution is wanted. Proposed loading patterns have to be evaluated by core neutronics simulator, which can impose unrealistic computer time requirements. A new loading pattern optimization code Monte Carlo Loading Pattern Search has been developed by coupling the simulated annealing optimization algorithm with a fast one-and-a-half dimensional core depletion simulator. The structure of the optimization method provides more efficient performance and allows the user to empty precious experience in the search process, thus reducing the search space size. Hereinafter, we discuss the characteristics of the method and illustrate them on the results obtained by solving the PWR reload problem. (authors). 7 refs., 1 tab., 1 fig

  5. The simulation status of particle transport system JPTS

    International Nuclear Information System (INIS)

    Deng, L.

    2015-01-01

    'Full text:' Particle transport system JPTS has been developed by IAPCM. It is based on the three support frustrations (JASMIN, JAUMIN and JCOGIN) and is used to simulate the reactor full core and radiation shielding problems. The system has been realized the high fidelity. In this presentation, analysis of the H-M, BEAVRS, VENUS-III and SG-III models are shown. Analyze HZP conditions of BEAVRS model with Monte Carlo code JMCT, MC21 and OpenMC to assess code accuracy against available data. Assess the feasibility of analysis of a PWR using JMCT. The large scale depletion solver is also shown. Assess the feasibility of analysis of radiation shielding using JSNT. JPTS has been proved with the capability of the full-core pin-by-pin and radiation shielding. (author)

  6. 3D-FE Modeling of 316 SS under Strain-Controlled Fatigue Loading and CFD Simulation of PWR Surge Line

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Barua, Bipul [Argonne National Lab. (ANL), Argonne, IL (United States); Listwan, Joseph [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    In financial year 2017, we are focusing on developing a mechanistic fatigue model of surge line pipes for pressurized water reactors (PWRs). To that end, we plan to perform the following tasks: (1) conduct stress- and strain-controlled fatigue testing of surge-line base metal such as 316 stainless steel (SS) under constant, variable, and random fatigue loading, (2) develop cyclic plasticity material models of 316 SS, (3) develop one-dimensional (1D) analytical or closed-form model to validate the material models and to understand the mechanics associated with 316 SS cyclic hardening and/or softening, (4) develop three-dimensional (3D) finite element (FE) models with implementation of evolutionary cyclic plasticity, and (5) develop computational fluid dynamics (CFD) model for thermal stratification, thermal-mechanical stress, and fatigue of example reactor components, such as a PWR surge line under plant heat-up, cool-down, and normal operation with/without grid-load-following. This semi-annual progress report presents the work completed on the above tasks for a 316 SS laboratory-scale specimen subjected to strain-controlled cyclic loading with constant, variable, and random amplitude. This is the first time that the accurate 3D-FE modeling of the specimen for its entire fatigue life, including the hardening and softening behavior, has been achieved. We anticipate that this work will pave the way for the development of a fully mechanistic-computer model that can be used for fatigue evaluation of safety-critical metallic components, which are traditionally evaluated by heavy reliance on time-consuming and costly test-based approaches. This basic research will not only help the nuclear reactor industry for fatigue evaluation of reactor components in a cost effective and less time-consuming way, but will also help other safety-related industries, such as aerospace, which is heavily dependent on test-based approaches, where a single full-scale fatigue test can cost

  7. Valve testing for UK PWR safety applications

    International Nuclear Information System (INIS)

    George, P.T.; Bryant, S.

    1989-01-01

    Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

  8. Mars and Venus: unequal planets.

    Science.gov (United States)

    Zimmerman, T S; Haddock, S A; McGeorge, C R

    2001-01-01

    Self-help books, a pervasive and influential aspect of society, can have a beneficial or detrimental effect on the therapeutic process. This article describes a thematic analysis and feminist critique of the best-selling self-help book, Men are from Mars, Women are from Venus. This analysis revealed that the author's materials are inconsistent with significant family therapy research findings and key principles of feminist theories. His descriptions of each gender and his recommendations for improving relationships serve to endorse and encourage power differentials between women and men.

  9. Distant interplanetary wake of Venus: plasma observations from pioneer Venus

    International Nuclear Information System (INIS)

    Mihalov, J.D.; Barnes, A.

    1982-01-01

    In June 1979 the Pioneer Venus orbiter made its first series of passes through the distant solar wind wake of Venus at distances of 8--12 R/sub V/ behind the planet. During this period the plasma analyzer aboard the spacecraft detected disturbed magnetosheath plasma that intermittently disappeared and reappeared, suggesting a tattered, filamentary cavity trailing behind the planet. The magnetosheath dropouts almost always occurred inside the region of 'magnetotail' observed by Russell et al. Sporadic bursts of energetic ions (E/q> or approx. =4kV) are detected inside and, occasionally, outside the magnetotail; all such bursts are consistent with identification of the ion as O + of planetary origin moving at the local magnetosheath flow speed. The morphology of the plasma dropouts and of the O + bursts is analyzed in detail. The cavity appears to contract at times of high solar wind dynamic pressure. The intensity of the O + component is highly variable, and appears not to be strongly correlated with solar wind dynamic pressure. The most intense bursts correspond to a flux 7 ions cm - 2 s - 1 . This maximum flux, if steady and filling a cylinder 1 R/sub V/ in radius would correspond to a mass loss rate of 25 ions s - 1 ; the intermittency and variability of the flux suggest that the true mean loss rate is very much lower. The kinetic temperature of the O + component is estimated as 10 5 --10 6 K in order of magnitude

  10. Sizewell 'B' PWR reference design

    International Nuclear Information System (INIS)

    1982-04-01

    The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

  11. Methane measurement by the Pioneer Venus large probe neutral mass spectrometer

    Science.gov (United States)

    Donahue, T. M.; Hodges, R. R., Jr.

    1992-12-01

    The Pioneer Venus Large Probe Mass Spectrometer detected a large quantity of methane as it descended below 20 km in the atmosphere of Venus. Terrestrial methane and Xe-136, both originating in the same container and flowing through the same plumbing, were deliberately released inside the mass spectrometer for instrumental reasons. However, the Xe-136 did not exhibit behavior similar to methane during Venus entry, nor did CH4 in laboratory simulations. The CH4 was deuterium poor compared to Venus water and hydrogen. While the inlet to the mass spectrometer was clogged with sulfuric acid droplets, significant deuteration of CH4 and its H2 progeny was observed. Since the only source of deuterium identifiable was water from sulfuric acid, we have concluded that we should correct the HDO/H2O ratio in Venus water from 3.2 x 10-2 to (5 plus or minus 0.7) x 10-2. When the probe was in the lower atmosphere, transfer of deuterium from Venus HDO and HD to CH4 can account quantitatively for the deficiencies recorded in HDO and HD below 10 km, and consequently, the mysterious gradients in water vapor and hydrogen mixing ratios we have reported. The revision in the D/H ratio reduces the mixing ratio of water vapor (and H2) reported previously by a factor of 3.2/5. We are not yet able to say whether the methane detected was atmospheric or an instrumental artifact. If it was atmospheric, its release must have been episodic and highly localized. Otherwise, the large D/H ratio in Venus water and hydrogen could not be maintained.

  12. Data assimilation and PWR primary measurement

    International Nuclear Information System (INIS)

    Mercier, Thibaud

    2015-01-01

    A Pressurized Water Reactor (PWR) Reactor Coolant System (RCS) is a highly complex physical process: heterogeneous power, flow and temperature distributions are difficult to be accurately measured, since instrumentations are limited in number, thus leading to the relevant safety and protection margins. EDF R and D is seeking to assess the potential benefits of applying Data Assimilation to a PWR's RCS (Reactor Coolant System) measurements, in order to improve the estimators for parameters of a reactor's operating setpoint, i.e. improving accuracy and reducing uncertainties and biases of measured RCS parameters. In this thesis, we define a 0D semi-empirical model for RCS, satisfying the description level usually chosen by plant operators, and construct a Monte-Carlo Method (inspired from Ensemble Methods) in order to use this model with Data Assimilation tools. We apply this method on simulated data in order to assess the reduction of uncertainties on key parameters: results are beyond expectations, however strong hypothesis are required, implying a careful preprocessing of input data. (author)

  13. Modeling of PWR fuel at extended burnup

    International Nuclear Information System (INIS)

    Dias, Raphael Mejias

    2016-01-01

    This work studies the modifications implemented over successive versions in the empirical models of the computer program FRAPCON used to simulate the steady state irradiation performance of Pressurized Water Reactor (PWR) fuel rods under high burnup condition. In the study, the empirical models present in FRAPCON official documentation were analyzed. A literature study was conducted on the effects of high burnup in nuclear fuels and to improve the understanding of the models used by FRAPCON program in these conditions. A steady state fuel performance analysis was conducted for a typical PWR fuel rod using FRAPCON program versions 3.3, 3.4, and 3.5. The results presented by the different versions of the program were compared in order to verify the impact of model changes in the output parameters of the program. It was observed that the changes brought significant differences in the results of the fuel rod thermal and mechanical parameters, especially when they evolved from FRAPCON-3.3 version to FRAPCON-3.5 version. Lower temperatures, lower cladding stress and strain, lower cladding oxide layer thickness were obtained in the fuel rod analyzed with the FRAPCON-3.5 version. (author)

  14. The Atmosphere and Climate of Venus

    Science.gov (United States)

    Bullock, M. A.; Grinspoon, D. H.

    Venus lies just sunward of the inner edge of the Sun's habitable zone. Liquid water is not stable. Like Earth and Mars, Venus probably accreted at least an ocean's worth of water, although there are alternative scenarios. The loss of this water led to the massive, dry CO2 atmosphere, extensive H2SO4 clouds (at least some of the time), and an intense CO2 greenhouse effect. This chapter describes the current understanding of Venus' atmosphere, established from the data of dozens of spacecraft and atmospheric probe missions since 1962, and by telescopic observations since the nineteenth century. Theoretical work to model the temperature, chemistry, and circulation of Venus' atmosphere is largely based on analogous models developed in the Earth sciences. We discuss the data and modeling used to understand the temperature structure of the atmosphere, as well as its composition, cloud structure, and general circulation. We address what is known and theorized about the origin and early evolution of Venus' atmosphere. It is widely understood that Venus' dense CO2 atmosphere is the ultimate result of the loss of an ocean to space, but the timing of major transitions in Venus' climate is very poorly constrained by the available data. At present, the bright clouds allow only 20% of the sunlight to drive the energy balance and therefore determine conditions at Venus' surface. Like Earth and Mars, differential heating between the equator and poles drives the atmospheric circulation. Condensable species in the atmosphere create clouds and hazes that drive feedbacks that alter radiative forcing. Also in common with Earth and Mars, the loss of light, volatile elements to space produces long-term changes in composition and chemistry. As on Earth, geologic processes are most likely modifying the atmosphere and clouds by injecting gases from volcanos as well as directly through chemical reactions with the surface. The sensitivity of Venus' atmospheric energy balance is quantified in

  15. PWR AXIAL BURNUP PROFILE ANALYSIS

    International Nuclear Information System (INIS)

    J.M. Acaglione

    2003-01-01

    The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

  16. Global Geological Map of Venus

    Science.gov (United States)

    Ivanov, M. A.

    2008-09-01

    Introduction: The Magellan SAR images provide sufficient data to compile a geological map of nearly the entire surface of Venus. Such a global and selfconsistent map serves as the base to address the key questions of the geologic history of Venus. 1) What is the spectrum of units and structures that makes up the surface of Venus [1-3]? 2) What volcanic/tectonic processes do they characterize [4-7]? 3) Did these processes operated locally, regionally, or globally [8- 11]? 4) What are the relationships of relative time among the units [8]? 5) At which length-scale these relationships appear to be consistent [8-10]? 6) What is the absolute timing of formation of the units [12-14]? 7) What are the histories of volcanism, tectonics and the long-wavelength topography on Venus? 7) What model(s) of heat loss and lithospheric evolution [15-21] do these histories correspond to? The ongoing USGS program of Venus mapping has already resulted in a series of published maps at the scale 1:5M [e.g. 22-30]. These maps have a patch-like distribution, however, and are compiled by authors with different mapping philosophy. This situation not always results in perfect agreement between the neighboring areas and, thus, does not permit testing geological hypotheses that could be addressed with a self-consistent map. Here the results of global geological mapping of Venus at the scale 1:10M is presented. The map represents a contiguous area extending from 82.5oN to 82.5oS and comprises ~99% of the planet. Mapping procedure: The map was compiled on C2- MIDR sheets, the resolution of which permits identifying the basic characteristics of previously defined units. The higher resolution images were used during the mapping to clarify geologic relationships. When the map was completed, its quality was checked using published USGS maps [e.g., 22-30] and the catalogue of impact craters [31]. The results suggest that the mapping on the C2-base provided a highquality map product. Units and

  17. High Temperature, Wireless Seismometer Sensor for Venus

    Science.gov (United States)

    Ponchak, George E.; Scardelletti, Maximilian C.; Taylor, Brandt; Beard, Steve; Meredith, Roger D.; Beheim, Glenn M.; Hunter Gary W.; Kiefer, Walter S.

    2012-01-01

    Space agency mission plans state the need to measure the seismic activity on Venus. Because of the high temperature on Venus (462? C average surface temperature) and the difficulty in placing and wiring multiple sensors using robots, a high temperature, wireless sensor using a wide bandgap semiconductor is an attractive option. This paper presents the description and proof of concept measurements of a high temperature, wireless seismometer sensor for Venus. A variation in inductance of a coil caused by the movement of an aluminum probe held in the coil and attached to a balanced leaf-spring seismometer causes a variation of 700 Hz in the transmitted signal from the oscillator/sensor system at 426? C. This result indicates that the concept may be used on Venus.

  18. Resfria - a computational routine for thermal-hydraulic analysis of a cooldown in the PWR

    International Nuclear Information System (INIS)

    Silva Neto, A.J. da; Maciel Filho, L.A.

    1989-01-01

    This paper presents the computer code RESFRIA, designed to calculate the process parameters in a PWR nuclear power plant during a cooldown normal procedure. The procedure is described and some of the models developed to the simulation of systems and equipments are presented. A simplified flowchart of the computational routine and the results in the form of a diagram, for a typical PWR nuclear power plant, are also presented. (author)

  19. Bias identification in PWR pressurizer instrumentation using the generalized liklihood-ratio technique

    International Nuclear Information System (INIS)

    Tylee, J.L.

    1981-01-01

    A method for detecting and identifying biases in the pressure and level sensors of a pressurized water reactor (PWR) pressurizer is described. The generalized likelihood ratio (GLR) technique performs statistical tests on the innovations sequence of a Kalman filter state estimator and is capable of determining when a bias appears, in what sensor the bias exists, and estimating the bias magnitude. Simulation results using a second-order linear, discrete PWR pressurizer model demonstrate the capabilities of the GLR method

  20. Influence of boron reduction strategies on PWR accident management flexibility

    International Nuclear Information System (INIS)

    Papukchiev, Angel Aleksandrov; Liu, Yubo; Schaefer, Anselm

    2007-01-01

    In conventional pressurized water reactor (PWR) designs, soluble boron is used for reactivity control over core fuel cycle. Design changes to reduce boron concentration in the reactor coolant are of general interest regarding three aspects - improved reactivity feedback properties, lower impact of boron dilution scenarios on PWR safety and eventually more flexible accident management procedures. In order to assess the potential advantages through the introduction of boron reduction strategies in current PWRs, two low boron core configurations based on fuel with increased utilization of gadolinium and erbium burnable absorbers have been developed. The new PWR designs permit to reduce the natural boron concentration in reactor coolant at begin of cycle to 518 ppm and 805 ppm. For the assessment of the potential safety advantages of these cores a hypothetical beyond design basis accident has been simulated with the system code ATHLET. The analyses showed improved inherent safety and increased accident management flexibility of the low boron cores in comparison with the standard PWR. (author)

  1. Venus and the Archean Earth: Thermal considerations

    International Nuclear Information System (INIS)

    Sleep, N.H.

    1989-01-01

    The Archean Era of the Earth is not a direct analog of the present tectonics of Venus. In this regard, it is useful to review the state of the Archean Earth. Most significantly, the temperature of the adiabatic interior of the Earth was 200 to 300 C hotter than the current temperature. Preservation biases limit what can be learned from the Archean record. Archean oceanic crust, most of the planetary surface at any one time, has been nearly all subducted. More speculatively, the core of the Earth has probably cooled more slowly than the mantle. Thus the temperature contrast above the core-mantle boundary and the vigor of mantle plumes has increased with time on the Earth. The most obvious difference between Venus and the present Earth is the high surface temperature and hence a low effective viscosity of the lithosphere. In addition, the temperature contrast between the adiabatic interior and the surface, which drives convection, is less on Venus than on the Earth. It appears that the hot lithosphere enhanced tectonics on the early Venus significantly enough that its interior cooled faster than the Earth's. The best evidence for a cool interior of Venus comes from long wavelength gravity anomalies. The low interior temperatures retard seafloor spreading on Venus. The high surface temperatures on Venus enhance crustal deformation. That is, the lower crust may become ductile enough to permit significant flow between the upper crust and the mantle. There is thus some analogy to modern and ancient areas of high heat flow on the Earth. Archean crustal blocks typically remained stable for long intervals and thus overall are not good analogies to the deformation style on Venus

  2. Geology of Maxwell Montes, Venus

    Science.gov (United States)

    Head, J. W.; Campbell, D. B.; Peterfreund, A. R.; Zisk, S. A.

    1984-01-01

    Maxwell Montes represent the most distinctive topography on the surface of Venus, rising some 11 km above mean planetary radius. The multiple data sets of the Pioneer missing and Earth based radar observations to characterize Maxwell Montes are analyzed. Maxwell Montes is a porkchop shaped feature located at the eastern end of Lakshmi Planum. The main massif trends about North 20 deg West for approximately 1000 km and the narrow handle extends several hundred km West South-West WSW from the north end of the main massif, descending down toward Lakshmi Planum. The main massif is rectilinear and approximately 500 km wide. The southern and northern edges of Maxwell Montes coincide with major topographic boundaries defining the edge of Ishtar Terra.

  3. A dynamic model of Venus's gravity field

    Science.gov (United States)

    Kiefer, W. S.; Richards, M. A.; Hager, B. H.; Bills, B. G.

    1984-01-01

    Unlike Earth, long wavelength gravity anomalies and topography correlate well on Venus. Venus's admittance curve from spherical harmonic degree 2 to 18 is inconsistent with either Airy or Pratt isostasy, but is consistent with dynamic support from mantle convection. A model using whole mantle flow and a high viscosity near surface layer overlying a constant viscosity mantle reproduces this admittance curve. On Earth, the effective viscosity deduced from geoid modeling increases by a factor of 300 from the asthenosphere to the lower mantle. These viscosity estimates may be biased by the neglect of lateral variations in mantle viscosity associated with hot plumes and cold subducted slabs. The different effective viscosity profiles for Earth and Venus may reflect their convective styles, with tectonism and mantle heat transport dominated by hot plumes on Venus and by subducted slabs on Earth. Convection at degree 2 appears much stronger on Earth than on Venus. A degree 2 convective structure may be unstable on Venus, but may have been stabilized on Earth by the insulating effects of the Pangean supercontinental assemblage.

  4. The applicability of CFD to simulate and study the mixing process and the thermo-hydraulic consequences of a main steam line break in PWR model

    Directory of Open Access Journals (Sweden)

    Farkas Istvan

    2017-01-01

    Full Text Available This paper focuses on the validation and applicability of CFD to simulate and analyze the thermo-hydraulic consequences of a main steam line break. Extensive validation data come from experiments performed using the Rossendorf coolant mixing model facility. For the calculation, the range of 9 to 12 million hexahe¬dral cells was constructed to capture all details in the interrogation domain in the system. The analysis was performed by running a time-dependent calculation, Detailed analyses were made at different cross-sections in the system to evaluate not only the value of the maximum and minimum temperature, but also the loca¬tion and the time at which it occurs during the transient which is considered to be indicator for the quality of mixing in the system. CFD and experimental results were qualitatively compared; mixing in the cold legs with emergency core cooling systems was overestimated. This could be explained by the sensitivity to the bound¬ary conditions. In the downcomer, the experiments displayed higher mixing: by our assumption this related to the dense measurement grid (they were not modelled. The temperature distribution in the core inlet plane agreed with the measurement results. Minor deviations were seen in the quantitative comparisons: the maximum temperature difference was 2ºC.

  5. PWR-to-PWR fuel cycle model using dry process

    International Nuclear Information System (INIS)

    Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

    2002-03-01

    PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

  6. Simplified model of a PWR primary circuit

    International Nuclear Information System (INIS)

    Souza, A.L.; Faya, A.J.G.

    1988-07-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analyzed by a nodal model. Average and hot channels are treated so that bulk response of the core and DNBR can be evaluated. A homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  7. Calculation of drop course of control rod assembly in PWR

    International Nuclear Information System (INIS)

    Zhou Xiaojia; Mao Fei; Min Peng; Lin Shaoxuan

    2013-01-01

    The validation of control rod drop performance is an important part of safety analysis of nuclear power plant. Development of computer code for calculating control rod drop course will be useful for validating and improving the design of control rod drive line. Based on structural features of the drive line, the driving force on moving assembly was analyzed and decomposed, the transient value of each component of the driving force was calculated by choosing either theoretical method or numerical method, and the simulation code for calculating rod cluster control assembly (RCCA) drop course by time step increase was achieved. The analysis results of control rod assembly drop course calculated by theoretical model and numerical method were validated by comparing with RCCA drop test data of Qinshan Phase Ⅱ 600 MW PWR. It is shown that the developed RCCA drop course calculation code is suitable for RCCA in PWR and can correctly simulate the drop course and the stress of RCCA. (authors)

  8. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Post test investigations of bundle test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.

    1986-11-01

    This KfK report describes the post test investigation of bundle experiment ESBU-2a. ESBU-2a was the second of two bundle tests on the temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS-Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (central tungsten heater, UO 2 -ring pellet and zircaloy cladding). The length was 0.4 meter. The bundle was heated to a maximum temperature of 2175 0 C. Molten cladding which dissolved part of the UO 2 pellets and slumped away from the already oxidized cladding formed a lump in the lower part of the bundle. After the test the bundle was embedded in epoxy and sectioned with a diamand saw, in the region of the refrozen melt. The cross sections were investigated by metallographic examination. The refrozen (U,Zr,O) melt consists variously of three phases with increasing oxygen content (metallic α-Zry, metallic (U,Zr) alloy and a (U,Zr)O 2 mixed oxide), two phases (α-Zry, (U,Zr)O 2 mixed oxide), or one phase ((U,Zr)O 2 mixed oxide). The cross sections show the increasing oxidation of the cladding with increasing elevation (temperature). A strong azimuthal dependency of the oxidation is found. In regions where the initial oxidized cladding is contacted by the melt one can recognize the interaction between the metallic melt and ZrO 2 of the cladding. Oxygen is taken away from the ZrO 2 . If the melt is in direct contact with steam a relatively well defined oxide layer is formed. (orig.) [de

  9. PWR plant construction in Japan

    International Nuclear Information System (INIS)

    Tamura, Toshifumi

    2002-01-01

    The construction methods based on the experiences on the Nuclear Island, which is a critical path in the total construction schedule, have been studied and reconsidered in order to construct by more reliable and economical method. So various improved construction method are being applied and the duration of construction is being reduced continuously. So various improved construction method are being applied and the duration of construction is being reduced continuously. In this paper, the history of construction of twenty-three (23) PWR Plant, the actual construction methods and schedule of Ohi-3/4, to which the many improved methods were applied during their construction, are introduced mainly with the improved points for previously constructed plants. And also the situation of construction method for the next PWR Plant is simply explained

  10. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  11. Fuel rod behavior of a PWR during load following

    International Nuclear Information System (INIS)

    Perrotta, J.A.; Andrade, G.G. de

    1982-01-01

    The behavior of a PWR fuel rod when operating in normal power cycles, excluding in case of accidents, is analysed. A computer code, that makes the mechanical analysis of the cladding using the finite element method was developed. The ramps and power cycles were simulated suposing the existence of cracks in pellets when the cladding-pellet interaction are done. As a result, an operation procedure of the fuel rod in power cycle is recommended. (E.G.) [pt

  12. Corrosion of PWR steam generators

    International Nuclear Information System (INIS)

    Garnsey, R.

    1979-01-01

    Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

  13. PWR system reliability improvement activities

    International Nuclear Information System (INIS)

    Yoshikawa, Yuichiro

    1985-01-01

    In Japan lacking in energy resources, it is our basic energy policy to accelerate the development program of nuclear power, thereby reducing our dependence. As referred to in the foregoing, every effort has been exerted on our part to improve the PWR system reliability by dint of the so-called 'HOMEMADE' TQC activities, which is our brain-child as a result of applying to the energy industry the quality control philosophy developed in the field of manufacturing industry

  14. GALILEO ORBITER V POS VENUS TRAJECTORY V1.0

    Data.gov (United States)

    National Aeronautics and Space Administration — Galileo Orbiter 60 second sampled trajectory data from the Venus flyby in Venus Solar Orbital (VSO) coordinates. These data cover the interval 1990-02-09 00:00 to...

  15. PWR secondary water chemistry guidelines: Revision 3

    International Nuclear Information System (INIS)

    Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

    1993-05-01

    An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

  16. VICI (Venus In Situ Composition Investigations): The Next Step in Understanding Venus Climate Evolution

    Science.gov (United States)

    Glaze, L. S.; Garvin, J. B.

    2017-12-01

    Venus provides a natural laboratory to explore an example of terrestrial planet evolution that may be cosmically ubiquitous. By better understanding the composition of the Venus atmosphere and surface, we can better constrain the efficiency of the Venusian greenhouse. VICI is a proposed NASA New Frontiers mission that delivers two landers to Venus on two separate Venus fly-bys. Following six orbital remote sensing missions to Venus (since 1978), VICI would be the first mission to land on the Venus surface since 1985, and the first U.S. mission to enter the Venus atmosphere in 49 years. The four major VICI science objectives are: Atmospheric origin and evolution: Understand the origin of the Venus atmosphere, how it has evolved, including how recently Venus lost its oceans, and how and why it is different from the atmospheres of Earth and Mars, through in situ measurements of key noble gases, nitrogen, and hydrogen. Atmospheric composition and structure: Reveal the unknown chemical processes and structure in Venus' deepest atmosphere that dominate the current climate through two comprehensive, in situ vertical profiles. Surface properties and geologic evolution: For the first time ever, explore the tessera from the surface, specifically to test hypotheses of ancient content-building cycles, erosion, and links to past climates using multi-point mineralogy, elemental chemistry, imaging and topography. Surface-atmosphere interactions: Characterize Venus' surface weathering environment and provide insight into the sulfur cycle at the surface-atmosphere interface by integrating rich atmospheric composition and structure datasets with imaging, surface mineralogy, and elemental rock composition. VICI is designed to study Venus' climate history through detailed atmospheric composition measurements not possible on earlier missions. In addition, VICI images the tessera surface during descent enabling detailed topography to be generated. Finally, VICI makes multiple elemental

  17. Mantle plumes on Venus revisited

    Science.gov (United States)

    Kiefer, Walter S.

    1992-01-01

    The Equatorial Highlands of Venus consist of a series of quasicircular regions of high topography, rising up to about 5 km above the mean planetary radius. These highlands are strongly correlated with positive geoid anomalies, with a peak amplitude of 120 m at Atla Regio. Shield volcanism is observed at Beta, Eistla, Bell, and Atla Regiones and in the Hathor Mons-Innini Mons-Ushas Mons region of the southern hemisphere. Volcanos have also been mapped in Phoebe Regio and flood volcanism is observed in Ovda and Thetis Regiones. Extensional tectonism is also observed in Ovda and Thetis Regiones. Extensional tectonism is also observed in many of these regions. It is now widely accepted that at least Beta, Atla, Eistla, and Bell Regiones are the surface expressions of hot, rising mantel plumes. Upwelling plumes are consistent with both the volcanism and the extensional tectonism observed in these regions. The geoid anomalies and topography of these four regions show considerable variation. Peak geoid anomalies exceed 90 m at Beta and Atla, but are only 40 m at Eistla and 24 m at Bell. Similarly, the peak topography is greater at Beta and Atla than at Eistla and Bell. Such a range of values is not surprising because terrestrial hotspot swells also have a side range of geoid anomalies and topographic uplifts. Kiefer and Hager used cylindrical axisymmetric, steady-state convection calculations to show that mantle plumes can quantitatively account for both the amplitude and the shape of the long-wavelength geoid and topography at Beta and Atla. In these models, most of the topography of these highlands is due to uplift by the vertical normal stress associated with the rising plume. Additional topography may also be present due to crustal thickening by volcanism and crustal thinning by rifting. Smrekar and Phillips have also considered the geoid and topography of plumes on Venus, but they restricted themselves to considering only the geoid-topography ratio and did not

  18. Large Volcanic Rises on Venus

    Science.gov (United States)

    Smrekar, Suzanne E.; Kiefer, Walter S.; Stofan, Ellen R.

    1997-01-01

    Large volcanic rises on Venus have been interpreted as hotspots, or the surface manifestation of mantle upwelling, on the basis of their broad topographic rises, abundant volcanism, and large positive gravity anomalies. Hotspots offer an important opportunity to study the behavior of the lithosphere in response to mantle forces. In addition to the four previously known hotspots, Atla, Bell, Beta, and western Eistla Regiones, five new probable hotspots, Dione, central Eistla, eastern Eistla, Imdr, and Themis, have been identified in the Magellan radar, gravity and topography data. These nine regions exhibit a wider range of volcano-tectonic characteristics than previously recognized for venusian hotspots, and have been classified as rift-dominated (Atla, Beta), coronae-dominated (central and eastern Eistla, Themis), or volcano-dominated (Bell, Dione, western Eistla, Imdr). The apparent depths of compensation for these regions ranges from 65 to 260 km. New estimates of the elastic thickness, using the 90 deg and order spherical harmonic field, are 15-40 km at Bell Regio, and 25 km at western Eistla Regio. Phillips et al. find a value of 30 km at Atla Regio. Numerous models of lithospheric and mantle behavior have been proposed to interpret the gravity and topography signature of the hotspots, with most studies focusing on Atla or Beta Regiones. Convective models with Earth-like parameters result in estimates of the thickness of the thermal lithosphere of approximately 100 km. Models of stagnant lid convection or thermal thinning infer the thickness of the thermal lithosphere to be 300 km or more. Without additional constraints, any of the model fits are equally valid. The thinner thermal lithosphere estimates are most consistent with the volcanic and tectonic characteristics of the hotspots. Estimates of the thermal gradient based on estimates of the elastic thickness also support a relatively thin lithosphere (Phillips et al.). The advantage of larger estimates of

  19. Comparative pick-up ion distributions at Mars and Venus: Consequences for atmospheric deposition and escape

    Science.gov (United States)

    Curry, Shannon M.; Luhmann, Janet; Ma, Yingjuan; Liemohn, Michael; Dong, Chuanfei; Hara, Takuya

    2015-09-01

    Without the shielding of a substantial intrinsic dipole magnetic field, the atmospheres of Mars and Venus are particularly susceptible to similar atmospheric ion energization and scavenging processes. However, each planet has different attributes and external conditions controlling its high altitude planetary ion spatial and energy distributions. This paper describes analogous test particle simulations in background MHD fields that allow us to compare the properties and fates, precipitation or escape, of the mainly O+ atmospheric pick-up ions at Mars and Venus. The goal is to illustrate how atmospheric and planetary scales affect the upper atmospheres and space environments of our terrestrial planet neighbors. The results show the expected convection electric field-related hemispheric asymmetries in both precipitation and escape, where the degree of asymmetry at each planet is determined by the planetary scale and local interplanetary field strength. At Venus, the kinetic treatment of O+ reveals a strong nightside source of precipitation while Mars' crustal fields complicate the simple asymmetry in ion precipitation and drive a dayside source of precipitation. The pickup O+ escape pattern at both Venus and Mars exhibits low energy tailward escape, but Mars exhibits a prominent, high energy 'polar plume' feature in the hemisphere of the upward convection electric field while the Venus ion wake shows only a modest poleward concentration. The overall escape is larger at Venus than Mars (2.1 ×1025 and 4.3 ×1024 at solar maximum, respectively), but the efficiency (likelihood) of O+ escaping is 2-3 times higher at Mars. The consequences of these comparisons for pickup ion related atmospheric energy deposition, loss rates, and detection on spacecraft including PVO, VEX, MEX and MAVEN are considered. In particular, both O+ precipitation and escape show electric field controlled asymmetries that grow with energy, while the O+ fluxes and energy spectra at selected spatial

  20. Advanced accumulator for PWR

    International Nuclear Information System (INIS)

    Ichimura, Taiki; Chikahata, Hideyuki

    1997-01-01

    Advanced accumulators have been incorporated into the APWR design in order to simplify the safety system configuration and to improve reliability. The advanced accumulators refill the reactor vessel with a large discharge flow rate in a large LOCA, then switch to a small flow rate to continue safety injection for core reflooding. The functions of the conventional accumulator and the low head safety injection pump are integrated into this advanced accumulator. Injection performance tests simulating LOCA conditions and visualization tests for new designs have been carried out. This paper describes the APWR ECCS configuration, the advanced accumulator design and some of the injection performance and visualization test results. It was verified that the flow resistance of the advanced accumulator is independent of the model scale. The similarity law and performance data of the advanced accumulator for applying APWR was established. (author)

  1. Material property changes of stainless steels under PWR irradiation

    International Nuclear Information System (INIS)

    Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

    2009-01-01

    Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

  2. Innovative measurement within the atmosphere of Venus.

    Science.gov (United States)

    Ekonomov, Alexey; Linkin, Vyacheslav; Manukin, Anatoly; Makarov, Vladislav; Lipatov, Alexander

    The results of Vega project experiments with two balloons flew in the cloud layer of the atmosphere of Venus are analyzed as to the superrotation nature and local dynamic and thermodynamic characteristics of the atmosphere. These balloons in conjunction with measurements of temperature profiles defined by the Fourier spectrometer measurements from the spacecraft Venera 15 allow us to offer a mechanism accelerating the atmosphere to high zonal velocities and supporting these speeds, the atmosphere superrotation in general. Spectral measurements with balloons confirm the possibility of imaging the planet's surface from a height of not more than 55 km. Promising experiments with balloons in the atmosphere of Venus are considered. In particular, we discuss the possibility of measuring the geopotential height, as Venus no seas and oceans to vertical positioning of the temperature profiles. As an innovative research facilities within the atmosphere overpressure balloon with a lifetime longer than 14 Earth days and vertical profile microprobes are considered.

  3. Transits of Venus and Mercury as muses

    Science.gov (United States)

    Tobin, William

    2013-11-01

    Transits of Venus and Mercury have inspired artistic creation of all kinds. After having been the first to witness a Venusian transit, in 1639, Jeremiah Horrocks expressed his feelings in poetry. Production has subsequently widened to include songs, short stories, novels, novellas, sermons, theatre, film, engravings, paintings, photography, medals, sculpture, stained glass, cartoons, stamps, music, opera, flower arrangements, and food and drink. Transit creations are reviewed, with emphasis on the English- and French-speaking worlds. It is found that transits of Mercury inspire much less creation than those of Venus, despite being much more frequent, and arguably of no less astronomical significance. It is suggested that this is primarily due to the mythological associations of Venus with sex and love, which are more powerful and gripping than Mercury's mythological role as a messenger and protector of traders and thieves. The lesson for those presenting the night sky to the public is that sex sells.

  4. Venus: radar determination of gravity potential.

    Science.gov (United States)

    Shapiro, I I; Pettengill, G H; Sherman, G N; Rogers, A E; Ingalls, R P

    1973-02-02

    We describe a method for the determination of the gravity potential of Venus from multiple-frequency radar measurements. The method is based on the strong frequency dependence of the absorption of radio waves in Venus' atmosphere. Comparison of the differing radar reflection intensities at several frequencies yields the height of the surface relative to a reference pressure contour; combination with measurements of round-trip echo delays allows the pressure, and hence the gravity potential contour, to be mapped relative to the mean planet radius. Since calibration data from other frequencies are unavailable, the absorption-sensitive Haystack Observatory data have been analyzed under the assumption of uniform surface reflectivity to yield a gravity equipotential contour for the equatorial region and a tentative upper bound of 6 x 10(-4) on the fractional difference of Venus' principal equatorial moments of inertia. The minima in the equipotential contours appear to be associated with topographic minima.

  5. Modeling Venus-like Worlds Through Time and Implications for the Habitable Zone

    Science.gov (United States)

    Way, M.; Del Genio, A. D.; Amundsen, D. S.; Sohl, L. E.; Kiang, N. Y.; Aleinov, I. D.; Kelley, M.

    2017-12-01

    In recent work [1] we demonstrated that the climatic history of Venus may have allowed for surface liquid water to exist for several billion years using a 3D GCM [2]. Model resolution was 4x5 latitude x longitude, 20 atmospheric layers and a 13 layer fully coupled ocean. Several assumptions were made based on what data we have for early Venus: a.) Used a solar spectrum from 2.9 billion years ago, and 715 million years ago for the incident radiation. b.) Assumed Venus had the same slow modern retrograde rotation throughout the 2.9 to 0.715 Gya history explored, although one simulation at faster rotation rate was shown not to be in the HZ. c.) Used atmospheric constituents similar to modern Earth: 1 bar N2, 400ppmv CO2, 1ppmv CH4. d.) Gave the planet a shallow 310m deep ocean constrained by published D/H ratio observations. e.) Used present day Venus topography and one run with Earth topography.In all cases except the faster rotating one the planet was able to maintain surface liquid water. We have now inserted the SOCRATES [3] radiation scheme into our 3D GCM to more accurately calculate heating fluxes for different atmospheric constituents. Using SOCRATES we have explored a number of other possible early histories for Venus including: f.) An aquaplanet configuration at 2.9Gya with present day rotation period.g.) A Land planet configuration at 2.9Gya with the equivalent of 10m of water in soil and lakes. h.) A synchronously rotating version of a, f, and g (supported by recent work of [4] and older work of [5]) i.) A Venus topography with a 310m ocean, but using present day insolation (1.9 x Earth). j.) Versions of most of the worlds above but with solar insolations >1.9 to explore more Venus-like exoplanetary worlds around G-type stars. In these additional cases the planet still resides in the liquid water habitable zone. Studies such as these should help Astronomers better understand whether exoplanets found in the Venus zone [6] are capable of hosting liquid water

  6. Venus project : experimentation at ENEA`s pilot site

    Energy Technology Data Exchange (ETDEWEB)

    Bargellini, M L; Fontana, F [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dip. Innovazione; Bucci, C; Ferrara, F; Sottile, P A [GESI s.r.l., Rome (Italy); Niccolai, L; Scavino, G [Rome Univ. Sacro Cuore (Italy); Mancini, R; Levialdi, S [Rome Univ. La Sapienza (Italy). Dip. di Scienze dell` Informazione

    1996-12-01

    The document describes the ENEA`s (Italian Agency for New Technologies, Energy and the Environment) experience in the Venus Project (Esprit III ). Venus is an advanced visual interface based on icon representation that permits to end-user to inquiry databases. VENUS interfaces to ENEA`s databases: cometa materials Module, Cometa Laboratories Module and European Programs. This report contents the results of the experimentation and of the validation carried out in ENEA`s related to the Venus generations. Moreover, the description of the architecture, the user requirements syntesis and the validation methodology of the VENUS systems have been included.

  7. Conceptual design of simplified PWR

    International Nuclear Information System (INIS)

    Tabata, Hiroaki

    1996-01-01

    The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

  8. Influences of boric acid and lithium hydroxide on oxide film of type 316 stainless steel in PWR simulated primary water; PWR 1次冷却材模擬環境中の316ステンレス鋼に生成した皮膜性状に及ぼすほう酸および水酸化リチウムの影響

    Energy Technology Data Exchange (ETDEWEB)

    Fukumura, Takuya; Fukuya, Koji; Arioka, Koji [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan)

    2012-06-15

    In order to understand the influences of boric acid and lithium hydroxide on the IGSCC of type 316 stainless steel, an oxide film was analyzed in simulated PWR primary water while varying the boric acid and lithium hydroxide concentrations. It was found that, although boric acid and lithium hydroxide did not affect the structure and chemical composition of the surface oxide film remarkably, a lower boric acid concentration or a higher lithium concentration produced an oxide film with a thicker surface. It was considered that the lower boric acid concentration and higher lithium hydroxide concentration caused a higher magnetite solubility at the surface of the material and that the higher magnetite solubility caused a higher iron concentration gradient, which promoted iron dissolution from the material and the formation of a thicker oxide film. It was found that the thicker oxide film caused a higher IGSCC susceptibility and that the corrosion was the dominant factor of the IGSCC mechanism. No significant change was found in the morphologies of crack tip oxide in different bulk water chemistry systems, thus producing CT specimens with similar crack growth rates. (author)

  9. Galileo infrared imaging spectroscopy measurements at venus

    Science.gov (United States)

    Carlson, R.W.; Baines, K.H.; Encrenaz, Th.; Taylor, F.W.; Drossart, P.; Kamp, L.W.; Pollack, James B.; Lellouch, E.; Collard, A.D.; Calcutt, S.B.; Grinspoon, D.; Weissman, P.R.; Smythe, W.D.; Ocampo, A.C.; Danielson, G.E.; Fanale, F.P.; Johnson, T.V.; Kieffer, H.H.; Matson, D.L.; McCord, T.B.; Soderblom, L.A.

    1991-01-01

    During the 1990 Galileo Venus flyby, the Near Infrared Mapping Spectrometer investigated the night-side atmosphere of Venus in the spectral range 0.7 to 5.2 micrometers. Multispectral images at high spatial resolution indicate substantial cloud opacity variations in the lower cloud levels, centered at 50 kilometers altitude. Zonal and meridional winds were derived for this level and are consistent with motion of the upper branch of a Hadley cell. Northern and southern hemisphere clouds appear to be markedly different. Spectral profiles were used to derive lower atmosphere abundances of water vapor and other species.

  10. Astrobiology: The Case for Venus

    Science.gov (United States)

    Landis, Geoffrey A.

    2003-01-01

    The scientific discipline of astrobiology addresses one of the most fundamental unanswered questions of science: are we alone? Is there life elsewhere in the universe, or is life unique to Earth? The field of astrobiology includes the study of the chemical precursors for life in the solar system; it also includes the search for both presently existing life and fossil signs of previously existing life elsewhere in our own solar system, as well as the search for life outside the solar system. Two of the promising environments within the solar system being currently considered are the surface of the planet Mars, and the hypothesized oceans underneath the ice covering the moon Europa. Both of these environments differ in several key ways from the environments where life is found on Earth; the Mars environment in most places too cold and at too low pressure for liquid water to be stable, and the sub-ice environment of Europa lacking an abundance of free energy in the form of sunlight. The only place in the solar system where we know that life exists today is the Earth. To look for life elsewhere in the solar system, one promising search strategy would be to find and study the environment in the solar system with conditions that are most similar to the environmental conditions where life thrives on the Earth. Specifically, we would like to study a location in the solar system with atmospheric pressure near one bar; temperature in the range where water is liquid, 0 to 100 C; abundant solar energy; and with the primary materials required for life, carbon, oxygen, nitrogen, and hydrogen, present. Other than the surface of the Earth, the only other place where these conditions exist is the atmosphere of Venus, at an altitude of about fifty kilometers above the surface.

  11. Surveillance of vibrations in PWR

    International Nuclear Information System (INIS)

    Espefaelt, R.; Lorenzen, J.; Aakerhielm, F.

    1980-07-01

    The core of a PWR - including fuel elements, internal structure, control rods and core support structure inside the pressure vessel - is subjected to forces which can cause vibrations. One sensitive means to detect and analyse such vibrations is by means of the noise from incore and excore neutron detector signals. In this project noise recordings have been made on two occasions in the Ringhals 2 plant and the obtained data been analysed using the Studsvik Noise Analysis Program System (SNAPS). The results have been intepreted and a detailed description of the vibrational status of the core and pressure vessel internals has been produced. On the basis of the obtained results it is proposed that neutron signal noise analysis should be performed at each PWR plant in the beginning, middle and end of each fuel cycle and an analysis be made using the methods developed in the project. It would also provide a contribution to a higher degree of preparedness for diagnostic tasks in case of unexpected and abnormal events. (author)

  12. Analytical technical of lightning surges induced on grounding mesh of PWR nuclear power plant

    International Nuclear Information System (INIS)

    Ikeda, I.; Tani, M.; Yonezawa, T.

    1990-01-01

    An analytical lightning surge technique is needed to make a qualitative and predictive evaluation of transient voltages induced on local grounding meshes and instrumentation cables by a lightning strike on a lightning rod in a PWR plant. This paper discusses an experiment with lightning surge impulses in a PWR plant which was setup to observe lightning caused transient voltages. Experimental data when compared with EMTP simulation results improved the simulation method. The improved method provides a good estimation of induced voltages on grounding meshes and instrumentation cables

  13. Performance of high burned PWR fuel during transient

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Fujishiro, Toshio

    1992-01-01

    In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

  14. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  15. Practical Observations of the Transit of Venus

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 9; Issue 5. Practical Observations of the Transit of Venus. B S Shyalaja. Classroom Volume 9 Issue 5 May 2004 pp 79-83. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/009/05/0079-0083 ...

  16. Tidal constraints on the interior of Venus

    Science.gov (United States)

    Dumoulin, C.; Tobie, G.; Verhoeven, O.; Rosenblatt, P.; Rambaux, N.

    2017-12-01

    As a prospective study for a future exploration of Venus, we compute the tidal response of Venus' interior assuming various mantle compositions and temperature profiles representative of different scenarios of Venus' formation and evolution. The mantle density and seismic velocities are modeled from thermodynamical equilibria of mantle minerals and used to predict the moment of inertia, Love numbers, and tide-induced phase lag characterizing the signature of the internal structure in the gravity field. The viscoelasticity of the mantle is parameterized using an Andrade rheology. From the models considered here, the moment of inertia lies in the range of 0.327 to 0.342, corresponding to a core radius of 2900 to 3450 km. Viscoelasticity of the mantle strongly increases the potential Love number relative to previously published elastic models. Due to the anelasticity effects, we show that the possibility of a completely solid metal core inside Venus cannot be ruled out based on the available estimate of k2 from the Magellan mission (Konopliv and Yoder, 1996). A Love number k2 lower than 0.27 would indicate the presence of a fully solid iron core, while for larger values, solutions with an entirely or partially liquid core are possible. Precise determination of the Love numbers, k2 and h2, together with an estimate of the tidal phase lag, are required to determine the state and size of the core, as well as the composition and viscosity of the mantle.

  17. 10. The surface and interior of venus

    Science.gov (United States)

    Masursky, H.; Kaula, W.M.; McGill, G.E.; Pettengill, G.H.; Phillips, R.J.; Russell, C.T.; Schubert, G.; Shapiro, I.I.

    1977-01-01

    Present ideas about the surface and interior of Venus are based on data obtained from (1) Earth-based radio and radar: temperature, rotation, shape, and topography; (2) fly-by and orbiting spacecraft: gravity and magnetic fields; and (3) landers: winds, local structure, gamma radiation. Surface features, including large basins, crater-like depressions, and a linear valley, have been recognized from recent ground-based radar images. Pictures of the surface acquired by the USSR's Venera 9 and 10 show abundant boulders and apparent wind erosion. On the Pioneer Venus 1978 Orbiter mission, the radar mapper experiment will determine surface heights, dielectric constant values and small-scale slope values along the sub-orbital track between 50??S and 75??N. This experiment will also estimate the global shape and provide coarse radar images (40-80 km identification resolution) of part of the surface. Gravity data will be obtained by radio tracking. Maps combining radar altimetry with spacecraft and ground-based images will be made. A fluxgate magnetometer will measure the magnetic fields around Venus. The radar and gravity data will provide clues to the level of crustal differentiation and tectonic activity. The magnetometer will determine the field variations accurately. Data from the combined experiments may constrain the dynamo mechanism; if so, a deeper understanding of both Venus and Earth will be gained. ?? 1977 D. Reidel Publishing Company.

  18. Abrir una Venus: Hablar con ella

    Directory of Open Access Journals (Sweden)

    Ginnette Barrantes Sáenz

    2013-09-01

    Se propone a Alicia como la Venus abierta que  incita, mediante  la  cita cinematográfica del cine mudo en el cine de Almodóvar, la no tan conocida figura de  amar a una  dormida( Allouch, 2005

  19. Solar Airplane Concept Developed for Venus Exploration

    Science.gov (United States)

    Landis, Geoffrey A.

    2004-01-01

    An airplane is the ideal vehicle for gathering atmospheric data over a wide range of locations and altitudes, while having the freedom to maneuver to regions of scientific interest. Solar energy is available in abundance on Venus. Venus has an exoatmospheric solar flux of 2600 W/m2, compared with Earth's 1370 W/m2. The solar intensity is 20 to 50 percent of the exoatmospheric intensity at the bottom of the cloud layer, and it increases to nearly 95 percent of the exoatmospheric intensity at 65 km. At these altitudes, the temperature of the atmosphere is moderate, in the range of 0 to 100 degrees Celsius, depending on the altitude. A Venus exploration aircraft, sized to fit in a small aeroshell for a "Discovery" class scientific mission, has been designed and analyzed at the NASA Glenn Research Center. For an exploratory aircraft to remain continually illuminated by sunlight, it would have to be capable of sustained flight at or above the wind speed, about 95 m/sec at the cloud-top level. The analysis concluded that, at typical flight altitudes above the cloud layer (65 to 75 km above the surface), a small aircraft powered by solar energy could fly continuously in the atmosphere of Venus. At this altitude, the atmospheric pressure is similar to pressure at terrestrial flight altitudes.

  20. 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.

  1. Sampling the Cloudtop Region on Venus

    Science.gov (United States)

    Limaye, Sanjay; Ashish, Kumar; Alam, Mofeez; Landis, Geoffrey; Widemann, Thomas; Kremic, Tibor

    2014-05-01

    The details of the cloud structure on Venus continue to be elusive. One of the main questions is the nature and identity of the ultraviolet absorber(s). Remote sensing observations from Venus Express have provided much more information about the ubiquitous cloud cover on Venus from both reflected and emitted radiation from Venus Monitoring Camera (VMC) and Visible InfraRed Imaging Spectrometer (VIRTIS) observations. Previously, only the Pioneer Venus Large Probe has measured the size distribution of the cloud particles, and other probes have measured the bulk optical properties of the cloud cover. However, the direct sampling of the clouds has been possible only below about 62 km, whereas the recent Venus Express observations indicate that the cloud tops extend from about 75 km in equatorial region to about 67 km in polar regions. To sample the cloud top region of Venus, other platforms are required. An unmanned aerial vehicle (UAV) has been proposed previously (Landis et al., 2002). Another that is being looked into, is a semi-buoyant aerial vehicle that can be powered using solar cells and equipped with instruments to not only sample the cloud particles, but also to make key atmospheric measurements - e.g. atmospheric composition including isotopic abundances of noble and other gases, winds and turbulence, deposition of solar and infrared radiation, electrical activity. The conceptual design of such a vehicle can carry a much more massive payload than any other platform, and can be controlled to sample different altitudes and day and night hemispheres. Thus, detailed observations of the surface using a miniature Synthetic Aperture Radar are possible. Data relay to Earth will need an orbiter, preferably in a low inclination orbit, depending on the latitude region selected for emphasis. Since the vehicle has a large surface area, thermal loads on entry are low, enabling deployment without the use of an aeroshell. Flight characteristics of such a vehicle have been

  2. Future Drag Measurements from Venus Express

    Science.gov (United States)

    Keating, Gerald; Mueller-Wodarg, Ingo; Forbes, Jeffrey M.; Yelle, Roger; Bruinsma, Sean; Withers, Paul; Lopez-Valverde, Miguel Angel; Theriot, Res. Assoc. Michael; Bougher, Stephen

    Beginning in July 2008 during the Venus Express Extended Mission, the European Space Agency will dramatically drop orbital periapsis from near 250km to near 180km above the Venus North Polar Region. This will allow orbital decay measurements of atmospheric densities to be made near the Venus North Pole by the VExADE (Venus Express Atmospheric Drag Experiment) whose team leader is Ingo Mueller-Wodarg. VExADE consists of two parts VExADE-ODA (Orbital Drag Analysis from radio tracking data) and VExADE-ACC (Accelerometer in situ atmospheric density measurements). Previous orbital decay measurements of the Venus thermosphere were obtained by Pioneer Venus from the 1970's into the 1990's and from Magellan in the 1990's. The major difference is that the Venus Express will provide measurements in the North Polar Region on the day and night sides, while the earlier measurements were obtained primarily near the equator. The periapsis will drift upwards in altitude similar to the earlier spacecraft and then be commanded down to its lower original values. This cycle in altitude will allow estimates of vertical structure and thus thermospheric temperatures in addition to atmospheric densities. The periapsis may eventually be lowered even further so that accelerometers can more accurately obtain density measurements of the polar atmosphere as a function of altitude, latitude, longitude, local solar time, pressure, Ls, solar activity, and solar wind on each pass. Bias in accelerometer measurements will be determined and corrected for by accelerometer measurements obtained above the discernable atmosphere on each pass. The second experiment, VExADE-ACC, is similar to the accelerometer experiments aboard Mars Global Surveyor, Mars Odyssey, and Mars Reconnaissance Orbiter that carried similar accelerometers in orbit around Mars. The risk involved in the orbital decay and accelerometer measurements is minimal. We have not lost any spacecraft orbiting Venus or Mars due to unexpected

  3. PWR thermocouple mechanical sealing structure

    International Nuclear Information System (INIS)

    Shen Qiuping; He Youguang

    1991-08-01

    The PWR in-core temperature detection device, which is one of measures to insure reactor safety operation, is to monitor and diagnose reactor thermal power output and in-core power distribution. The temperature detection device system uses thermocouples as measuring elements with stainless steel protecting sleeves. The thermocouple has a limited service time and should be replaced after its service time has reached. A new sealing device for the thermocouples of reactor in-core temperature detection system has been developed to facilitate replacement. The structure is complete tight under high temperature and pressure without any leakage and seepage, and easy to be assembled or disassembled in radioactive environment. The device is designed to make it possible to replace the thermocouple one by one if necessary. This is a new, simple and practical structure

  4. PWR standardization: The French experience

    International Nuclear Information System (INIS)

    Bacher, P.E.

    1987-01-01

    After a short historical review of the French PWR programme with 45000 MWe in operation and 15000 MWe under construction, the paper first develops the objectives and limits of the standardizatoin policy. Implementation of standardization is described through successive reactor series and feedback of experience, together with its impact on safety and on codes and standards. Present benefits of standardization range from low engineering costs to low backfitting costs, via higher quality, reduction in construction times and start-up schedules and improved training of operators. The future of the French programme into the 1990's is again with an advanced standardized series, the N4-1400 MW plant. There is no doubt that the very positive experience with standardization is relevant to any country trying to achieve self-reliance in the nuclear power field. (author)

  5. Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

    International Nuclear Information System (INIS)

    Lee, C.B.

    2002-01-01

    CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

  6. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Directory of Open Access Journals (Sweden)

    Krása Antonín

    2017-01-01

    Full Text Available VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector. Discrepancies between experiments and Monte Carlo calculations (MCNP5 of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2 are presented.

  7. Comparative study on neutron data in integral experiments of MYRRHA mockup critical cores in the VENUS-F reactor

    Science.gov (United States)

    Krása, Antonín; Kochetkov, Anatoly; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente

    2017-09-01

    VENUS-F is a fast, zero-power reactor with 30% wt. metallic uranium fuel and solid lead as coolant simulator. It serves as a mockup of the MYRRHA reactor core. This paper describes integral experiments performed in two critical VENUS-F core configurations (with and without graphite reflector). Discrepancies between experiments and Monte Carlo calculations (MCNP5) of keff, fission rate spatial distribution and reactivity effects (lead void and fuel Doppler) depending on a nuclear data library used (JENDL-4.0, ENDF-B-VII.1, JEFF-3.1.2, 3.2, 3.3T2) are presented.

  8. Babcock and Wilcox advanced PWR development

    International Nuclear Information System (INIS)

    Kulynych, G.E.; Lemon, J.E.

    1986-01-01

    The Babcock and Wilcox 600 MWe PWR design is discussed. Main features of the new B-600 design are improvements in reactor system configuration, glandless coolant pumps, safety features, core design and steam generators

  9. Hydrogen production in a PWR during LOCA

    International Nuclear Information System (INIS)

    Cassette, P.

    1983-12-01

    The purpose of this paper is to provide information on hydrogen generation during LOCA in French 900 MW PWR power plants. The design basis accident is taken into account as well as more severe accidents assuming failure of emergency systems

  10. Economic optimization of PWR cores with ROSA

    International Nuclear Information System (INIS)

    Verhagen, F.C.M.; Wakker, P.H.

    2005-01-01

    The core-loading pattern is decisive for fuel cycle economics, fuel safety parameters and economic planning for future cycles. ROSA, NRG's loading pattern optimization code system for PWRs, has proven for over a decade to be a valuable tool to reactor operators for improving their fuel management economics. ROSA uses simulated annealing as loading pattern optimization technique, in combination with an extremely fast 3-D neutronics code for loading pattern calculations. The code is continuously extended with new optimization parameters and rules. This paper outlines recent developments of the ROSA code system and discusses results of PWR specific applications of ROSA. Core designs with a large variety of challenging constraints have been realized with ROSA. As a typical example, for the 193 assembly, Vantage 5H/RFA-2 fueled TVA's Watts Bar unit 1, a cycle 4 core with 76 feed assemblies was designed. This was followed by a high-energy cycle 5 with only 77 feed assemblies and approximately 535 days of natural cycle length. Subsequently, an economical core using 72 bundles was designed for cycle 6. This resulted in considerable savings in the cost of feed assemblies for reloads. The typical accuracy of ROSA compared to results of license codes in within ±0.02 for normalized assembly powers, ±0.03 for maximum enthalpy rise hot channel factor (F ΔH ), and ±3 days for natural cycle length. (author)

  11. The Venus Emissivity Mapper - gaining a global perspective on the surface composition of Venus

    Science.gov (United States)

    Helbert, Joern; Dyar, Melinda; Widemann, Thomas; Marcq, Emmanuel; Maturilli, Alessandro; Mueller, Nils; Kappel, David; Ferrari, Sabrina; D'Amore, Mario; Tsang, Constantine; Arnold, Gabriele; Smrekar, Suzanne; VEM Team

    2017-10-01

    The permanent cloud cover of Venus prohibits observations of the surface with traditional imaging techniques over much of the EM spectral range, leading to the false notion that information about the composition of Venus’ surface could only be derived from lander missions. However, harsh environmental conditions on the surface cause landed missions to be sole site, highly complex, and riskier than orbiting missions.It is now known that 5 transparency windows occur in the Venus atmosphere, ranging from 0.86 µm to 1.18 µm. Recent advances in high temperature laboratory spectroscopy at the PSL at DLR these windows are highly diagnostic for surface mineralogy. Mapping of the southern hemisphere of Venus with VIRTIS on VEX in the 1.02 µm band was a proof-of-concept for an orbital remote sensing approach to surface composition and weathering studies[1-3]. The Venus Emissivity Mapper [4] proposed for the NASA’s Venus Origins Explorer (VOX) and the ESA EnVision proposal builds on these recent advances. It is the first flight instrument specially designed with a sole focus on mapping the surface of Venus using the narrow atmospheric windows around 1 µm. Operating in situ from Venus orbit, VEM will provide a global map of surface composition as well as redox state of the surface, providing a comprehensive picture of surface-atmosphere interaction and support for landing site selection. Continuous observation of the thermal emission of the Venus will provide tight constraints on the current day volcanic activity[5]. This is complemented by measurements of atmospheric water vapor abundance as well as cloud microphysics and dynamics. These data will allow for accurate correction of atmospheric interference on the surface measurements, which provide highly valuable science on their own. A mission combining VEM with a high-resolution radar mapper such as VOX or EnVision in a low circular orbit will provide key insights into the divergent evolution of Venus.1. Smrekar, S

  12. MEETING VENUS. A Collection of Papers presented at the Venus Transit Conference Tromsoe 2012

    Science.gov (United States)

    Sterken, Christiaan; Aspaas, Per Pippin

    2013-05-01

    On 2-3 June 2012, the University of Tromsoe hosted a conference about the cultural and scientific history of the transits of Venus. The conference took place in Tromsoe for two very specific reasons. First and foremost, the last transit of Venus of this century lent itself to be observed on the disc of the Midnight Sun in this part of Europe during the night of 5 to 6 June 2012. Second, several Venus transit expeditions in this region were central in the global enterprise of measuring the scale of the solar system in the eighteenth century. The site of the conference was the Nordnorsk Vitensenter (Science Centre of Northern Norway), which is located at the campus of the University of Tromsoe. After the conference, participants were invited to either stay in Tromsoe until the midnight of 5-6 June, or take part in a Venus transit voyage in Finnmark, during which the historical sites Vardoe, Hammerfest, and the North Cape were to be visited. The post-conference program culminated with the participants observing the transit of Venus in or near Tromsoe, Vardoe and even from a plane near Alta. These Proceedings contain a selection of the lectures delivered on 2-3 June 2012, and also a narrative description of the transit viewing from Tromsoe, Vardoe and Alta. The title of the book, Meeting Venus, refers the title of a play by the Hungarian film director, screenwriter and opera director Istvan Szabo (1938-). The autobiographical movie Meeting Venus (1991) directed by him is based on his experience directing Tannhauser at the Paris Opera in 1984. The movie brings the story of an imaginary international opera company that encounters a never ending series of difficulties and pitfalls that symbolise the challenges of any multicultural and international endeavour. As is evident from the many papers presented in this book, Meeting Venus not only contains the epic tales of the transits of the seventeenth, eighteenth and nineteenth centuries, it also covers the conference

  13. The escape of natural satellites from Mercury and Venus

    International Nuclear Information System (INIS)

    Kumar, S.S.

    1977-01-01

    It is suggested that the slow rotations of Mercury and Venus may be connected with the absence of natural satellites around them. If Mercury or Venus possessed a satellite at the time of formation, the tidal evolution would have caused the satellite to recede. At a sufficiently large distance from the planet, the Sun's gravitational influence makes the satellite orbit unstable. The natural satellites of Mercury and Venus might have escaped as a consequence of this instability. (Auth.)

  14. Escape of natural satellites from Mercury and Venus

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, S S [Virginia Univ., Charlottesville (USA)

    1977-09-01

    It is suggested that the slow rotations of Mercury and Venus may be connected with the absence of natural satellites around them. If Mercury or Venus possessed a satellite at the time of formation, the tidal evolution would have caused the satellite to recede. At a sufficiently large distance from the planet, the Sun's gravitational influence makes the satellite orbit unstable. The natural satellites of Mercury and Venus might have escaped as a consequence of this instability.

  15. PWR core follow calculations using the ELCOS code system

    International Nuclear Information System (INIS)

    Grimm, P.; Paratte, J.M.

    1990-01-01

    The ELCOS code system developed at PSI is used to simulate a cycle of a PWR in which one fifth of the assemblies are MOX fuel. The reactor and the calculational methods are briefly described. The calculated critical boron concentrations and power distributions are compared with the measurements at the plant. Although the critical boron concentration is somewhat overpredicted and the computed power distributions are slightly flatter than the measured ones the results of the calculations agree generally well with the measured data. (author) 1 tab., 8 figs., 6 refs

  16. Study of the noise propagation in PWR with coupled codes

    International Nuclear Information System (INIS)

    Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

    2011-01-01

    The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

  17. Thermal-hydraulic analysis of PWR cores in transient condition

    International Nuclear Information System (INIS)

    Silva Galetti, M.R. da.

    1984-01-01

    A calculational methodology for thermal - hydraulic analysis of PWR cores under steady-state and transient condition was selected and made available to users. An evaluation of the COBRA-IIIP/MIT code, used for subchannel analysis, was done through comparison of the code results with experimental data on steady state and transient conditions. As a result, a comparison study allowing spatial and temporal localization of critical heat flux was obtained. A sensitivity study of the simulation model to variations in some empirically determined parameter is also presented. Two transient cases from Angra I FSAR were analysed, showing the evolution of minimum DNBR with time. (Author) [pt

  18. Conversion ratio in epithermal PWR, in thorium and uranium cycle

    International Nuclear Information System (INIS)

    Barroso, D.E.G.

    1982-01-01

    Results obtained for the conversion ratio in PWR reactors with close lattices, operating in thorium and uranium cycles, are presented. The study of those reactors is done in an unitary fuel cell of the lattices with several ratios V sub(M)/V sub(F), considering only the equilibrium cycles and adopting a non-spatial depletion calculation model, aiming to simulate mass flux of reactor heavy elements in the reactor. The neutronic analysis and the cross sections generation are done with Hammer computer code, with one critical apreciation about the application of this code in epithermal systems and with modifications introduced in the library of basic data. (E.G.) [pt

  19. Substitution of cobalt alloying in PWR primary circuit gate valves

    International Nuclear Information System (INIS)

    Cachon, L.; Sudreau, F.; Brunel, L.

    1995-01-01

    The object of this study is qualify cobalt-free alternative alloys for valve applications. This paper focus on tribological characterization of numerous coatings is done by using the first one, of a classical type. Then tests are performed with the second one which simulates solicitations supported by gate valves in primary circuit of PWR. 35% Ni-Cr - 65% Cr 3 C 2 coating, deposited by detonation gun technology, gives us hope to find a substitute of Stelite 6. (author). 5 refs., 16 figs., 2 tabs

  20. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  1. First stage of cosmic expedition Vega: Venus investigations

    International Nuclear Information System (INIS)

    Balebanov, V.M.; Moroz, V.I.; Mukhin, L.M.

    1985-01-01

    Main results of the first (Venus) stage of the international complex program ''Venus - Halley'' (''Vega'' for short) are presented. The program is aimed at transporting descent space vehicles to the Venus to explore its atmosphere and surface. Then automatic interplanetary stations (AIS) will be directed to the Halley's comet. In June 1985 the descent space vehicles AIS ''Vega-1'' and ''Vega-2'' have landed softly on the Venus surface, aerostat probes have been launched to the planet atmosphere. The design of the descent space vehicle, structure and chemical composition of the atmosphere, ground composition are briefly outlined

  2. Energetic particles at venus: galileo results.

    Science.gov (United States)

    Williams, D J; McEntire, R W; Krimigis, S M; Roelof, E C; Jaskulek, S; Tossman, B; Wilken, B; Stüdemann, W; Armstrong, T P; Fritz, T A; Lanzerotti, L J; Roederer, J G

    1991-09-27

    At Venus the Energetic Particles Detector (EPD) on the Galileo spacecraft measured the differential energy spectra and angular distributions of ions >22 kiloelectron volts (keV) and electrons > 15 keV in energy. The only time particles were observed by EPD was in a series of episodic events [0546 to 0638 universal time (UT)] near closest approach (0559:03 UT). Angular distributions were highly anisotropic, ordered by the magnetic field, and showed ions arriving from the hemisphere containing Venus and its bow shock. The spectra showed a power law form with intensities observed into the 120- to 280-keV range. Comparisons with model bow shock calculations show that these energetic ions are associated with the venusian foreshock-bow shock region. Shock-drift acceleration in the venusian bow shock seems the most likely process responsible for the observed ions.

  3. Electron plasma oscillations in the Venus foreshock

    Science.gov (United States)

    Crawford, G. K.; Strangeway, R. J.; Russell, C. T.

    1990-01-01

    Plasma waves are observed in the solar wind upstream of the Venus bow shock by the Pioneer Venus Orbiter. These wave signatures occur during periods when the interplanetary magnetic field through the spacecraft position intersects the bow shock, thereby placing the spacecraft in the foreshock region. The electron foreshock boundary is clearly evident in the data as a sharp onset in wave activity and a peak in intensity. Wave intensity is seen to drop rapidly with increasing penetration into the foreshock. The peak wave electric field strength at the electron foreshock boundary is found to be similar to terrestrial observations. A normalized wave spectrum was constructed using measurements of the electron plasma frequency and the spectrum was found to be centered about this value. These results, along with polarization studies showing the wave electric field to be field aligned, are consistent with the interpretation of the waves as electron plasma oscillations.

  4. Electron plasma oscillations in the Venus foreshock

    International Nuclear Information System (INIS)

    Crawford, G.K.; Strangeway, R.J.; Russell, C.T.

    1990-01-01

    Plasma waves are observed in the solar wind upstream of the Venus bow shock by the Pioneer Venus Orbiter. These wave signatures occur during periods when the interplanetary magnetic field through the spacecraft position intersects the bow shock, thereby placing the spacecraft in the foreshock region. The electron foreshock boundary is clearly evident in the data as a sharp onset in wave activity and a peak in intensity. Wave intensity is seen to drop rapidly with increasing penetration into the foreshock. The peak wave electric field strength at the electron foreshock boundary is found to be similar to terrestrial observations. A normalized wave spectrum was constructed using measurements of the electron plasma frequency and the spectrum was found to be centered about this value. These results, along with polarization studies showing the wave electric field to be field aligned, are consistent with the interpretation of the waves as electron plasma oscillations

  5. Venus y el fin del mundo

    Directory of Open Access Journals (Sweden)

    Gonzalo Munévar

    2006-01-01

    Full Text Available Este artículo busca demostrar que los argumentos generales acerca de la exploración científica valen también para las ciencias espaciales. El trabajo se basa en el ejemplo de la exploración de Venus y lo que esta nos dice acerca de nuestro propio planeta. Argumenta que el concepto de la probabilidad de Leslie es incorrecto, como también lo son las dudas sobre la evidencia Venusiana. Así mismo, concluye que no se puede rechazar la importancia que tienen los descubrimientos inesperados que han resultado de la exploración de Venus para ayudarnos a comprender nuestro propio planeta. Y que si van a ser rechazados estos descubrimientos debe ser por razones científicas, no por intuiciones acerca de la probabilidad.

  6. Deuterium content of the Venus atmosphere

    International Nuclear Information System (INIS)

    Bertaux, -J.-L.; Clarke, J.T.

    1989-01-01

    The abundance of deuterium in the atmosphere of Venus is an important clue to the planet's history, because ordinary and deuterated water escape at different rates. Using the high-resolution mode of the International Ultraviolet Explorer (IUE), we measured hydrogen Lyman-α-emission but found only an upper limit on deuterium Lyman-α-emission, from which we inferred a D/H ratio of less than 2-5 x 10 -3 . This is smaller by a factor of 3-8 than the D/H ratio derived from measurements by the Pioneer Venus Large Probe, and may indicate either a stratification of D/H ratio with altitude or a smaller overall ratio than previously thought. (author)

  7. Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Xiangjie Liu

    2014-01-01

    Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

  8. Analysis of dynamic behavior of a PWR utilizing the computer program SARDAN 2

    International Nuclear Information System (INIS)

    Pessanha, J.A.O.

    1982-07-01

    In the design of a PWR nuclear plant it is necessary to verify if the design limits are respected, even under abnormal operation condition. An evolution of SARDAN code, developed to simulate transients in PWR, are presented. The new aspects incorporeted in SARDAN 2 are: the fuel ROD analysis in finite-diference, an open channel model for the critic subchannel analysis and the introduction of a simplified model for the automatic control system. The program has been tested in accident condition II, in special, uncontrolled ROD cluster assembly bank withoraw, dropped full-length assembly group, uncontrolled Boron dilution, and the results obtained were considered satisfactory. (Author) [pt

  9. Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

    2015-01-01

    Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

  10. VLF imaging of the Venus foreshock

    Science.gov (United States)

    Crawford, G. K.; Strangeway, R. J.; Russell, C. T.

    1993-01-01

    VLF plasma wave measurements obtained from the Pioneer Venus Orbiter Electric Field Detector (OEFD) have been used to construct statistical images of the Venus foreshock. Our data set contains all upstream measurements from an entire Venus year (approximately 200 orbits). Since the foreshock VLF characteristics vary with Interplanetary Magnetic Field (IMF) orientation we restrict the study to IMF orientations near the nominal Parker spiral angle (25 to 45). Our results show a strong decrease in 30 kHz wave intensity with both foreshock depth and distance. There is also an asymmetry in the 30 kHz emissions from the upstream and downstream foreshocks. The ion foreshock is characterized by strong emissions in the 5.4 kHz OEFD channel which are positioned much deeper in the foreshock than expected from terrestrial observations. No activity is observed in the region where field aligned ion distributions are expected. ULF wave activity, while weaker than at Earth, shows similar behavior and may indicate the presence of similar ion distributions.

  11. Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

    International Nuclear Information System (INIS)

    Muhammad Subekti

    2009-01-01

    Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

  12. Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D. [KAERI, Daejeon (Korea, Republic of)

    2012-01-15

    This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study.

  13. Quantitative analysis technique for Xenon in PWR spent fuel by using WDS

    International Nuclear Information System (INIS)

    Kwon, H. M.; Kim, D. S.; Seo, H. S.; Ju, J. S.; Jang, J. N.; Yang, Y. S.; Park, S. D.

    2012-01-01

    This study includes three processes. First, a peak centering of the X-ray line was performed after a diffraction for Xenon La1 line was installed. Xe La1 peak was identified by a PWR spent fuel sample. Second, standard intensities of Xe was obtained by interpolation of the La1 intensities from a series of elements on each side of xenon. And then Xe intensities across the radial direction of a PWR spent fuel sample were measured by WDS-SEM. Third, the electron and X-ray depth distributions for a quantitative electron probe micro analysis were simulated by the CASINO Monte Carlo program to do matrix correction of a PWR spent fuel sample. Finally, the method and the procedure for local quantitative analysis of Xenon was developed in this study

  14. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  15. PWR secondary water chemistry study

    International Nuclear Information System (INIS)

    Pearl, W.L.; Sawochka, S.G.

    1977-02-01

    Several types of corrosion damage are currently chronic problems in PWR recirculating steam generators. One probable cause of damage is a local high concentration of an aggressive chemical even though only trace levels are present in feedwater. A wide variety of trace chemicals can find their way into feedwater, depending on the sources of condenser cooling water and the specific feedwater treatment. In February 1975, Nuclear Water and Waste Technology Corporation (NWT), was contracted to characterize secondary system water chemistry at five operating PWRs. Plants were selected to allow effects of cooling water chemistry and operating history on steam generator corrosion to be evaluated. Calvert Cliffs 1, Prairie Island 1 and 2, Surry 2, and Turkey Point 4 were monitored during the program. Results to date in the following areas are summarized: (1) plant chemistry variations during normal operation, transients, and shutdowns; (2) effects of condenser leakage on steam generator chemistry; (3) corrosion product transport during all phases of operation; (4) analytical prediction of chemistry in local areas from bulk water chemistry measurements; and (5) correlation of corrosion damage to chemistry variation

  16. Design and Development of Virtual Reactivity System for PWR

    International Nuclear Information System (INIS)

    Anwar, M. I.

    2012-01-01

    The reactivity monitoring and investigation is an important mean to ensure the safety operation of a nuclear power plant. But the reactivity of the nuclear reactor usually cannot be directly measured. It should be computed with certain estimation method. In this thesis, an effort has been made using an artificial neural network and highly fluctuating experimental data for predicting the total reactivity of the nuclear reactor based on all components of net reactivity. This virtual reactivity system is designed by taking advantage of neural network's nonlinear mapping capability. Based on analysis of the reactivity contributing factors, several neural network models are built separately for control rod, boron, poisons, fuel Doppler Effect and moderator effect. Extensive simulation and validation tests for PWR show that satisfied results have been obtained with the proposed approach. It presents a new idea to estimate the PWR's reactivity using artificial intelligence. All the design and simulation work is carried out in MATLAB and a real time programming environment is chosen for the computation and prediction of reactivity. (author)

  17. Four-fluid model of PWR degraded cores

    International Nuclear Information System (INIS)

    Dearing, J.F.

    1985-01-01

    This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

  18. Venus Express uurib Maa kurja kaksikut / ref. Triin Thalheim

    Index Scriptorium Estoniae

    2005-01-01

    9. novembril startis Baikonuri kosmodroomilt Veenusele Euroopa Kosmoseagentuuri sond Venus Express, mis peaks planeedi atmosfääri sisenema aprillis. Teadlaste sõnul peab sondi saadetav info aitama mõista naaberplaneedi kliimat ja atmosfääri ning tooma selgust, kas Maa võib kunagi Veenuse sarnaseks muutuda. Lisaks joonis: Venus Express

  19. Reassessment of planetary protection requirements for Venus missions

    Science.gov (United States)

    Szostak, J.; Riemer, R.; Smith, D.; Rummel, J.

    In 2005 the US Space Studies Board SSB was asked by NASA to reexamine the planetary protection requirements for spacecraft missions to Venus In particular the SSB was tasked to 1 Assess the surface and atmospheric environments of Venus with respect to their ability to support the survival and growth of Earth-origin microbial contamination by future spacecraft missions and 2 Provide recommendations related to planetary protection issues associated with the return to Earth of samples from Venus The task group established by the SSB to address these issues assessed the known aspects of the present-day environment of Venus and the ability of Earth organisms to survive in the physical and chemical conditions found on the planet s surface or in the clouds in the planet s atmosphere As a result of its deliberations the task group found compelling evidence against there being significant dangers of forward or reverse biological contamination as a result of contact between a spacecraft and the surface of Venus or the clouds in the atmosphere of Venus regardless of the current unknowns The task group did however conclude that Venus is a body of interest relative to the process of chemical evolution and the origin of life As a result the task group endorses NASA s current policy of subjecting missions to Venus to the requirements imposed by planetary protection Category II rather than the less restrictive Category I recommended by COSPAR

  20. Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

    International Nuclear Information System (INIS)

    Sayyaadi, Hoseyn; Sabzaligol, Tooraj

    2010-01-01

    A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

  1. Chandra Captures Venus In A Whole New Light

    Science.gov (United States)

    2001-11-01

    Scientists have captured the first X-ray view of Venus using NASA's Chandra X-ray Observatory. The observations provide new information about the atmosphere of Venus and open a new window for examining Earth's sister planet. Venus in X-rays looks similar to Venus in visible light, but there are important differences. The optically visible Venus is due to the reflection of sunlight and, for the relative positions of Venus, Earth and Sun during these observations, shows a uniform half-crescent that is brightest toward the middle. The X-ray Venus is slightly less than a half-crescent and brighter on the limbs. The differences are due to the processes by which Venus shines in visible and X-ray light. The X-rays from Venus are produced by fluorescence, rather than reflection. Solar X-rays bombard the atmosphere of Venus, knock electrons out of the inner parts of the atoms, and excite the atoms to a higher energy level. The atoms almost immediately return to their lower energy state with the emission of a fluorescent X-ray. A similar process involving ultraviolet light produces the visible light from fluorescent lamps. For Venus, most of the fluorescent X-rays come from oxygen and carbon atoms between 120 and 140 kilometers (74 to 87 miles) above the planet's surface. In contrast, the optical light is reflected from clouds at a height of 50 to 70 kilometers (31 to 43 miles). As a result, Venus' Sun-lit hemisphere appears surrounded by an almost-transparent luminous shell in X-rays. Venus looks brightest at the limb since more luminous material is there. Venus X-ray/Optical Composite of Venus Credit: Xray: NASA/CXC/MPE/K.Dennerl et al., Optical: Konrad Dennerl "This opens up the exciting possibility of using X-ray observations to study regions of the atmosphere of Venus that are difficult to investigate by other means," said Konrad Dennerl of the Max Planck Institute for Extraterrestrial Physics in Garching, Germany, leader of an international team of scientists that

  2. Structure of the middle atmosphere of Venus and future observation with PFS on Venus Express.

    Science.gov (United States)

    Zasova, L. V.; Formisano, V.; Moroz, V. I.; Ignatiev, N. I.; Khatountsev, I. A.

    Investigation of the middle atmosphere of Venus (55 -- 100 km) will allow to advance our knowledge about the most puzzling phenomena of the Venus dynamics -- its superrotation. More than 70% of all absorbed by Venus Solar energy is deposited there, results in the thermal tides generation and giving energy to support the superrotation. The importance of the tides in the middle atmosphere is manifested by the tidal character of the local time variation of the structure of the thermal field, zonal wind field (especially, behavior of the wind speed in the mid latitude jet), upper clouds, with amplitudes depending on the altitude and latitude. Investigation of the middle atmosphere is a scientific goal of the long wavelength channel of PFS on Venus Express, as well as of its short wavelength channel (the latter on the day side). The 3D temperature, aerosol, thermal wind and SO2 abundance fields, spatial distribution of abundance of H2O (possibly vertical profile), CO, HCl, HF will be obtained.

  3. ABB advanced BWR and PWR fuel

    International Nuclear Information System (INIS)

    Junkrans, S.; Helmersson, S.; Andersson, S.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

  4. Preliminary study on direct recycling of spent PWR fuel in PWR system

    International Nuclear Information System (INIS)

    Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

    2012-01-01

    Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

  5. Low-frequency magnetic field fluctuations in Venus' solar wind interaction region: Venus Express observations

    Directory of Open Access Journals (Sweden)

    L. Guicking

    2010-04-01

    Full Text Available We investigate wave properties of low-frequency magnetic field fluctuations in Venus' solar wind interaction region based on the measurements made on board the Venus Express spacecraft. The orbit geometry is very suitable to investigate the fluctuations in Venus' low-altitude magnetosheath and mid-magnetotail and provides an opportunity for a comparative study of low-frequency waves at Venus and Mars. The spatial distributions of the wave properties, in particular in the dayside and nightside magnetosheath as well as in the tail and mantle region, are similar to observations at Mars. As both planets do not have a global magnetic field, the interaction process of the solar wind with both planets is similar and leads to similar instabilities and wave structures. We focus on the spatial distribution of the wave intensity of the fluctuating magnetic field and detect an enhancement of the intensity in the dayside magnetosheath and a strong decrease towards the terminator. For a detailed investigation of the intensity distribution we adopt an analytical streamline model to describe the plasma flow around Venus. This allows displaying the evolution of the intensity along different streamlines. It is assumed that the waves are generated in the vicinity of the bow shock and are convected downstream with the turbulent magnetosheath flow. However, neither the different Mach numbers upstream and downstream of the bow shock, nor the variation of the cross sectional area and the flow velocity along the streamlines play probably an important role in order to explain the observed concentration of wave intensity in the dayside magnetosheath and the decay towards the nightside magnetosheath. But, the concept of freely evolving or decaying turbulence is in good qualitative agreement with the observations, as we observe a power law decay of the intensity along the streamlines. The observations support the assumption of wave convection through the magnetosheath, but

  6. Energy consumption analysis of the Venus Deep Space Station (DSS-13)

    Science.gov (United States)

    Hayes, N. V.

    1983-01-01

    This report continues the energy consumption analysis and verification study of the tracking stations of the Goldstone Deep Space Communications Complex, and presents an audit of the Venus Deep Space Station (DSS 13). Due to the non-continuous radioastronomy research and development operations at the station, estimations of energy usage were employed in the energy consumption simulation of both the 9-meter and 26-meter antenna buildings. A 17.9% decrease in station energy consumption was experienced over the 1979-1981 years under study. A comparison of the ECP computer simulations and the station's main watt-hour meter readings showed good agreement.

  7. Determination of the time to failure curve as a function of stress for a highly irradiated AISI 304 stainless steel after constant load tests in simulated PWR water environment

    International Nuclear Information System (INIS)

    Pokor, C.; Massoud, J.P.; Wintergerst, M.; Toivonen, A.; Ehrnsten, U.; Karlsen, W.

    2011-01-01

    The structures of Reactor Pressure Vessel Internals are subjected to an intense neutron flux. Under these operating conditions, the microstructure and the mechanical properties of the austenitic stainless steel components change. In addition, these components are subjected to stress of either manufacturing origin or generated under operation. Cases of baffle bolts cracking have occurred in CP0 Nuclear Power Plant units. The mechanism of degradation of these bolts is Irradiation-Assisted Stress Corrosion Cracking. In order to obtain a better understanding of this mechanism and its principal parameters of influence, a set of stress corrosion tests (mainly constant load tests) were launched within the framework of the EDF project 'PWR Internals' using materials from a Chooz A baffle corner (SA 304). These tests aim to quantify the influence on IASCC of the applied stress, temperature and environment (primary water, higher lithium concentration, inert environment) for an irradiation dose close to 30 dpa. A curve showing time to failure as a function of the stress was determined. The shape of this curve is consistent with the few data that are available in the literature. A stress threshold of about 50 % of the yield strength value at the test temperature has been determined, below which cracking in that environment seems impossible. After irradiation this material is sensitive to intergranular fracture in a primary environment, but also in an inert environment (argon) at 340 C. The tests also showed a negative effect of increased lithium concentration on the time to failure and on the stress threshold. (authors)

  8. Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

    International Nuclear Information System (INIS)

    Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

    1993-03-01

    As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

  9. Venus gravity - Analysis of Beta Regio

    Science.gov (United States)

    Esposito, P. B.; Sjogren, W. L.; Mottinger, N. A.; Bills, B. G.; Abbott, E.

    1982-01-01

    Radio tracking data acquired over Beta Regio were analyzed to obtain a surface mass distribution from which a detailed vertical gravity field was derived. In addition, a corresponding vertical gravity field was evaluated solely from the topography of the Beta region. A comparison of these two maps confirms the strong correlation between gravity and topography which was previously seen in line-of-sight gravity maps. It also demonstrates that the observed gravity is a significant fraction of that predicted from the topography alone. The effective depth of complete isostatic compensation for the Beta region is estimated to be 330 km, which is somewhat deeper than that found for other areas of Venus.

  10. Uvmas: Venus Ultraviolet-visual Mapping Spectrometer

    Science.gov (United States)

    Bellucci, G.; Zasova, L.; Altieri, F.; Formisano, V.; Ignatiev, N.; Moroz, V.

    We present the concept of an instrument for remote sensing of Venus from a planetary orbiter. The main characteristics of the instrument are the following: A~é· Spectral range: 0.190 A~é­ 0.490 A~éµm A~é· Spectral resolution: 0.4 nm (/= 500 at 0.2 A~éµ m) A~é· Angular resolution: 0.4 mrad at max A~é· Spatial resolution: 200 meters at 500 Km A~é· Field of view = 5.7A~é° A~é· S/N: 70 at 0.2 A~éµ m at 1 sec exp time given albedo = 0.03. The scientific objectives are the following: Dynamic investigation (0.2 5 µm). Mapping facility will allow the tracking of the UV features and will define the velocities in the atmosphere near the cloud top level. Detailed mapping of velocities of UV features at high spatial resolution, their variation with latitude, altitude and local time will advance our knowledge in understanding the puzzles of Venus dynamics like how and what mechanism drives the Venus atmospheric mass from equator to pole against temperature gradient and what is the mechanism supporting the zonal superrotation. What is the polar vortex organization, at what latitudes there is the descending branch of the Hadley cell. SO2 and SO in the range 0.232 µm. In this spectral range the SO2 and SO bands are observed. They present unresolved features with 10 Å width. Vertical profiles of these components may be obtained above the cloud and below the upper cloud boundary. Vertical, horizontal, local time and temporal variation will be obtained. This allows to create a photochemical model of the atmosphere above the clouds, and to understand a mechanism of cloud aerosol formation. "Unknown" UV- absorber, in the range 0.3 5 µm. It absorbs 50 % of the solar energy deposited on Venus. It exists only in the upper clouds. It is not known if it is in gaseous phase or included in the aerosol particles. This absorber is not homogeneously distributed and is responsible for the UV atmospheric contrast from 0.32­0.5 µm; it correlates

  11. Venus Surface Composition Constrained by Observation and Experiment

    Science.gov (United States)

    Gilmore, Martha; Treiman, Allan; Helbert, Jörn; Smrekar, Suzanne

    2017-11-01

    New observations from the Venus Express spacecraft as well as theoretical and experimental investigation of Venus analogue materials have advanced our understanding of the petrology of Venus melts and the mineralogy of rocks on the surface. The VIRTIS instrument aboard Venus Express provided a map of the southern hemisphere of Venus at ˜1 μm allowing, for the first time, the definition of surface units in terms of their 1 μm emissivity and derived mineralogy. Tessera terrain has lower emissivity than the presumably basaltic plains, consistent with a more silica-rich or felsic mineralogy. Thermodynamic modeling and experimental production of melts with Venera and Vega starting compositions predict derivative melts that range from mafic to felsic. Large volumes of felsic melts require water and may link the formation of tesserae to the presence of a Venus ocean. Low emissivity rocks may also be produced by atmosphere-surface weathering reactions unlike those seen presently. High 1 μm emissivity values correlate to stratigraphically recent flows and have been used with theoretical and experimental predictions of basalt weathering to identify regions of recent volcanism. The timescale of this volcanism is currently constrained by the weathering of magnetite (higher emissivity) in fresh basalts to hematite (lower emissivity) in Venus' oxidizing environment. Recent volcanism is corroborated by transient thermal anomalies identified by the VMC instrument aboard Venus Express. The interpretation of all emissivity data depends critically on understanding the composition of surface materials, kinetics of rock weathering and their measurement under Venus conditions. Extended theoretical studies, continued analysis of earlier spacecraft results, new atmospheric data, and measurements of mineral stability under Venus conditions have improved our understanding atmosphere-surface interactions. The calcite-wollastonite CO2 buffer has been discounted due, among other things, to

  12. Behaviour of organic iodides under pwr accident conditions

    International Nuclear Information System (INIS)

    Alm, M.

    1982-01-01

    Laboratory experiments were performed to study the behaviour of radioactive methyl iodide under PWR loss-of-coolant conditions. The pressure relief equipment consisted of an autoclave for simulating the primary circuit and of an expansion vessel for simulating the conditions after a rupture in the reactor coolant system. After pressure relief, the composition of the CH 3 sup(127/131)I-containing steam-air mixture within the expansion vessel was analysed at 80 0 C over a period of 42 days. On the basis of the values measured and of data taken from the literature, both qualitative and quantitative assessments have been made as to the behaviour of radioactive methyl iodide in the event of loss-of-coolant accidents. (author)

  13. Analysis of reactivity insertion accidents in PWR reactors

    International Nuclear Information System (INIS)

    Camargo, C.T.M.

    1978-06-01

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

  14. Venus Express set for launch to the cryptic planet

    Science.gov (United States)

    2005-10-01

    On Wednesday, 26 October 2005, the sky over the Baikonur Cosmodrome, Kazakhstan, will be illuminated by the blast from a Soyuz-Fregat rocket carrying this precious spacecraft aloft. The celestial motion of the planets in our Solar System has given Venus Express the window to travel to Venus on the best route. In fact, every nineteen months Venus reaches the point where a voyage from Earth is the most fuel-efficient. To take advantage of this opportunity, ESA has opted to launch Venus Express within the next ‘launch window’, opening on 26 October this year and closing about one month later, on 24 November. Again, due to the relative motion of Earth and Venus, plus Earth’s daily rotation, there is only one short period per day when it is possible to launch, lasting only a few seconds. The first launch opportunity is on 26 October at 06:43 Central European Summer Time (CEST) (10:43 in Baikonur). Venus Express will take only 163 days, a little more than five months, to reach Venus. Then, in April 2006, the adventure of exploration will begin with Venus finally welcoming a spacecraft, a fully European one, more than ten years after humankind paid the last visit. The journey starts at launch One of the most reliable launchers in the world, the Soyuz-Fregat rocket, will set Venus Express on course for its target. Soyuz, procured by the European/Russian Starsem company, consists of three main stages with an additional upper stage, Fregat, atop. Venus Express is attached to this upper stage. The injection of Venus Express into the interplanetary trajectory which will bring it to Venus consists of three phases. In the first nine minutes after launch, Soyuz will perform the first phase, that is an almost vertical ascent trajectory, in which it is boosted to about 190 kilometres altitude by its three stages, separating in sequence. In the second phase, the Fregat-Venus Express ‘block’, now free from the Soyuz, is injected into a circular parking orbit around Earth

  15. Existence of collisional trajectories of Mercury, Mars and Venus with the Earth.

    Science.gov (United States)

    Laskar, J; Gastineau, M

    2009-06-11

    It has been established that, owing to the proximity of a resonance with Jupiter, Mercury's eccentricity can be pumped to values large enough to allow collision with Venus within 5 Gyr (refs 1-3). This conclusion, however, was established either with averaged equations that are not appropriate near the collisions or with non-relativistic models in which the resonance effect is greatly enhanced by a decrease of the perihelion velocity of Mercury. In these previous studies, the Earth's orbit was essentially unaffected. Here we report numerical simulations of the evolution of the Solar System over 5 Gyr, including contributions from the Moon and general relativity. In a set of 2,501 orbits with initial conditions that are in agreement with our present knowledge of the parameters of the Solar System, we found, as in previous studies, that one per cent of the solutions lead to a large increase in Mercury's eccentricity-an increase large enough to allow collisions with Venus or the Sun. More surprisingly, in one of these high-eccentricity solutions, a subsequent decrease in Mercury's eccentricity induces a transfer of angular momentum from the giant planets that destabilizes all the terrestrial planets approximately 3.34 Gyr from now, with possible collisions of Mercury, Mars or Venus with the Earth.

  16. Development of MHI PWR fuel assembly with high thermal performance

    International Nuclear Information System (INIS)

    Yasushi Makino; Masaya Hoshi; Masaji Mori; Hidetoshi Kido; Kazuo Ikeda

    2005-01-01

    Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a PWR fuel assembly to meet the needs of Japanese fuel market with mainly improving its reliability such as a mechanical strength, a seismic strength and endurance. For burn-up extension of the fuel to 55 GWd/t, MHI has introduced a Zircaloy spacer grid with better neutron economics with retaining the reliability in an operating core. However, for a future power up-rating and a longer cycle operation, a higher thermal performance is required for PWR fuel assembly. To meet the needs of fuel market, MHI has developed an advanced type of Zircaloy spacer grid with a greater DNB performance while retaining the reliability of a fuel and a relatively low pressure drop. For the greater DNB performance, MHI optimized geometrical shape of mixing vane to promote a fluid mixing performance. In this report, higher DNB performance provided by the advanced Zircaloy spacer grid is presented. The results of 3D simulation for the flow behavior in 5 x 5 partial assembly, a mixing test and a water DNB test were compared between the current and the advanced spacer grids. Consequently, it was confirmed that a crossover vane enhanced a fluid mixing and the advanced spacer grid could significantly improve DNB performance compared with the current design of spacer grids. (authors)

  17. A new approach to PWR power control using intelligent techniques

    International Nuclear Information System (INIS)

    Boroushaki, M.; Ghofrani, M.B.; Lucas, C.; Yazdanpanah, M.J.; Sadati, N.

    2004-01-01

    Improved load following capability is one of the main technical performances of advanced PWR(APWR). Controlling the nuclear reactor core during load following operation encounters some difficulties. These difficulties mainly arise from nuclear reactor core limitations in local power peaking, while the core is subject to large and sharp variation of local power density during transients. Axial offset (A.O) is the parameter usually used to represent of core power peaking, in form of a practical parameter. This paper, proposes a new intelligent approach to A.o control of PWR nuclear reactors core during load following operation. This method uses a neural network model of the core to predict the dynamic behavior of the core and a fuzzy critic based on the operator knowledge and experience for the purpose of decision-making during load following operations. Simulation results show that this method can use optimum control rod groups maneuver with variable overlapping and may improve the reactor load following capability

  18. Transit of Venus Culture: A Celestial Phenomenon Intrigues the Public

    Science.gov (United States)

    Bueter, Chuck

    2012-01-01

    When Jeremiah Horrocks first observed it in 1639, the transit of Venus was a desirable telescopic target because of its scientific value. By the next transit of Venus in 1761, though, the enlightened public also embraced it as a popular celestial phenomenon. Its stature elevated over the centuries, the transit of Venus has been featured in music, poetry, stamps, plays, books, and art. The June 2004 transit emerged as a surprising global sensation, as suggested by the search queries it generated. Google's Zeitgeist deemed Venus Transit to be the #1 Most Popular Event in the world for that month. New priorities, technologies, and media have brought new audiences to the rare alignment. As the 2012 transit of Venus approaches, the trend continues with publicly accessible capabilities that did not exist only eight years prior. For example, sites from which historic observations have been made are plotted and readily available on Google Earth. A transit of Venus phone app in development will, if fully funded, facilitate a global effort to recreate historic expeditions by allowing smartphone users to submit their observed transit timings to a database for quantifying the Astronomical Unit. While maintaining relevance in modern scientific applications, the transit of Venus has emerged as a cultural attraction that briefly intrigues the mainstream public and inspires their active participation in the spectacle.

  19. PWR reactors for BBR nuclear power plants

    International Nuclear Information System (INIS)

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

  20. Reliability of PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Ribeiro, A.A.T.; Muniz, A.A.

    1978-12-01

    Results of the analysis of factors influencing the reliability of international nuclear power plants of the PWR type are presented. The reliability factor is estimated and the probability of its having lower values than a certain specified value is discussed. (Author) [pt

  1. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  2. Secondary systems of PWR and BWR

    International Nuclear Information System (INIS)

    Schindler, N.

    1981-01-01

    The secondary systems of a nuclear power plant comprises the steam, condensate and feedwater cycle, the steam plant auxiliary or ancillary systems and the cooling water systems. The presentation gives a general review about the main systems which show a high similarity of PWR and BWR plants. (orig./RW)

  3. Utilization of thorium in PWR type reactors

    International Nuclear Information System (INIS)

    Correa, F.

    1977-01-01

    Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

  4. Improvement of PWR reliability by corrosion prevention

    International Nuclear Information System (INIS)

    Takamatsu, Hiroshi

    1999-01-01

    Since first PWR in Japan started commercial operation in 1970, we have encountered the various modes of corrosion on primary and secondary side components. We have paid much efforts for resolving these corrosion problems, that is, investigating the causes of corrosion and establishing the countermeasures for these corrosion. We summarize these efforts in this article. (author)

  5. Status of developing advanced PWR in Japan

    International Nuclear Information System (INIS)

    Iida, Yotaro

    1982-01-01

    During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

  6. An evaluation of tight - pitch PWR cores

    International Nuclear Information System (INIS)

    Correa, F.

    1980-01-01

    The subtask of a project carried out at MIT (Massachusetts Institute of Technology) for DOE (Department of Energy) as part of their NASAP/INFCE - related effects involving the optimization of PWR lattices in the recycle model is summarized. (E.G.) [pt

  7. Characterizing Volcanic Eruptions on Venus: Some Realistic (?) Scenarios

    Science.gov (United States)

    Stofan, E. R.; Glaze, L. S.; Grinspoon, D. H.

    2011-01-01

    When Pioneer Venus arrived at Venus in 1978, it detected anomalously high concentrations of SO2 at the top of the troposphere, which subsequently declined over the next five years. This decline in SO2 was linked to some sort of dynamic process, possibly a volcanic eruption. Observations of SO2 variability have persisted since Pioneer Venus. More recently, scientists from the Venus Express mission announced that the SPICAV (Spectroscopy for Investigation of Characteristics of the Atmosphere of Venus) instrument had measured varying amounts of SO2 in the upper atmosphere; VIRTIS (Visible and Infrared Thermal Imaging Spectrometer) measured no similar variations in the lower atmosphere (ESA, 4 April, 2008). In addition, Fegley and Prinn stated that venusian volcanoes must replenish SO2 to the atmosphere, or it would react with calcite and disappear within 1.9 my. Fegley and Tremain suggested an eruption rate on the order of approx 1 cubic km/year to maintain atmospheric SO2; Bullock and Grinspoon posit that volcanism must have occurred within the last 20-50 my to maintain the sulfuric acid/water clouds on Venus. The abundance of volcanic deposits on Venus and the likely thermal history of the planet suggest that it is still geologically active, although at rates lower than Earth. Current estimates of resurfacing rates range from approx 0.01 cubic km/yr to approx 2 cubic km/yr. Demonstrating definitively that Venus is still volcanically active, and at what rate, would help to constrain models of evolution of the surface and interior, and help to focus future exploration of Venus.

  8. Corona Associations and Their Implications for Venus

    Science.gov (United States)

    Chapman, M.G.; Zimbelman, J.R.

    1998-01-01

    Geologic mapping principles were applied to determine genetic relations between coronae and surrounding geomorphologic features within two study areas in order to better understand venusian coronae. The study areas contain coronae in a cluster versus a contrasting chain and are (1) directly west of Phoebe Regio (quadrangle V-40; centered at latitude 15??S, longitude 250??) and (2) west of Asteria and Beta Regiones (between latitude 23??N, longitude 239?? and latitude 43??N, longitude 275??). Results of this research indicate two groups of coronae on Venus: (1) those that are older and nearly coeval with regional plains, and occur globally; and (2) those that are younger and occur between Beta, Atla, and Themis Regiones or along extensional rifts elsewhere, sometimes showing systematic age progressions. Mapping relations and Earth analogs suggest that older plains coronae may be related to a near-global resurfacing event perhaps initiated by a mantle superplume or plumes. Younger coronae of this study that show age progression may be related to (1) a tectonic junction of connecting rifts resulting from local mantle upwelling and spread of a quasi-stationary hotspot plume, and (2) localized spread of post-plains volcanism. We postulate that on Venus most of the young, post-resurfacing coronal plumes may be concentrated within an area defined by the bounds of Beta, Atla, and Themis Regiones. ?? 1998 Academic Press.

  9. The Venus flybys opportunity with BEPICOLOMBO

    Science.gov (United States)

    Mangano, Valeria; de la Fuente, Sara; Montagnon, Elsa; Benkhoff, Johannes; Zender, Joe; Orsini, Stefano

    2017-04-01

    BepiColombo is a dual spacecraft mission to Mercury to be launched in October 2018 and carried out jointly between the European Space Agency (ESA) and the Japanese Aerospace Exploration Agency (JAXA). The Mercury Planetary Orbiter (MPO) payload comprises eleven experiments and instrument suites. It will focus on a global characterization of Mercury through the investigation of its interior, surface, exosphere and magnetosphere. In addition, it will test Einstein's theory of general relativity. The second spacecraft, the Mercury Magnetosphere Orbiter (MMO), will carry five experiments or instrument suites to study the environment around the planet including the planet's exosphere and magnetosphere, and their interaction processes with the solar wind. The composite spacecraft made of MPO, MMO, a transfer module (MTM) and a sunshield (MOSIF) will be launched on an escape trajectory that will bring it into heliocentric orbit on its way to Mercury. During the cruise of 7.2 years toward the inner part of the Solar System, BepiColombo will make 1 flyby to the Earth, 2 to Venus, and 6 to Mercury. Only part of its payload will be obstructed by the sunshield and the cruise spacecraft configuration, so that the two flybys to Venus will allow operations of many instruments, like: spectrometers at many wavelengths, accelerometer, radiometer, ion and electron detectors. A scientific working group has recently formed from the BepiColombo community to identify potentially interesting scientific cases and to analyse operation timelines. Preliminary outputs will be presented and discussed.

  10. VLF emissions in the Venus foreshock - Comparison with terrestrial observations

    Science.gov (United States)

    Crawford, G. K.; Strangeway, R. J.; Russell, C. T.

    1993-01-01

    An examination is conducted of ELF/VLF emissions observed in the solar wind upstream of the Venus shock, for the 100 Hz-30 kHz range, using data from the Pioneer Venus Orbiter's electric field detector and magnetometer instruments. Detailed comparisons are made with terrestrial measurements for both the electron and ion foreshocks. The results obtained support the Crawford et al. (1990) identification of the Venus electron foreshock emissions as electron plasma oscillations, whose waves are generated in situ and act to isotropize the electron distributions.

  11. Pioneer Venus and near-earth observations of interplanetary shocks

    International Nuclear Information System (INIS)

    Mihalov, J.D.; Russell, C.T.; Knudsen, W.C.; Scarf, F.L.

    1987-01-01

    Twenty-three transient interplanetary shocks observed near earth during 1978-1982, and mostly reported in the literature, have also been identified at the Pioneer Venus Orbiter spacecraft. There seems to be a fairly consistent trend for lower shock speeds, farther from the sun. Shock normals obtained using the Pioneer Venus data correspond well with published values from near earth. By referring to the portion of the Pioneer Venus plasma data used here from locations at longitudes within 37 degree of earth, it is found that shocks are weaker at earth, compared with closer to the sun

  12. Geology of the Venus equatorial region from Pioneer Venus radar imaging

    International Nuclear Information System (INIS)

    Senske, D.A.; Head, J.W.

    1989-01-01

    The surface characteristics and morphology of the equatorial region of Venus were first described by Masursky et al. who showed this part of the planet to be characterized by two topographic provinces, rolling plains and highlands, and more recently by Schaber who described and interpreted tectonic zones in the highlands. Using Pioneer Venus (PV) radar image data (15 deg S to 45 deg N), Senske and Head examined the distribution, characteristics, and deposits of individual volcanic features in the equatorial region, and in addition classified major equatorial physiographic and tectonic units on the basis of morphology, topographic signature, and radar properties derived from the PV data. Included in this classification are: plains (undivided), inter-highland tectonic zones, tectonically segmented linear highlands, upland rises, tectonic junctions, dark halo plains, and upland plateaus. In addition to the physiographic units, features interpreted as coronae and volcanic mountains have also been mapped. The latter four of the physiographic units along with features interpreted to be coronae

  13. PWR simplified fuel element simulation using calculation trailer ANSYS CFX and PARCS including pressure drop and turbulence in the spacer; Simulacion de un elemento combustible PWR simplicificado mediante el calculo acoplado ANSYS CFX y PARCS incluyendo caida de presion y turbulencia en el espaciador

    Energy Technology Data Exchange (ETDEWEB)

    Pena-Monferrer, C.; Chiva, S.; Miro, R.; Barrachina, T.; Pellacani, F.; Macian-Juan, R.

    2012-07-01

    With the recent development of a new computational tool for calculations of nuclear reactors based on the coupling between the PARCS neutron transport code and computational fluid dynamics commercial code (CFD) ANSYS CFX opens new possibilities in the fuel element design that contributes to a better understanding and a better simulation of the processes of heat transfer and specific phenomena of fluid dynamics as the {sup c}rossflow{sup .}.

  14. 3-D full core calculations for the long-term behaviour of PWR's

    International Nuclear Information System (INIS)

    Winter, H.J.; Koebke, K.; Wagner, M.R.

    1987-01-01

    Presently, the most realistic simulation of a PWR core is by means of three-dimensional (3-D) full core calculations. Only by such 3-D representations can the large scope of axial effects be treated in an accurate and direct way, without the need to perform various auxiliary calculations. Although the computationally efficient burnup-corrected nodal expansion method (NEM-BC) is used, the computing effort for 3-D reactor calculations becomes rather high, e.g. a storage of about 320000 words is required to describe a 1300 MWe PWR. NEM-BC was introduced (1979) into KWU's package of PWR design codes because of its high accuracy and the great reduction of computing time and storage requirements in comparison to other methods. The application of NEM-BC to 3-dimensional PWR design is strongly correlated with the progress achieved in the solution of the homogenization and dehomogenization problem. By means of suitable methods (equivalence theory) the transport-theoretical information of the pinwise power and burnup distribution for the heterogeneous fuel assemblies is transferred in a consistent manner to the full core reactor solution. The new methods and the corresponding code system are explained in some detail. (orig.)

  15. The verification of PWR-fuel code for PWR in-core fuel management

    International Nuclear Information System (INIS)

    Surian Pinem; Tagor M Sembiring; Tukiran

    2015-01-01

    In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

  16. Volcano morphometry and volume scaling on Venus

    Science.gov (United States)

    Garvin, J. B.; Williams, R. S., Jr.

    1994-01-01

    A broad variety of volcanic edifices have been observed on Venus. They ranged in size from the limits of resolution of the Magellan SAR (i.e., hundreds of meters) to landforms over 500 km in basal diameter. One of the key questions pertaining to volcanism on Venus concerns the volume eruption rate or VER, which is linked to crustal productivity over time. While less than 3 percent of the surface area of Venus is manifested as discrete edifices larger than 50 km in diameter, a substantial component of the total crustal volume of the planet over the past 0.5 Ga is related to isolated volcanoes, which are certainly more easily studied than the relatively diffusely defined plains volcanic flow units. Thus, we have focused our efforts on constraining the volume productivity of major volcanic edifices larger than 100 km in basal diameter. Our approach takes advantage of the topographic data returned by Magellan, as well as our database of morphometric statistics for the 20 best known lava shields of Iceland, plus Mauna Loa of Hawaii. As part of this investigation, we have quantified the detailed morphometry of nearly 50 intermediate to large scale edifices, with particular attention to their shape systematics. We found that a set of venusian edifices which include Maat, Sapas, Tepev, Sif, Gula, a feature at 46 deg S, 215 deg E, as well as the shield-like structure at 10 deg N, 275 deg E are broadly representative of the approx. 400 volcanic landforms larger than 50 km. The cross-sectional shapes of these 7 representative edifices range from flattened cones (i.e., Sif) similar to classic terrestrial lava shields such as Mauna Loa and Skjaldbreidur, to rather dome-like structures which include Maat and Sapas. The majority of these larger volcanoes surveyed as part of our study displayed cross-sectional topographies with paraboloidal shaped, in sharp contrast with the cone-like appearance of most simple terrestrial lava shields. In order to more fully explore the

  17. Exploration of Venus' Deep Atmosphere and Surface Environment

    Science.gov (United States)

    Glaze, L. S.; Amato, M.; Garvin, J. B.; Johnson, N. M.

    2017-01-01

    Venus formed in the same part of our solar system as Earth, apparently from similar materials. Although both planets are about the same size, their differences are profound. Venus and Earth experienced vastly different evolutionary pathways resulting in unexplained differences in atmospheric composition and dynamics, as well as in geophysical processes of the planetary surfaces and interiors. Understanding when and why the evolutionary pathways of Venus and Earth diverged is key to understanding how terrestrial planets form and how their atmospheres and surfaces evolve. Measurements made in situ, within the near-surface or surface environment, are critical to addressing unanswered questions. We have made substantial progress modernizing and maturing pressure vessel technologies to enable science operations in the high temperature and pressure near-surface/surfaceenvironment of Venus.

  18. Earth-type planets (Mercury, Venus, and Mars)

    Science.gov (United States)

    Marov, M. Y.; Davydov, V. D.

    1975-01-01

    Spacecraft- and Earth-based studies on the physical nature of the planets Mercury, Venus, and Mars are reported. Charts and graphs are presented on planetary surface properties, rotational parameters, atmospheric compositions, and astronomical characteristics.

  19. Engineers are from Mars and educators are from Venus: Research ...

    African Journals Online (AJOL)

    ... are from Venus: Research supervision in engineering and educational collaboration. ... The projects usually entailed an interdisciplinary thesis that addressed an ... in chemical engineering, the work-readiness of civil engineering students, ...

  20. Solar Wind Interaction and Impact on the Venus Atmosphere

    Science.gov (United States)

    Futaana, Yoshifumi; Stenberg Wieser, Gabriella; Barabash, Stas; Luhmann, Janet G.

    2017-11-01

    Venus has intrigued planetary scientists for decades because of its huge contrasts to Earth, in spite of its nickname of "Earth's Twin". Its invisible upper atmosphere and space environment are also part of the larger story of Venus and its evolution. In 60s to 70s, several missions (Venera and Mariner series) explored Venus-solar wind interaction regions. They identified the basic structure of the near-Venus space environment, for example, existence of the bow shock, magnetotail, ionosphere, as well as the lack of the intrinsic magnetic field. A huge leap in knowledge about the solar wind interaction with Venus was made possible by the 14-year long mission, Pioneer Venus Orbiter (PVO), launched in 1978. More recently, ESA's probe, Venus Express (VEX), was inserted into orbit in 2006, operated for 8 years. Owing to its different orbit from that of PVO, VEX made unique measurements in the polar and terminator regions, and probed the near-Venus tail for the first time. The near-tail hosts dynamic processes that lead to plasma energization. These processes in turn lead to the loss of ionospheric ions to space, slowly eroding the Venusian atmosphere. VEX carried an ion spectrometer with a moderate mass-separation capability and the observed ratio of the escaping hydrogen and oxygen ions in the wake indicates the stoichiometric loss of water from Venus. The structure and dynamics of the induced magnetosphere depends on the prevailing solar wind conditions. VEX studied the response of the magnetospheric system on different time scales. A plethora of waves was identified by the magnetometer on VEX; some of them were not previously observed by PVO. Proton cyclotron waves were seen far upstream of the bow shock, mirror mode waves were observed in magnetosheath and whistler mode waves, possibly generated by lightning discharges were frequently seen. VEX also encouraged renewed numerical modeling efforts, including fluid-type of models and particle-fluid hybrid type of models

  1. The multistring model VENUS for ultrarelativistic heavy ion collisions

    International Nuclear Information System (INIS)

    Werner, K.

    1988-02-01

    The event generator VENUS is based on a multistring model for heavy ion collisions at ultrarelativistic energies. The model is a straightforward extension of a successful model for soft proton-proton scattering, the latter one being consistent with e/sup /plus//e/sup /minus// annihilation and deep inelastic lepton scattering. Comparisons of VENUS results with pA and recent AA data alow some statements about intranuclear cascading. 18 refs., 7 figs

  2. Electrochemical measurements in PWR steam generators to follow crevice chemistry

    International Nuclear Information System (INIS)

    Feron, D.

    1991-01-01

    In PWR steam generator crevices, the evolution of chemistry is important for the understanding of corrosion phenomena. Electrochemical measurements have been performed in high temperature simulated crevice environments in order to follow hideout processes and remedial actions (on-line addition of boric acid). Reported tests have been conducted with model boilers of AJAX facilities. Eccentric and concentric tube support plate crevices have been instrumented with platinum electrodes. Electrochemical measurements have been collected when model boiler was under nominal conditions (primary temperature: 335 deg C, secondary temperature: 280 deg C). They include Electrochemical Impedance Spectroscopy (EIS) and potential measurements: with EIS, sodium and boric acid hideouts have been detected and followed. Potential measurements have been performed in an attempt to measure crevice PH evolution

  3. B ampersand W PWR advanced control system algorithm development

    International Nuclear Information System (INIS)

    Winks, R.W.; Wilson, T.L.; Amick, M.

    1992-01-01

    This paper discusses algorithm development of an Advanced Control System for the B ampersand W Pressurized Water Reactor (PWR) nuclear power plant. The paper summarizes the history of the project, describes the operation of the algorithm, and presents transient results from a simulation of the plant and control system. The history discusses the steps in the development process and the roles played by the utility owners, B ampersand W Nuclear Service Company (BWNS), Oak Ridge National Laboratory (ORNL), and the Foxboro Company. The algorithm description is a brief overview of the features of the control system. The transient results show that operation of the algorithm in a normal power maneuvering mode and in a moderately large upset following a feedwater pump trip

  4. Numerical regulation of a test facility of materials for PWR

    International Nuclear Information System (INIS)

    Zauq, M.H.

    1982-02-01

    The installation aims at testing materials used in nuclear power plants; tests consists in simulations of a design basis accident (failure of a primary circuit of a PWR type reactor) for a qualification of these materials. A description of the test installation, of the thermodynamic control, and of the control system is presented. The organisation of the software is then given: description of the sequence chaining monitor, operation, list and function of the programs. The analog information processing is also presented (data transmission). A real-time microcomputer and clock are used for this work. The microprocessor is the 6800 of MOTOROLA. The microcomputer used has been built around the MC 6800; its structure is described. The data acquisition include an analog data acquisition system and a numerical data acquisition system. Laboratory and on-site tests are finally presented [fr

  5. Overview PWR-Blowdown Heat Transfer Separate-Effects Program

    International Nuclear Information System (INIS)

    White, J.D.

    1978-01-01

    The ORNL Pressurized Water Reactor Blowdown Heat Transfer Program (PWR-BDHT) is a separate-effects experimental study of thermal-hydraulic phenomena occurring during the first 20 sec of a hypothetical LOCA. Specific objectives include the determination, for a wide range of parameters, of time to CHF and the following variables for both pre- and post-CHF: heat fluxes, ΔT (temperature difference between pin surface and fluid), heat transfer coefficients, and local fluid properties. A summary of the most interesting results from the program obtained during the past year is presented. These results are in the area of: (1) RELAP verification, (2) electric pin calibration, (3) time to critical heat flux (CHF), (4) heat transfer coefficient comparisons, and (5) nuclear fuel pin simulation

  6. Simplified model of a PWR primary coolant circuit

    International Nuclear Information System (INIS)

    Souza, A.L. de; Faya, A.J.G.

    1988-01-01

    The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

  7. Measurement of gas-liquid two-phase flow around horizontal tube bundle using SF6-water. Simulating high-pressure high-temperature gas-liquid two-phase flow of PWR/SG secondary coolant side at normal pressure

    International Nuclear Information System (INIS)

    Ishikawa, Atsushi; Imai, Ryoj; Tanaka, Takahiro

    2014-01-01

    In order to improve prediction accuracy of analysis code used for design and development of industrial products, technology had been developed to create and evaluate constitutive equation incorporated in analysis code. The experimental facility for PWR/SG U tubes part was manufactured to measure local void fraction and gas-liquid interfacial velocity with forming gas-liquid upward two-phase flow simulating high-pressure high-temperature secondary coolant (water-steam) rising vertically around horizontal tube bundle. The experimental facility could reproduce flow field having gas-liquid density ratio equivalent to real system with no heating using SF6 (Sulfur Hexafluoride) gas at normal temperature and pressure less than 1 MPa, because gas-liquid density ratio, surface tension and gas-liquid viscosity ratio were important parameters to determine state of gas-liquid two-phase flow and gas-liquid density ratio was most influential. Void fraction was measured by two different methods of bi-optical probe and conductivity type probe. Test results of gas-liquid interfacial velocity vs. apparent velocity were in good agreement with existing empirical equation within 10% error, which could confirm integrity of experimental facility and appropriateness of measuring method so as to set up original constitutive equation in the future. (T. Tanaka)

  8. The Venus Emissivity Mapper - Investigating the Atmospheric Structure and Dynamics of Venus' Polar Region

    Science.gov (United States)

    Widemann, T.; Marcq, E.; Tsang, C.; Mueller, N. T.; Kappel, D.; Helbert, J.; Dyar, M. D.; Smrekar, S. E.

    2017-12-01

    Venus' climate evolution is driven by the energy balance of its global cloud layers. Venus displays the best-known case of polar vortices evolving in a fast-rotating atmosphere. Polar vortices are pervasive in the Solar System and may also be present in atmosphere-bearing exoplanets. While much progress has been made since the early suggestion that the Venus clouds are H2O-H2SO4 liquid droplets (Young 1973), several cloud parameters are still poorly constrained, particularly in the lower cloud layer and optically thicker polar regions. The average particle size is constant over most of the planet but increases toward the poles. This indicates that cloud formation processes are different at latitudes greater than 60°, possibly as a result of the different dynamical regimes that exist in the polar vortices (Carlson et al. 1993, Wilson et al. 2008, Barstow et al. 2012). Few wind measurements exist in the polar region due to unfavorable viewing geometry of currently available observations. Cloud-tracking data indicate circumpolar circulation close to solid-body rotation. E-W winds decrease to zero velocity close to the pole. N-S circulation is marginal, with extremely variable morphology and complex vorticity patterns (Sanchez-Lavega et al. 2008, Luz et al. 2011, Garate-Lopez et al. 2013). The Venus Emissivity Mapper (VEM; Helbert et al., 2016) proposed for NASA's Venus Origins Explorer (VOX) and the ESA M5/EnVision orbiters has the capability to better constrain the microphysics (vertical, horizontal, time dependence of particle size distribution, or/and composition) of the lower cloud particles in three spectral bands at 1.195, 1.310 and 1.510 μm at a spatial resolution of 10 km. Circular polar orbit geometry would provide an unprecedented study of both polar regions within the same mission. In addition, VEM's pushbroom method will allow short timescale cloud dynamics to be assessed, as well as local wind speeds, using repeated imagery at 90 minute intervals

  9. The new lattice code Paragon and its qualification for PWR core applications

    International Nuclear Information System (INIS)

    Ouisloumen, M.; Huria, H.C.; Mayhue, L.T.; Smith, R.M.; Kichty, M.J.; Matsumoto, H.; Tahara, Y.

    2003-01-01

    Paragon is a new two-dimensional transport code based on collision probability with interface current method and written entirely in Fortran 90/95. The qualification of Paragon has been completed and the results are very good. This qualification included a number of critical experiments. Comparisons to the Monte Carlo code MCNP for a wide variety of PWR assembly lattice types were also performed. In addition, Paragon-based core simulator models have been compared against PWR plant startup and operational data for a large number of plants. Some results of these calculations and also comparisons against models developed with a licensed Westinghouse lattice code, Phoenix-P, are presented. The qualification described in this paper provided the basis for the qualification of Paragon both as a validated transport code and as the nuclear data source for core simulator codes

  10. Model-based fault detection and isolation of a PWR nuclear power plant using neural networks

    International Nuclear Information System (INIS)

    Far, R.R.; Davilu, H.; Lucas, C.

    2008-01-01

    The proper and timely fault detection and isolation of industrial plant is of premier importance to guarantee the safe and reliable operation of industrial plants. The paper presents application of a neural networks-based scheme for fault detection and isolation, for the pressurizer of a PWR nuclear power plant. The scheme is constituted by 2 components: residual generation and fault isolation. The first component generates residuals via the discrepancy between measurements coming from the plant and a nominal model. The neutral network estimator is trained with healthy data collected from a full-scale simulator. For the second component detection thresholds are used to encode the residuals as bipolar vectors which represent fault patterns. These patterns are stored in an associative memory based on a recurrent neutral network. The proposed fault diagnosis tool is evaluated on-line via a full-scale simulator detected and isolate the main faults appearing in the pressurizer of a PWR. (orig.)

  11. PWR reactor vessel in-service-inspection according to RSEM

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    customers, specific techniques and tools were developed whenever deemed necessary: manipulators, probes, pusher-pullers, transducers and software products. NDT systems allow all kinds of remote-controlled acquisition and data processing. Intercontrole has performed more than 280 reactor vessel pre- and in-service inspections in France and abroad: Reactor vessel examinations in France since 1974, including more than 240 vessel pre- and in-service inspections; PSI and first ISI of Daya Bay (China) RPV. The contract for the Daya Bay first 10-year RPV ISI, to be carried out in 2005 for the unit 2 and in 2006 for the Unit 1 has been awarded; Reactor vessel examination in accordance with ASME Sections V and XI since 1981, which includes 41 reactor vessel inspections; ISI of KRSKO (Slovenia) RPV (full and partial scope), following ASME code requirements. Every year Intercontrole performs the inspection of more than 10 RPV worldwide. The paper has the following sections: Nuclear services experience; NDE capabilities and experience; Customer requirements for PWR reactor vessel in-service-inspection according to RSEM; Intercontrole solutions to comply with customer expectations; MIS manipulator; Data acquisition system; Analysis process and tools; Computer simulation. In conclusion, up to October 2005 Intercontrole successfully performed four RPV full ten-year RSEM In-Service Inspections with its new MIS. Each time the Vessel Occupation Time was lower than expected (from 10 to 20%). This proved the efficiency of the concept (manipulator + SaphirPLUS data acquisition system + Civacuve analysis software)

  12. Analysis of C/E results of fission rate ratio measurements in several fast lead VENUS-F cores

    Science.gov (United States)

    Kochetkov, Anatoly; Krása, Antonín; Baeten, Peter; Vittiglio, Guido; Wagemans, Jan; Bécares, Vicente; Bianchini, Giancarlo; Fabrizio, Valentina; Carta, Mario; Firpo, Gabriele; Fridman, Emil; Sarotto, Massimo

    2017-09-01

    During the GUINEVERE FP6 European project (2006-2011), the zero-power VENUS water-moderated reactor was modified into VENUS-F, a mock-up of a lead cooled fast spectrum system with solid components that can be operated in both critical and subcritical mode. The Fast Reactor Experiments for hybrid Applications (FREYA) FP7 project was launched in 2011 to support the designs of the MYRRHA Accelerator Driven System (ADS) and the ALFRED Lead Fast Reactor (LFR). Three VENUS-F critical core configurations, simulating the complex MYRRHA core design and one configuration devoted to the LFR ALFRED core conditions were investigated in 2015. The MYRRHA related cores simulated step by step design peculiarities like the BeO reflector and in pile sections. For all of these cores the fuel assemblies were of a simple design consisting of 30% enriched metallic uranium, lead rodlets to simulate the coolant and Al2O3 rodlets to simulate the oxide fuel. Fission rate ratios of minor actinides such as Np-237, Am-241 as well as Pu-239, Pu-240, Pu-242 and U-238 to U-235 were measured in these VENUS-F critical assemblies with small fission chambers in specially designed locations, to determine the spectral indices in the different neutron spectrum conditions. The measurements have been analyzed using advanced computational tools including deterministic and stochastic codes and different nuclear data sets like JEFF-3.1, JEFF-3.2, ENDF/B7.1 and JENDL-4.0. The analysis of the C/E discrepancies will help to improve the nuclear data in the specific energy region of fast neutron reactor spectra.

  13. Analysis of the VENUS-3 experiments

    International Nuclear Information System (INIS)

    Maerker, R.E.; D'hondt, P.; Leenders, L.; Fabry, A.

    1990-01-01

    The results of applying a hybrid superposition-synthesis calculational method to a mockup of a three-dimensional geometry involving a partial length shield assembly at the VENUS-3 facility in Mol, Belgium, are described. Comparisons of transport calculations using the method and many measurements involving nickel, indium, and aluminum dosimeters indicate agreement usually to within measurement uncertainties estimated at around 5%, if effects of inaccuracies in the dosimeter cross sections are minimized and proper orientation of the coordinate system used in the synthesis procedure is observed. These conclusions suggest a solution to the problem of predicting pressure vessel fluence in reactors modified by these partial-length shield assemblies may already exist. 7 refs., 2 figs., 1 tab

  14. Lightning measurements from the Pioneer Venus Orbiter

    Science.gov (United States)

    Scarf, F. L.; Russell, C. T.

    1983-01-01

    The plasma wave instrument on the Pioneer Venus Orbiter frequently detects strong and impulsive low-frequency signals when the spacecraft traverses the nightside ionosphere near periapsis. These particular noise bursts appear only when the local magnetic field is strong and steady and when the field is oriented to point down to the ionosphere thus; the signals have all characteristics of lightning whistlers. We have tried to identify lightning sources between the cloud layers and the planet itself by tracing rays along the B-field from the Orbiter down toward the surface. An extensive data set, consisting of measurements through Orbit 1185, strongly indicates a clustering of lightning sources near the Beta and Phoebe Regios, with an additional significant cluster near the Atla Regio at the eastern edge of Aphrodite Terra. These results suggest that there are localized lightning sources at or near the planetary surface.

  15. Venus - Volcanic features in Atla Region

    Science.gov (United States)

    1991-01-01

    This Magellan image from the Atla region of Venus shows several types of volcanic features and superimposed surface fractures. The area in the image is approximately 350 kilometers (217 miles) across, centered at 9 degrees south latitude, 199 degrees east longitude. Lava flows emanating from circular pits or linear fissures form flower-shaped patterns in several areas. A collapse depression approximately 20 kilometers by 10 kilometers (12 by 6 miles) near the center of the image is drained by a lava channel approximately 40 kilometers (25 miles) long. Numerous surface fractures and graben (linear valleys) criss-cross the volcanic deposits in north to northeast trends. The fractures are not buried by the lavas, indicating that the tectonic activity post-dates most of the volcanic activity.

  16. Venus radar mapper attitude reference quaternion

    Science.gov (United States)

    Lyons, D. T.

    1986-01-01

    Polynomial functions of time are used to specify the components of the quaternion which represents the nominal attitude of the Venus Radar mapper spacecraft during mapping. The following constraints must be satisfied in order to obtain acceptable synthetic array radar data: the nominal attitude function must have a large dynamic range, the sensor orientation must be known very accurately, the attitude reference function must use as little memory as possible, and the spacecraft must operate autonomously. Fitting polynomials to the components of the desired quaternion function is a straightforward method for providing a very dynamic nominal attitude using a minimum amount of on-board computer resources. Although the attitude from the polynomials may not be exactly the one requested by the radar designers, the polynomial coefficients are known, so they do not contribute to the attitude uncertainty. Frequent coefficient updates are not required, so the spacecraft can operate autonomously.

  17. An Encounter between the Sun and Venus

    CERN Multimedia

    2004-01-01

    The astronomical event of the year will take place on Tuesday, 8 June, when Venus transits across the disk of the sun. In the framework of CERN's 50th anniversary celebrations, the CERN Astronomy Club and the Orion Club invite you to attend their observation of the event on the car park of the Val-Thoiry shopping centre (France) between 7.15 a.m. and 1.30 p.m. Various instruments will be set up in a special tent so that the event can be observed without any risk of damage to the eyes. As the observation of this astronomical event will depend on the weather forecast, confirmation of the above arrangements will be given on the 50th anniversary website the day before.

  18. Mariner-Venus-Mercury optical navigation demonstration - Results and implications for future missions

    Science.gov (United States)

    Acton, C. H., Jr.; Ohtakay, H.

    1975-01-01

    Optical navigation uses spacecraft television pictures of a target body against a known star background in a process which relates the spacecraft trajectory to the target body. This technology was used in the Mariner-Venus-Mercury mission, with the optical data processed in near-real-time, simulating a mission critical environment. Optical data error sources were identified, and a star location error analysis was carried out. Several methods for selecting limb crossing coordinates were used, and a limb smear compensation was introduced. Omission of planetary aberration corrections was the source of large optical residuals.

  19. The Reappearance of Venus Observed 8 October 2015

    Science.gov (United States)

    Dunham, David W.; Dunham, Joan B.

    2018-01-01

    The reappearance of Venus on October 8, 2015 offered a unique opportunity to attempt observation of the ashen light of Venus as the unlit side of Venus emerged from behind the dark side of the Moon. The dark side of Venus would be offered to observers without interference from the bright side of Venus or of the Moon. Observations were made from Alice Springs, Australia visually with a 20-cm Schmidt-Cassegrain and with a low-light level surveillance camera on a 25-cm reflector. No evidence of the dark side was noted by the visual observer, the video shows little indication of Venus prior to the bright side reappearance. The conclusion reached is that the ashen light, as it was classically defined, is not observable visually or with small telescopes in the visual regime.The presentation describes the prediction, observation technique, and various analyses by the authors and others to draw conclusions from the data.To date, the authors have been unable to locate any reports of others attempting to observe this unique event. That is a pity since, not only was it interesting for an attempt to verify past observations of the ashen light, it was also a visually stunning event.

  20. Surface age of venus: use of the terrestrial cratering record

    International Nuclear Information System (INIS)

    Schaber, G.G.; Shoemaker, E.M.; Kozak, R.C.

    1987-01-01

    The average crater age of Venus' northern hemisphere may be less than 250 m.y. assuming equivalence between the recent terrestrial cratering rate and that on Venus for craters ≥ 20 km in diameter. For craters larger than this threshold size, below which crater production is significantly affected by the Venusian atmosphere, there are fairly strong observational grounds for concluding that such an equivalence in cratering rates on Venus and Earth may exist. However, given the uncertainties in the role of both active and inactive comet nuclei in the cratering history of Earth, we conclude that the age of the observed surface in the northern hemisphere of Venus could be as great as the 450-m.y. mean age of the Earth's crust. The observed surface of Venus might be even older, but no evidence from the crater observations supports an age as great as 1 b.y. If the age of the observed Venusian surface were 1 b.y., it probably should bear the impact scars of a half dozen or more large comet nuclei that penetrated the atmosphere and formed craters well over 100 km in diameter. Venera 15/16 mapped only about 25% of Venus; the remaining 75% may tell us a completely different story

  1. Estimating lithospheric properties at Atla Regio, Venus

    Science.gov (United States)

    Phillips, Roger J.

    1994-01-01

    Magellan spehrical harmonic gravity and topography models are used to estimate lithospheric properties at Alta Regio, Venus, a proposed hotspot with dynamic support from mantle plume(s). Global spherical harmonic and local representations of the gravity field share common properties in the Atla region interms of their spectral behavior over a wavelength band from approximately 2100 to approximately 700 km. The estimated free-air admittance spectrum displays a rather featureless long-wavelength portion followed by a sharp rise at wavelengths shorter than about 1000 km. This sharp rise requires significant flexural support of short-wavelength structures. The Bouguer coherence also displays a sharp drop in this wavelength band, indicating a finite flexural rigidity of the lithosphere. A simple model for lithospheric loading from above and below is introduced (D. W. Forsyth, 1985) with four parameters: f, the ratio of bottom loading to top loading; z(sub m), crustal thickness; z(sub l) depth to bottom loading source; and T(sub e) elastic lithosphere thickness. A dual-mode compensation model is introduced in which the shorter wavelengths (lambda approximately less than 1000 km) might be explained best by a predominance of top loading by the large shield volcanoes Maat Mons, Ozza Mons, and Sapas Mons, and the longer wavelengths (lambda approximately greater than 1500 km) might be explained best by a deep depth of compensation, possibly representing bottom loading by a dynamic source. A Monte Carlo inversion technique is introduced to thoroughly search out the four-space of the model parameters and to examine parameter correlation in the solutions. Venus either is a considerabe deficient in heat sources relative to Earth, or the thermal lithosphere is overthickened in response to an earlier episode of significant heat loss from the planet.

  2. Transits of Venus and Colonial India

    Science.gov (United States)

    Kochhar, Rajesh

    2012-09-01

    Astronomical expeditions during the colonial period had a political and national significance also. Measuring the earth and mapping the sky were activities worthy of powerful and power- seeking nations. Such was the sanctity of global astronomical activity that many other agendas could be hidden under it. An early astronomy-related expedition turned out to be extremely beneficial, to botany. The expedition sent by the French Government in 1735 to South America under the leadership of Charles Marie de la Condamine (1701--1774) ostensibly for the measurement of an arc of the meridian at Quito in Ecuador surreptitiously collected data that enabled Linnaeus to describe the genus cinchona in 1742. When the pair of transits of Venus occurred in 1761 and 1769, France and England were engaged in a bitter rivalry for control of India. The observation of the transits became a part of the rivalry. A telescope presented by the British to a South Indian King as a decorative toy was borrowed back for actual use. Scientifically the transit observations were a wash out, but the exercise introduced Europe to details of living Indian tradition of eclipse calculations. More significantly, it led to the institutionalization of modern astronomy in India under the auspices of the English East India Company (1787). The transits of Venus of 1874 and 1882 were important not so much for the study of the events as for initiating systematic photography of the Sun. By this, Britain owned most of the world's sunshine, and was expected to help European solar physicists get data from its vast Empire on a regular basis. This and the then genuinely held belief that a study of the sun would help predict failure of monsoons led to the institutionalization of solar physics studies in India (1899). Of course, when the solar physicists learnt that solar activity did not quite determine rainfall in India, they forgot to inform the Government.

  3. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  4. The PWR spectral code GELS. Pt. 1

    International Nuclear Information System (INIS)

    Penndorf, K.; Schult, F.; Schulz, G.

    1976-01-01

    The code procedures group constant libraries for the static PWR design of whatever fuel cycle - Uranium, Thorium, or Plutonium. The whole reach of temperatures is covered and the treatment of strong lumped absorbers as control or burnable poison pins is included. The main features are: 1) Good accuracy in spite of not fitting the material data to critical experiments; 2) speed and relatively low computer equipment; 3) restriction to PWR's only. In case of demands for higher accuracy there is a further restriction concerning the library data of the epithermal resonance absorbers: They are strictly valid only for several special lattice geometrics. Three samples are given each representing a typical application of the code. Two of them likewise are demonstrations of recalculated experiments. (orig.) [de

  5. Fuel management optimization for a PWR

    International Nuclear Information System (INIS)

    Dumas, M.; Robeau, D.

    1981-04-01

    This study is aimed to optimize the refueling pattern of a PWR. Two methods are developed, they are based on a linearized form of the optimization problem. The first method determines a feasible solution in two steps; in the first one the original problem is replaced by a relaxed one which is solved by the Method of Approximation Programming. The second step is based on the Branch and Bound method to find the feasible solution closest to the solution obtained in the first step. The second method starts from a given refueling pattern and tries to improve this pattern by the calculation of the effects of 2 by 2, 3 by 3 and 4 by 4 permutations on the objective function. Numerical results are given for a typical PWR refueling using the two methods

  6. RSK-guidelines for PWR reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The RSK guidelines for PWA reactors of April 24, 1974, have been revised and amended in this edition. The RSK presents a summary of safety requirements to be observed in the design, construction, and operation of PWR reactors in the form of guidelines. From January 1979 onwards these guidelines will be the basis of siting and safety considerations for new PWR reactors, and newly built nuclear power plants will have to form these guidelines. They are not binding for existing nuclear power plants under construction or in operation. It will be a matter of individual discussion whether or not the guidelines will be applied in these plants. The main purpose of the guidelines is to facilitate discussion among RSK members and to give early information on necessary safety requirements. If the guidelines are observed by producers and operators, the RSK will make statements on individual projects at short notice. (orig./HP) [de

  7. Optimization of reload core design for PWR

    International Nuclear Information System (INIS)

    Shen Wei; Xie Zhongsheng; Yin Banghua

    1995-01-01

    A direct efficient optimization technique has been effected for automatically optimizing the reload of PWR. The objective functions include: maximization of end-of-cycle (EOC) reactivity and maximization of average discharge burnup. The fuel loading optimization and burnable poison (BP) optimization are separated into two stages by using Haling principle. In the first stage, the optimum fuel reloading pattern without BP is determined by the linear programming method using enrichments as control variable, while in the second stage the optimum BP allocation is determined by the flexible tolerance method using the number of BP rods as control variable. A practical and efficient PWR reloading optimization program based on above theory has been encoded and successfully applied to Qinshan Nuclear Power Plant (QNP) cycle 2 reloading design

  8. Lumped-parameter modeling of PWR downcomer and pressurizer for LOCA conditions

    International Nuclear Information System (INIS)

    Rohatgi, U.S.; Saha, P.; Dubow, A.A.

    1978-01-01

    Two lumped-parameter models, one for a PWR downcomer and the other for a pressurizer, are presented. The models are based on the transient, nonhomogeneous, drift-flux description of two-phase flow, and are suitable for simulating a hypothetical LOCA condition. Effects of thermal nonequilibrium are incorporated in the downcomer model, whereas the pressurizer model can track the interfaces among various flow regimes. Semiimplicit numerical schemes are used for solution. Encouraging results have been obtained for both the models. (author)

  9. 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

  10. Generalized perturbation theory error control within PWR core-loading pattern optimization

    International Nuclear Information System (INIS)

    Imbriani, J.S.; Turinsky, P.J.; Kropaczek, D.J.

    1995-01-01

    The fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for PWR reactors. The code couples the optimization technique of simulated annealing (SA) with a generalized perturbation theory (GPT) model for evaluating core physics characteristics. To ensure the accuracy of the GPT predictions, as well as to maximize the efficient of the SA search, a GPT error control method has been developed

  11. Thermal analysis of a one-element PWR spent fuel shipping cask

    International Nuclear Information System (INIS)

    Fields, S.R.

    1979-06-01

    The transient thermal behavior of a typical one-element PWR spent fuel shipping cask, following a hypothetical accident and fire, has been simulated. The objectives of the study were to determine the transient behavior of the cask and its spent fuel primary coolant through the pressure relief system and possible fuel pin clad failure due to overheating following loss of coolant. 15 figures, 7 tables

  12. PWR fuel behavior: lessons learned from LOFT

    International Nuclear Information System (INIS)

    Russell, M.L.

    1981-01-01

    A summary of the experience with the Loss-of-Fluid Test (LOFT) fuel during loss-of-coolant experiments (LOCEs), operational and overpower transient tests and steady-state operation is presented. LOFT provides unique capabilities for obtaining pressurized water reactor (PWR) fuel behavior information because it features the representative thermal-hydraulic conditions which control fuel behavior during transient conditions and an elaborate measurement system to record the history of the fuel behavior

  13. EDF/CIDEN - ONECTRA: PWR decontamination

    International Nuclear Information System (INIS)

    Fayolle, P.; Orcel, H.; Wertz, L.

    2010-01-01

    In the context of PWR circuit renewal (expected in 2011) and their decontamination, an analysis of data coming from cartography and on site decontamination measurements as well as from premise modelling by means of the PANTHERE radioprotection code, is presented. Several French PWRs have been studied. After a presentation of code principles and operation, the authors discuss the radiological context of a workstation, and give an assessment of the annual dose associated with maintenance operations with or without decontamination

  14. EPRI PWR primary water chemistry guidelines revision

    International Nuclear Information System (INIS)

    McElrath, Joel; Fruzzetti, Keith

    2014-01-01

    EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

  15. Optimum fuel use in PWR reactors

    International Nuclear Information System (INIS)

    Neubauer, W.

    1979-07-01

    An optimization program was developed to calculate minimum-cost refuelling schedules for PWR reactors. Optimization was made over several cycles, without any constraints (equilibrium cycle). In developing the optimization program, special consideration was given to an individual treatment of every fuel element and to a sufficiently accurate calculation of all the data required for safe reactor operation. The results of the optimization program were compared with experimental values obtained at Obrigheim nuclear power plant. (orig.) [de

  16. Chemical and radiochemical specifications - PWR power plants

    International Nuclear Information System (INIS)

    Stutzmann, A.

    1997-01-01

    Published by EDF this document gives the chemical specifications of the PWR (Pressurized Water Reactor) nuclear power plants. Among the chemical parameters, some have to be respected for the safety. These parameters are listed in the STE (Technical Specifications of Exploitation). The values to respect, the analysis frequencies and the time states of possible drops are noticed in this document with the motion STE under the concerned parameter. (A.L.B.)

  17. GAIA: AREVAs New PWR fuel assembly design

    Energy Technology Data Exchange (ETDEWEB)

    Vollmert, N.; Gentet, G.; Louf, P.H.; Mindt, M.; O' Brian, J.; Peucker, J.

    2015-07-01

    GAIA is the label of a new PWR Fuel Assembly design developed by AREVA with the objective to provide its customers an advanced fuel assembly design regarding both robustness and performance. Since 2012 GAIA lead fuel assemblies are under irradiation in a Swedish reactor and since 2015 in a U.S. reactor. Visual inspections and examinations carried out so far during the outages confirmed the intended reliability, robustness and the performance enhancement of the design. (Author)

  18. Shielding design for PWR in France

    International Nuclear Information System (INIS)

    Champion, G.; Charransol; Le Dieu de Ville, A.; Nimal, J.C.; Vergnaud, T.

    1983-05-01

    Shielding calculation scheme used in France for PWR is presented here for 900 MWe and 1300 MWe plants built by EDF the French utility giving electricity. Neutron dose rate at areas accessible by personnel during the reactor operation is calculated and compared with the measurements which were carried out in 900 MWe units up to now. Measurements on the first French 1300 MWe reactor are foreseen at the end of 1983

  19. Organization patterns of PWR power plants

    International Nuclear Information System (INIS)

    Leicman, J.

    1980-01-01

    Organization patterns are shown for the St. Lucia 1, North Anna, Sequoyah, and Beaver Valley nuclear power plants, for a typical PWR power plant in the USA, for the Biblis/RWE-KWU nuclear power plants and for a four-unit nuclear power plant operated by Electricite de France as well as for the Loviisa power plant. Organization patterns are also shown for relatively independent and non-independent nuclear power plants according to IAEA recommendations. (J.P.)

  20. Sensitivity analysis of a PWR pressurizer

    International Nuclear Information System (INIS)

    Bruel, Renata Nunes

    1997-01-01

    A sensitivity analysis relative to the parameters and modelling of the physical process in a PWR pressurizer has been performed. The sensitivity analysis was developed by implementing the key parameters and theoretical model lings which generated a comprehensive matrix of influences of each changes analysed. The major influences that have been observed were the flashing phenomenon and the steam condensation on the spray drops. The present analysis is also applicable to the several theoretical and experimental areas. (author)

  1. T Plant removal of PWR Chiller Subsystem

    International Nuclear Information System (INIS)

    Dana, C.M.

    1994-01-01

    The PWR Pool Chiller System is not longer required for support of the Shippingport Blanket Fuel Assemblies Storage. The Engineering Work Plan will provide the overall coordination of the documentation and physical changes to deactivate the unneeded subsystem. The physical removal of all energy sources for the Chiller equipment will be covered under a one time work plan. The documentation changes will be covered using approved Engineering Change Notices and Procedure Change Authorizations as needed

  2. Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

    International Nuclear Information System (INIS)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-01-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  3. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  4. The Conceptual Design of Innovative Safe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

  5. Materials performance in operating PWR steam generators

    International Nuclear Information System (INIS)

    Weeks, J.R.

    1975-01-01

    The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

  6. Aerosol properties in the upper clouds of Venus from glory observations by the Venus Monitoring Camera (Venus Express mission)

    Science.gov (United States)

    Markiewicz, Wojciech J.; Petrova, Elena V.; Shalygina, Oksana S.

    2018-01-01

    From the angular positions of the glory features observed on the upper cloud deck of Venus in three VMC channels (at 0.365, 0.513, and 0.965 μm), the dominating sizes of cloud particles and their refractive indices have been retrieved, and their spatial and temporal variations have been analyzed. For this, the phase profiles of brightness were compared to the single-scattering phase functions of particles of different sizes, since diffuse multiple scattering in the clouds does not move the angular positions of the glory, which is produced by the single scattering by cloud particles, but only makes them less pronounced. We presented the measured phase profiles in two ways: they were built for individual images and for individual small regions observed in series of successive images. The analysis of the data of both types has yielded consistent results. The presently retrieved radii of cloud particle average approximately 1.0-1.2 μm (though some values reach 1.4 μm) and demonstrate a variable pattern versus latitude and local solar time (LST). The decrease of particle sizes at high latitudes (down to 0.6 μm at 60°S) earlier found from the 0.965-μm and partly 0.365-μm data has been definitely confirmed in the analysis of the data of all three channels considered. To obtain the consistent estimates of particle sizes from the UV glory maximum and minimum positions, we have to vary the effective variance of the particle sizes, while it was fixed constant in our previous studies. The twofold increase of this parameter (from 0.07 to 0.14) diminishes the estimates of particle sizes by 10-15%, while the effect on the retrieved refractive index is negligible. The obtained estimates of the refractive index are more or less uniformly distributed over the covered latitude and LST ranges, and most of them are higher than those of concentrated sulfuric acid solution. This confirms our previous result obtained only at 0.965 μm, and now we may state that the cases of a

  7. Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

    1995-03-01

    The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

  8. Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

    2015-07-01

    In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

  9. TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

    International Nuclear Information System (INIS)

    MAJI, A. K.; MARSHALL, B.

    2000-01-01

    In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

  10. Venus Interior Probe Using In-situ Power and Propulsion (VIP-INSPR), Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — We envision a novel architecture for Venus Interior Probes based on in-situ resources for power generation (VIP-INSPR). Proposed Venus probe is based on the...

  11. The Creation of a Beneficial Bioshpere from Co2 in the Clouds of Venus

    Science.gov (United States)

    Linaraki, D. L.; Oungrinis, K. A.

    2017-02-01

    This research resulted in an architectural design for a Venus colony based on multiple factors combination, such as psychology of space, predicted near-future technology, and the identified environmental conditions on Venus.

  12. Venus winds at cloud level from VIRTIS during the Venus Express mission

    Science.gov (United States)

    Hueso, Ricardo; Peralta, Javier; Sánchez-Lavega, Agustín.; Pérez-Hoyos, Santiago; Piccioni, Giuseppe; Drossart, Pierre

    2010-05-01

    The Venus Express (VEX) mission has been in orbit to Venus for almost four years now. The VIRTIS instrument onboard VEX observes Venus in two channels (visible and infrared) obtaining spectra and multi-wavelength images of the planet. Images in the ultraviolet range are used to study the upper cloud at 66 km while images in the infrared (1.74 μm) map the opacity of the lower cloud deck at 48 km. Here we present our latest results on the analysis of the global atmospheric dynamics at these cloud levels using a large selection over the full VIRTIS dataset. We will show the atmospheric zonal superrotation at these levels and the mean meridional motions. The zonal winds are very stable in the lower cloud at mid-latitudes to the tropics while it shows different signatures of variability in the upper cloud where solar tide effects are manifest in the data. While the upper clouds present a net meridional motion consistent with the upper branch of a Hadley cell the lower cloud present almost null global meridional motions at all latitudes but with particular features traveling both northwards and southwards in a turbulent manner depending on the cloud morphology on the observations. A particular important atmospheric feature is the South Polar vortex which might be influencing the structure of the zonal winds in the lower cloud at latitudes from the vortex location up to 55°S. Acknowledgements This work has been funded by the Spanish MICIIN AYA2009-10701 with FEDER support and Grupos Gobierno Vasco IT-464-07.

  13. Venus - 3D Perspective View of Sapas Mons

    Science.gov (United States)

    1992-01-01

    Sapas Mons is displayed in the center of this computer-generated three-dimensional perspective view of the surface of Venus. The viewpoint is located 527 kilometers (327 miles) northwest of Sapas Mons at an elevation of 4 kilometers (2.5 miles) above the terrain. Lava flows extend for hundreds of kilometers across the fractured plains shown in the foreground to the base of Sapas Mons. The view is to the southeast with Sapas Mons appearing at the center with Maat Mons located in the background on the horizon. Sapas Mons, a volcano 400 kilometers (248 miles) across and 1.5 kilometers (0.9 mile) high is located at approximately 8 degrees north latitude, 188 degrees east longitude, on the western edge of Atla Regio. Its peak sits at an elevation of 4.5 kilometers (2.8 miles) above the planet's mean elevation. Sapas Mons is named for a Phoenician goddess. The vertical scale in this perspective has been exaggerated 10 times. Rays cast in a computer intersect the surface to create a three-dimensional perspective view. Simulated color and a digital elevation map developed by the U.S. Geological Survey are used to enhance small-scale structure. The simulated hues are based on color images recorded by the Soviet Venera 13 and 14 spacecraft. The image was produced by the Solar System Visualization project and the Magellan Science team at the JPL Multimission Image Processing Laboratory and is a single frame from a video released at the April 22, 1992 news conference.

  14. System for stress corrosion conditions tests on PWR reactors

    International Nuclear Information System (INIS)

    Castro, Andre Cesar de Jesus

    2007-01-01

    The study of environmentally assisted cracking (EAC) involves the consideration and evaluation of the inherent compatibility between a material and the environment under conditions of either applied or residual stress. EAC is a critical problem because equipment, components and structure are subject to the influence of mechanical stress, water environment of different composition, temperature and different material history. Testing for resistance to EAC is one of the most effective ways to determine the interrelationships among this variables on the process of EAC. Up to now, several experimental techniques have been developed worldwide, which address different aspects of environmental caused damage. Constant loading of CT specimens test is a typical example of test, which is used for the estimation of parameters of stress corrosion cracking. To assess the initiation stages and kinetics of crack growth, the testing facility should allow active loading of specimens in the environment that is close to the actual operation conditions of assessed component. This paper presents a testing facility for stress corrosion cracking to be installed at CDTN, which was designed and developed at CDTN. The facility is used to carry out constant load tests under simulated PWR environment, where temperature, water pressure and chemistry are controlled, which are considered the most important factors in SCC. Also, the equipment operational conditions, its applications, and restrictions are presented. The system was developed to operate at temperature until 380 degree C and pressure until 180 bar. It consists in a autoclave stuck at a mechanical system, responsible of producing load , a water treatment station, and a data acquisition system. This testing facility allows the evaluation of cracking progress, especially at PWR reactor. (author) operational conditions. (author)

  15. PWR and BWR spent fuel assembly gamma spectra measurements

    Energy Technology Data Exchange (ETDEWEB)

    Vaccaro, S. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Tobin, S.J.; Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Hu, J. [Oak Ridge National Laboratory, Oak Ridge (United States); Schwalbach, P. [European Commission, DG Energy, Directorate EURATOM Safeguards Luxembourg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company (SKB) (Sweden); Trellue, H.; Vo, D. [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2016-10-11

    A project to research the application of nondestructive assay (NDA) to spent fuel assemblies is underway. The research team comprises the European Atomic Energy Community (EURATOM), embodied by the European Commission, DG Energy, Directorate EURATOM Safeguards; the Swedish Nuclear Fuel and Waste Management Company (SKB); two universities; and several United States national laboratories. The Next Generation of Safeguards Initiative–Spent Fuel project team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. This study focuses on spectrally resolved gamma-ray measurements performed on a diverse set of 50 assemblies [25 pressurized water reactor (PWR) assemblies and 25 boiling water reactor (BWR) assemblies]; these same 50 assemblies will be measured with neutron-based NDA instruments and a full-length calorimeter. Given that encapsulation/repository and dry storage safeguards are the primarily intended applications, the analysis focused on the dominant gamma-ray lines of {sup 137}Cs, {sup 154}Eu, and {sup 134}Cs because these isotopes will be the primary gamma-ray emitters during the time frames of interest to these applications. This study addresses the impact on the measured passive gamma-ray signals due to the following factors: burnup, initial enrichment, cooling time, assembly type (eight different PWR and six different BWR fuel designs), presence of gadolinium rods, and anomalies in operating history. To compare the measured results with theory, a limited number of ORIGEN-ARP simulations were performed.

  16. Dynamics of the accumulation process of the Earth group of planets: Formation of the reverse rotation of Venus

    Science.gov (United States)

    Koslov, N. N.; Eneyev, T. M.

    1979-01-01

    A numerical simulation of the process of formation of the terrestrial planets is carried within the framework of a new theory for the accumulation of planetary and satellite systems. The numerical simulation permitted determining the parameters of the protoplanetary disk from which Mercury, Venus and the Earth were formed as result of the evolution. The acquisition of a slow retrograde rotation for Venus was discovered during the course of the investigation, whereas Mercury and the Earth acquired direct rotation about their axes. Deviations of the semimajor axes of these three planets as well as the masses of the Earth and Venus from the true values are small as a rule (l 10%). It is shown that during the accumulation of the terrestrial planets, there existed a profound relationship between the process of formation of the orbits and masses of the planet and the process of formation of their rotation about their axes. Estimates are presented for the radii of the initial effective bodies and the time of evolution for the terrestrial accumulation zone.

  17. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    International Nuclear Information System (INIS)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C.; Palma, Daniel A.P.

    2017-01-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  18. Determination of welding parameters for execution of weld overlayer on PWR nuclear reactor nozzles

    International Nuclear Information System (INIS)

    Ribeiro, Gabriela M.; Lima, Luciana I.; Quinan, Marco A.; Schvartzman, Monica M.

    2009-01-01

    In the PWR reactors, nickel based dissimilar welds have been presented susceptibilities the stress corrosion (S C). For the mitigation the problem a deposition of weld layers on the external surface of the nozzle is an alternative, viewing to provoke the compression of the region subjected to S C. This paper presents a preliminary study on the determination of welding parameters to obtain these welding overlayers. Welding depositions were performed on a test piece welded with nickel 182 alloy, simulating the conditions of a nozzle used in a PWR nuclear power plant. The welding process was the GTAW (Gas Tungsten Arc Welding), and a nickel 52 alloy as addition material. The overlayers were performed on the base metals, carbon steel an stainless steel, changing the welding parameters and verifying the the time of each weld filet. After that, the samples were micro structurally characterized. The macro structures and the microstructures obtained through optical microscopy and Vickers microhardness are presented. The preliminary results make evident the good weld quality. However, a small weld parameters influence used in the base material microstructure (carbon steel and stainless steel). The obtained results in this study will be used as reference in the construction of a mock up which will simulate all the conditions of a pressurizer nozzle of PWR reactor

  19. Development of a thermal-hydraulic code for reflood analysis in a PWR experimental loop

    Energy Technology Data Exchange (ETDEWEB)

    Alves, Sabrina P.; Mesquita, Amir Z.; Rezende, Hugo C., E-mail: sabrinapral@gmail.com, E-mail: amir@cdtn.brm, E-mail: hcr@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    A process of fundamental importance in the event of Loss of Coolant Accident (LOCA) in Pressurized Water nuclear Reactors (PWR) is the reflood of the core or rewetting of nuclear fuels. The Nuclear Technology Development Center (CDTN) has been developing since the 70’s programs to allow Brazil to become independent in the field of reactor safety analysis. To that end, in the 80’s was designed, assembled and commissioned one Rewetting Test Facility (ITR in Portuguese). This facility aims to investigate the phenomena involved in the thermal hydraulic reflood phase of a Loss of Coolant Accident in a PWR nuclear reactor. This work aim is the analysis of physical and mathematical models governing the rewetting phenomenon, and the development a thermo-hydraulic simulation code of a representative experimental circuit of the PWR reactors core cooling channels. It was possible to elaborate and develop a code called REWET. The results obtained with REWET were compared with the experimental results of the ITR, and with the results of the Hydroflut code, that was the old program previously used. An analysis was made of the evolution of the wall temperature of the test section as well as the evolution of the front for two typical tests using the two codes calculation, and experimental results. The result simulated by REWET code for the rewetting time also came closer to the experimental results more than those calculated by Hydroflut code. (author)

  20. 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

  1. Some thermalhydraulics of closure head adapters in a 3 loops PWR

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, F.; Daubert, O.; Hecker, M. [EDF/DER/National Hydraulics Laboratory, Chatou (France)] [and others

    1995-09-01

    In 1993 a R&D action, based on numerical simulations and experiments on PWR`s upper head was initiated. This paper presents the test facility TRAVERSIN (a scale model of a 900 MW PWR adapter) and the calculations performed on the geometry of different upper head sections with the Thermalhydraulic Finite Element Code N3S used for 2D and 3D computations. The paper presents the method followed to bring the adapter and upper head study to a successful conclusion. Two complementary approaches are performed to obtain global results on complete fluid flow in the upper head and local results on the flow around the adapters of closure head. A validation test case of these experimental and numerical tools is also presented.

  2. Where should one look for traces of life on Venus?

    Science.gov (United States)

    Vidmachenko, A. P.

    2018-05-01

    Now Venus is not very similar to a suitable place for living. It surface temperature exceeds 730 K, the pressure is 90 atmospheres, the cloud layer consists of sulfur dioxide, and the fog above cloud is a solution of sulfuric acid. But about 3 billion years ago, this planet among the Earth-type planets within the Solar System was perhaps the most suitable place for the existence of some form of life there. Measurements of the ratio of hydrogen isotopes in the atmosphere also showed that the planet once had much more water, and perhaps it was enough even for the oceans. In early years on Venus was similar to the earth's climate, have a satisfactory temperature and oceans of liquid water. That is, under the above conditions with moderate temperature, sufficient heat and liquid water, Venus would be quite suitable for the emergence of certain microorganisms and for the existence of primitive life there, especially in the oceans. One way to check whether the ancient Venus was once covered by the oceans is the study of the tremolite found on Earth. It is necessary to hope to find the tremolite at some depth below the surface of Venus. Also necessary to search for some biosignals in the form of petrified remains, of possibly simple thermophilic microorganisms. We believe that such an experiment can be prepared and technically carried out during the next decades.

  3. Benchmark calculations for VENUS-2 MOX -fueled reactor dosimetry

    International Nuclear Information System (INIS)

    Kim, Jong Kung; Kim, Hong Chul; Shin, Chang Ho; Han, Chi Young; Na, Byung Chan

    2004-01-01

    As a part of a Nuclear Energy Agency (NEA) Project, it was pursued the benchmark for dosimetry calculation of the VENUS-2 MOX-fueled reactor. In this benchmark, the goal is to test the current state-of-the-art computational methods of calculating neutron flux to reactor components against the measured data of the VENUS-2 MOX-fuelled critical experiments. The measured data to be used for this benchmark are the equivalent fission fluxes which are the reaction rates divided by the U 235 fission spectrum averaged cross-section of the corresponding dosimeter. The present benchmark is, therefore, defined to calculate reaction rates and corresponding equivalent fission fluxes measured on the core-mid plane at specific positions outside the core of the VENUS-2 MOX-fuelled reactor. This is a follow-up exercise to the previously completed UO 2 -fuelled VENUS-1 two-dimensional and VENUS-3 three-dimensional exercises. The use of MOX fuel in LWRs presents different neutron characteristics and this is the main interest of the current benchmark compared to the previous ones

  4. Study of anticipated transient without scram for PWR

    International Nuclear Information System (INIS)

    Pu Jilong.

    1985-01-01

    Anticipated Transient Without Scram (ATWS) of PWR, the one of the 'Unresolved Safety Issue' with NRC, has been investigated for many years. The latest analysis done by the author considers the PWR's inherent stability and long-term performence under the condition of ATWS combined with SBLOCA and studies the sensitivity of several assumptions, which shows positive results

  5. Pushing back the boundaries of PWR fuel performance

    International Nuclear Information System (INIS)

    Sofer, G.A.; Skogen, F.B.; Brown, C.A.; Fresk, Y.U.

    1985-01-01

    In today's fiercely competitive PWR reload market utilities are benefiting from a variety of design innovations which are helping to cut fuel cycle costs and to improve fuel performance. An advanced PWR fuel design from Exxon, for example, currently under evaluation at the Ginna plant in the United States, offers higher burn-up and greater power cycling. (author)

  6. Highlights of the French program on PWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pages, J P [CEA Centre d` Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Direction des Reacteurs Nucleaires

    1997-12-01

    The presentation reviews the French programme on PWR fuel including the overall results of the year 1996 for nuclear operation; fuel management and economy; French nuclear electricity generation sites; production of nuclear generated electricity; energy availability of the 900 and 1,300 Mw PWR units; average radioactive liquid releases excluding tritium per unit; plutonium recycling experience.

  7. An economic analysis code used for PWR fuel cycle

    International Nuclear Information System (INIS)

    Liu Dingqin

    1989-01-01

    An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost

  8. Sizewell: proposed site for Britain's first PWR power station

    International Nuclear Information System (INIS)

    1980-10-01

    The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

  9. Tectonic evolution of Lavinia Planitia, Venus

    Science.gov (United States)

    Squyres, Steven W.; Frank, Sharon L.; Mcgill, George E.; Solomon, Sean C.

    1991-01-01

    High resolution radar images from the Magellan spacecraft have revealed the first details of the morphology of the Lavinia Planitia region of Venus. Lavinia is a broad lowland over 2000 km across, centered at about 45 deg S latitude, 345 deg E longitude. Herein, the tectonic evolution of Lavinia is discussed, and its possible relationship to processes operating in the planet's interior. The discussion is restricted to the region from 37.3 to 52.6 deg S latitude and from about 340 to 0 deg E longitude. One of the most interesting characteristics of Lavinia is that the entire region possesses a regional tectonic framework of striking regularity. Lavinia is also transected by a complex pattern of belts of intense tectonic deformation known as ridge belts. Despite the gross topographic similarity of all of the ridge belts in Lavinia, they exhibit two rather distinct styles of near surface deformation. One is composed of sets of broad, arch-like ridges rising above the surrounding plains. In the other type, obvious fold-like ridges are rare to absent in the radar images. Both type show evidence for small amounts of shear distributed across the belts.

  10. Chemical decontamination solutions: Effects on PWR equipment

    International Nuclear Information System (INIS)

    Pezze, C.M.; Colvin, E.R.; Aspden, R.G.

    1992-01-01

    A critical objective for the nuclear industry is the reduction of personnel exposure to radiation. Reductions have been achieved through industry's radiation management programs including training and radiation awareness concepts. Increased plant maintenance and higher radiation fields at many sites continue to raise concerns. To alleviate the radiation exposure problem, the sources of radiation which contribute to personnel exposure must be removed from the plant. A feasible was of significantly reducing these sources from a Pressurized Water Reactor (PWR) is to chemically decontaminate the entire reactor coolant system (RCS). A program was conducted to determine the technical acceptability of using certain dilute chemical solvent processes for full RCS chemical decontamination. The two processes evaluated were CAN-DEREM and LOMI. The purpose of the program was to define and complete a systematic evaluation of the major issues that need to be addressed for the successful decontamination of the entire RCS and affected portions of the auxiliary systems of a four-loop PWR system. A test program was designed to evaluate the corrosion effects of the two decontamination processes under expected plant conditions. Materials and sample configurations dictated by generic PWR components were evaluated. The testing also included many standard corrosion coupons. The test data were then used to assess the impact of chemical decontamination on the physical condition and operability of the components, equipment and mechanical systems that make up the RCS. An overview of the test program, sample configurations, data and engineering evaluations is presented. The data demonstrate that through detailed engineering evaluations of corrosion data and equipment function, the impact of full RCS chemical decontamination on plant equipment is established

  11. Technical specifications for PWR secondary water chemistry

    International Nuclear Information System (INIS)

    Weeks, J.R.; van Rooyen, D.

    1977-08-01

    The bases for establishing Technical Specifications for PWR secondary water chemistry are reviewed. Whereas extremely stringent control of secondary water needs to be maintained to prevent denting in some units, sound bases for establishing limits that will prevent stress corrosion, wastage, and denting do not exist at the present time. This area is being examined very thoroughly by industry-sponsored research programs. Based on the evidence available to date, short term control limits are suggested; establishment of these or other limits as Technical Specifications is not recommended until the results of the research programs have been obtained and evaluated

  12. Technical basis for PWR emergency plans forming

    International Nuclear Information System (INIS)

    L'Homme, A.; Manesse, D.; Gauvain, J.; Crabol, B.

    1989-10-01

    Our speech first summarizes the french approach concerning the management of severe accidents which could occur on PWR stations. Then it defines the source-term which is being used as a general support for elaborating the emergency plans devoted to the protection of the population. It describes next the consequences of this source-term on the site and in the environment, which constitute the technical bases for defining actions of utilities and concerned authorities. It gives lastly information on the present status of the different emergency plans and the complementary work undertaken to improve them [fr

  13. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  14. Industrywide survey of PWR organics. Final report

    International Nuclear Information System (INIS)

    Richards, J.E.; Byers, W.A.

    1986-07-01

    Thirteen Pressurized Water reactor (PWR) secondary cycles were sampled for organic acids, total organic carbon, and inorganic anions. The distribution and removal of organics in a makeup water treatment system were investigted at an additional plant. TOC analyses were used for the analysis of makeup water systems; anion ion chromatography and ion exclusion chromatography were used for the analysis of secondary water systems. Additional information on plant operation and water chemistry was collected in a survey. The analytical and survey data were compared and correlations made

  15. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  16. Burst protected nuclear reactor plant with PWR

    International Nuclear Information System (INIS)

    Harand, E.; Michel, E.

    1978-01-01

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

  17. PWR life time: the EDF project

    International Nuclear Information System (INIS)

    Noel, R.; Reynes, L.; Mercier, J.P.

    1987-01-01

    Operating a very large number of standardized PWR units which supply today 70% of French power generation, Electricite de France is highly interested in getting the best estimate of the safe and economical life of these plants. An extensive program of work has been undertaken in this respect. The studies have first to go through all available data on aging process, survey and maintenance of a limited number of major components. This review will lead to recommendation of complementary work in these fields. The first conclusions are that these units are able to perform a long service time, under provision of careful survey and maintenance [fr

  18. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  19. Environmental surveillance of PWR power stations

    International Nuclear Information System (INIS)

    Conti, M.

    1980-01-01

    The action of Electricite de France with respect to the environment of PWR nuclear power stations is essentially centred on prevention. Controls are carried out at two levels: - before the power station goes on stream (radioecological study), - when the power station is operational. The purpose of the controls effected on the radioactive effluents and the environment is to check that the maximum discharge rate stipulated in the corresponding orders is complied with and to ensure that there are no anomalies in the environment [fr

  20. Advancing PWR fuel to meet customer needs

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, F W

    1987-03-01

    Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

  1. Recent development in PWR zinc injection

    International Nuclear Information System (INIS)

    Ocken, H.; Fruzzetti, K.; Frattini, P.; Wood, C.J.

    2002-01-01

    Zinc injection to the reactor coolant system (RCS) of PWRs holds the promise to alleviate two key challenges facing PWR plant operators: (1) reducing degradation of coolant system materials, including nickel-base alloy tubing and lower alloy penetrations due to stress corrosion cracking, and (2) lowering shutdown dose rates. Primary water stress corrosion cracking (PWSCC) is a dominant tube failure mode at many plants. This paper summarizes recent observations from U. S. and international PWRs that have implemented zinc injection, focusing primarily on coolant chemistry and dose rate issues. It also provides a look at the future direction of EPRI-sponsored projects on this topic. (authors)

  2. Spectroscopic characterization of Venus at the single molecule level.

    Science.gov (United States)

    David, Charlotte C; Dedecker, Peter; De Cremer, Gert; Verstraeten, Natalie; Kint, Cyrielle; Michiels, Jan; Hofkens, Johan

    2012-02-01

    Venus is a recently developed, fast maturating, yellow fluorescent protein that has been used as a probe for in vivo applications. In the present work the photophysical characteristics of Venus were analyzed spectroscopically at the bulk and single molecule level. Through time-resolved single molecule measurements we found that single molecules of Venus display pronounced fluctuations in fluorescence emission, with clear fluorescence on- and off-times. These fluorescence intermittencies were found to occupy a broad range of time scales, ranging from milliseconds to several seconds. Such long off-times can complicate the analysis of single molecule counting experiments or single-molecule FRET experiments. This journal is © The Royal Society of Chemistry and Owner Societies 2012

  3. Day and night models of the Venus thermosphere

    Science.gov (United States)

    Massie, S. T.; Hunten, D. M.; Sowell, D. R.

    1983-01-01

    A model atmosphere of Venus for altitudes between 100 and 178 km is presented for the dayside and nightside. Densities of CO2, CO, O, N2, He, and O2 on the dayside, for 0800 and 1600 hours local time, are obtained by simultaneous solution of continuity equations. These equations couple ionospheric and neutral chemistry and the transport processes of molecular and eddy diffusion. Photodissociation and photoionization J coefficients are presented to facilitate the incorporation of chemistry into circulation models of the Venus atmosphere. Midnight densities of CO2 CO, O, N2, He, and N are derived from integration of the continuity equations, subject to specified fluxes. The nightside densities and fluxes are consistent with the observed airglow of NO and O2(1 Delta). The homopause of Venus is located near 133 km on both the dayside and nightside.

  4. An intelligent pedagogic tool for teaching the operators of PWR type reactors

    International Nuclear Information System (INIS)

    Cordier, B.; Guillermard, M.

    1990-01-01

    A tool was developed for assisting the instruction of the operators of a PWR type nuclear power plant. For achieving the objectives, an expert system and a simulator were combined. The main objective of the system is to improve the work of the operators in performing remedial actions in case of accident. The simulator applies two IBM PC AT3 and a MC 680 20 microprocessor. The use and the validation of the expert system are presented. The perspectives for the system, implanted on the Tricastin nuclear power plant, are analyzed [fr

  5. Decision tree based knowledge acquisition and failure diagnosis using a PWR loop vibration model

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Ding, Y.

    1993-01-01

    An analytical vibration model of the primary system of a 1300 MW PWR was used for simulating mechanical faults. Deviations in the calculated power density spectra and coherence functions are determined and classified. The decision tree technique is then used for a personal computer supported knowledge presentation and for optimizing the logical relationships between the simulated faults and the observed symptoms. The optimized decision tree forms the knowledge base and can be used to diagnose known cases as well as to include new data into the knowledge base if new faults occur. (author)

  6. Experimental modelling of core debris dispersion from the vault under a PWR pressure vessel: Part 1

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Trenberth, R.

    1987-12-01

    Modelling experiments have been done on a 1/25 scale model in Perspex of the vault under a PWR pressure vessel. Various liquids have been used to simulate molten core debris assumed to have fallen on to the vault floor from a breach at the bottom of the pressure vessel. High pressure air and helium have been used to simulate the discharge of steam and gas from the breach. The dispersion of liquid via the vault access shafts has been measured. Photographs have been taken of fluid flow patterns and velocity profiles have been obtained. The requirements for further experiments are indicated. (author)

  7. Aeolian sand transport and aeolian deposits on Venus: A review

    Science.gov (United States)

    Kreslavsly, Mikhail A.; Bondarenko, Nataliya V.

    2017-06-01

    We review the current state of knowledge about aeolian sand transport and aeolian bedforms on planet Venus. This knowledge is limited by lack of observational data. Among the four planetary bodies of the Solar System with sufficient atmospheres in contact with solid surfaces, Venus has the densest atmosphere; the conditions there are transitional between those for terrestrial subaerial and subaqueous transport. The dense atmosphere causes low saltation threshold and short characteristic saltation length, and short scale length of the incipient dunes. A few lines of evidence indicate that the typical wind speeds exceed the saltation threshold; therefore, sand transport would be pervasive, if sand capable of saltation is available. Sand production on Venus is probably much slower than on the Earth; the major terrestrial sand sinks are also absent, however, lithification of sand through sintering is expected to be effective under Venus' conditions. Active transport is not detectable with the data available. Aeolian bedforms (transverse dunes) resolved in the currently available radar images occupy a tiny area on the planet; however, indirect observations suggest that small-scale unresolved aeolian bedforms are ubiquitous. Aeolian transport is probably limited by sand lithification causing shortage of saltation-capable material. Large impact events likely cause regional short-term spikes in aeolian transport by supplying a large amount of sand-size particles, as well as disintegration and activation of older indurated sand deposits. The data available are insufficient to understand whether the global aeolian sand transport occurs or not. More robust knowledge about aeolian transport on Venus is essential for future scientific exploration of the planet, in particular, for implementation and interpretation of geochemical studies of surface materials. High-resolution orbital radar imaging with local to regional coverage and desirable interferometric capabilities is the

  8. Tidal Venuses: triggering a climate catastrophe via tidal heating.

    Science.gov (United States)

    Barnes, Rory; Mullins, Kristina; Goldblatt, Colin; Meadows, Victoria S; Kasting, James F; Heller, René

    2013-03-01

    Traditionally, stellar radiation has been the only heat source considered capable of determining global climate on long timescales. Here, we show that terrestrial exoplanets orbiting low-mass stars may be tidally heated at high-enough levels to induce a runaway greenhouse for a long-enough duration for all the hydrogen to escape. Without hydrogen, the planet no longer has water and cannot support life. We call these planets "Tidal Venuses" and the phenomenon a "tidal greenhouse." Tidal effects also circularize the orbit, which decreases tidal heating. Hence, some planets may form with large eccentricity, with its accompanying large tidal heating, and lose their water, but eventually settle into nearly circular orbits (i.e., with negligible tidal heating) in the habitable zone (HZ). However, these planets are not habitable, as past tidal heating desiccated them, and hence should not be ranked highly for detailed follow-up observations aimed at detecting biosignatures. We simulated the evolution of hypothetical planetary systems in a quasi-continuous parameter distribution and found that we could constrain the history of the system by statistical arguments. Planets orbiting stars with massesplanet orbiting a 0.3 MSun star at 0.12 AU. We found that it probably did not lose its water via tidal heating, as orbital stability is unlikely for the high eccentricities required for the tidal greenhouse. As the inner edge of the HZ is defined by the onset of a runaway or moist greenhouse powered by radiation, our results represent a fundamental revision to the HZ for noncircular orbits. In the appendices we review (a) the moist and runaway greenhouses, (b) hydrogen escape, (c) stellar mass-radius and mass-luminosity relations, (d) terrestrial planet mass-radius relations, and (e) linear tidal theories.

  9. Mars ionopause during solar minimum: A lesson from Venus

    International Nuclear Information System (INIS)

    Mahajan, K.K.; Mayr, H.G.

    1990-01-01

    The ion densities measured by the Viking landers (Hanson et al., 1977) do not show an abrupt falloff with height, giving the false impression that Mars has no ionopause. On the basis of knowledge gained from the solar wind interaction at Venus during solar minimum, they demonstrate that the observed O 2 + profile above about 160 km on Mars is a distributed photodynamical ionosphere and can produce an ionopause at around 325 km, similar to that observed on Venus during solar minimum. They conclude that the solar wind interacts directly with the Mars ionosphere, suggesting that the planet does not have an intrinsic magnetic field of any consequence

  10. Magnetic field overshoots in the Venus blow shock

    International Nuclear Information System (INIS)

    Tatrallyay, M.; Luhmann, J.G.; Russell, C.T.

    1984-01-01

    An examination of Pioneer Venus Orbiter fluxgate magnetometer data has shown that magnetic field overshoots occur not only behind quasi-perpendicular bow shocks but also behind quasi-parallel shocks. Overshoots are assocciated only with supercritical shocks. Their amplitudes increase with increasing fast Mach number. Solar wind beta has a lesser effect. The thickness of the overshoot increases with decreasing Theta-BN. The thickness of apparent overshoots detected behind 4 strong fast interplanetary shocks (M greater than M/crit) is about 3 orders of magnitude larger. Multiple crossings of the Venus bow shock were observed mainly at turbulent shocks. Their occurence is not influenced by Theta-BN. 15 references

  11. Short Large-Amplitude Magnetic Structures (SLAMS) at Venus

    Science.gov (United States)

    Collinson, G. A.; Wilson, L. B.; Sibeck, D. G.; Shane, N.; Zhang, T. L.; Moore, T. E.; Coates, A. J.; Barabash, S.

    2012-01-01

    We present the first observation of magnetic fluctuations consistent with Short Large-Amplitude Magnetic Structures (SLAMS) in the foreshock of the planet Venus. Three monolithic magnetic field spikes were observed by the Venus Express on the 11th of April 2009. The structures were approx.1.5->11s in duration, had magnetic compression ratios between approx.3->6, and exhibited elliptical polarization. These characteristics are consistent with the SLAMS observed at Earth, Jupiter, and Comet Giacobini-Zinner, and thus we hypothesize that it is possible SLAMS may be found at any celestial body with a foreshock.

  12. Stratigraphy and Observations of Nepthys Mons Quadrangle (V54), Venus

    Science.gov (United States)

    Bridges, N. T.

    2001-01-01

    Initial mapping has begun in Venus' Nepthys Mons Quadrangle (V54, 300-330 deg. E, 25-50 deg. S). Major research areas addressed are how the styles of volcanism and tectonism have changed with time, the evolution of shield volcanoes, the evolution of coronae, the characteristics of plains volcanism, and what these observations tell us about the general geologic history of Venus. Reported here is a preliminary general stratigraphy and several intriguing findings. Additional information is contained in the original extended abstract.

  13. Propagation of the trip behavior in the VENUS vertex chamber

    International Nuclear Information System (INIS)

    Ohama, Taro; Yamada, Yoshikazu.

    1995-03-01

    The high voltage system of the VENUS vertex chamber occasionally trips by a discharge somewhere among cathode electrodes during data taking. This trip behavior induces often additional trips at other electrodes such as the skin and the grid electrodes in the vertex chamber. This propagation mechanism of trips is so complicated in this system related with multi-electrodes. Although the vertex chamber is already installed inside the VENUS detector and consequently the discharge is not able to observe directly, a trial to estimate the propagation has been done using only the information which appears around the trip circuits and the power supply of the vertex chamber. (author)

  14. Krypton and xenon in the atmosphere of Venus

    Science.gov (United States)

    Donahue, T. M.; Hoffman, J. H.; Hodges, R. R., Jr.

    1981-01-01

    The paper reports a determination by the Pioneer Venus large probe neutral mass spectrometer of upper limits to the concentration of krypton and xenon along with most of their isotopes in the atmosphere of Venus. The upper limit to the krypton mixing ratio is estimated at 47 ppb, with a very conservative estimate at 69 ppb. The probable upper limit to the sum of the mixing ratios of the isotopes Xe-128, Xe-129, Xe-130, Xe-131, and Xe-132 is 40 ppb by volume, with a very conservative upper limit three times this large.

  15. Advanced high conversion PWR: preliminary analysis

    International Nuclear Information System (INIS)

    Golfier, H.; Bellanger, V.; Bergeron, A.; Dolci, F.; Gastaldi, B.; Koberl, O.; Mignot, G.; Thevenot, C.

    2007-01-01

    In this paper, physical aspects of a HCPWR (High Conversion Light Water Reactor), which is an innovative PWR fuelled with mixed oxide and having a higher conversion ratio due to a lower moderation ratio. Moderation ratios lower than unity are considered which has led to low moderation PWR fuel assembly designs. The objectives of this parametric study are to define a feasibility area with regard to the following neutronic aspects: moderation ratio, Pu loading, reactor spectrum, irradiation time, and neutronic coefficients. Important thermohydraulic parameters are the pressure drop, the critical heat flux, the maximum temperature in the fuel rod and the pumping power. The thermohydraulic analysis shows that a range of moderation ratios from 0.8 to 1.2 is technically possible. A compromise between improved fuel utilization and research and development effort has been found for the moderation ration of about 1. The parametric study shows that there are 2 ranges of interest for the moderation ratio: -) moderation ratio between 0.8 and 1.2 with reduced fissile heights (> 3 m), hexagonal arrangement fuel assembly and square arrangement fuel assembly are possible; and -) moderation between 0.6 and 0.7 with a modification of the reactor operating conditions (reduction of the primary flow and of the thermal power), the fuel rods could be arranged inside a hexagonal fuel rod assembly. (A.C.)

  16. CECP, Decommissioning Costs for PWR and BWR

    International Nuclear Information System (INIS)

    Bierschbach, M.C.

    1997-01-01

    1 - Description of program or function: The Cost Estimating Computer Program CECP, designed for use on an IBM personal computer or equivalent, was developed for estimating the cost of decommissioning boiling water reactor (BWR) and light-water reactor (PWR) power stations to the point of license termination. 2 - Method of solution: Cost estimates include component, piping, and equipment removal costs; packaging costs; decontamination costs; transportation costs; burial volume and costs; and manpower staffing costs. Using equipment and consumables costs and inventory data supplied by the user, CECP calculates unit cost factors and then combines these factors with transportation and burial cost algorithms to produce a complete report of decommissioning costs. In addition to costs, CECP also calculates person-hours, crew-hours, and exposure person-hours associated with decommissioning. 3 - Restrictions on the complexity of the problem: The program is designed for a specific waste charge structure. The waste cost data structure cannot handle intermediate waste handlers or changes in the charge rate structures. The decommissioning of a reactor can be divided into 5 periods. 200 different items for special equipment costs are possible. The maximum amount for each special equipment item is 99,999,999$. You can support data for 10 buildings, 100 components each; ESTS1071/01: There are 65 components for 28 systems available to specify the contaminated systems costs (BWR). ESTS1071/02: There are 75 components for 25 systems available to specify the contaminated systems costs (PWR)

  17. Workers doses in central European PWR NPPs

    International Nuclear Information System (INIS)

    Janzekovic, H.; Krizman, M.

    2003-01-01

    As is stated, the ISOE database which was established in 1992 forms an excellent basis for studies and comparisons of occupational exposure data between nuclear power plants. In the year 2001, 69% of all participating reactors were pressurised water reactors. The ISOE database presents workers' exposure from 213 participating pressurised reactors (PWR) from 27 countries in that year. Among these 32 PWRs belong to six Central European Countries. The analysis of the exposure of workers based on radiation protection performance indicators (collective dose, average dose etc.) in these PWRs could be related to some nuclear safety performance indicators for recent years using ISOE database. The comparison is made to ISOE world - wide data. In the six Central European Countries altogether 32 PWR operated in the year 2001.The international databases of performance indicators related to radiation protection as for example the ISOE or the UNSCEAR database can be use as an efficient tool in the management of radiation protection of workers in a nuclear facilities and regulatory bodies. The databases enable the study of performance trends and the improvement of radiation protection. (authors)

  18. The deformation of PWR fuel in a LOCA

    International Nuclear Information System (INIS)

    Mann, C.A.; Hindle, E.D.; Parsons, P.D.

    1982-04-01

    Available world-wide published data on the deformation of PWR fuel in a loss-of-coolant accident are reviewed. Adequate data exist for the oxidation of Zircaloy up to about 1500 0 C; data are increasingly sparse above this temperature and lacking above the melting point. The US NRC criteria for embrittlement are discussed and considered adequate for undeformed cladding, though they may be less so for deformed thinned material. Cladding deformation and the factors controlling it are considered in the light of data from the US, Germany, Japan and the UK. It is concluded that strains in the range 30% - 70% can be produced in experiments simulating LOCA conditions. The behaviour of cladding is strongly influenced by the spatial distribution of temperature, which is in turn dependent on heat transfer mechanisms at the surfaces of the cladding. No realistic experiment, i.e. one with a multirod array and simulated cooling, has produced deformations which would inhibit quenching. Such experiments have not, however, as yet covered the entire range of conditions which might obtain following a LOCA. (author)

  19. Can Venus magnetosheath plasma evolve into turbulence?

    Science.gov (United States)

    Dwivedi, Navin; Schmid, Daniel; Narita, Yasuhito; Volwerk, Martin; Delva, Magda; Voros, Zoltan; Zhang, Tielong

    2014-05-01

    The present work aims to understand turbulence properties in planetary magnetosheath regions to obtain physical insight on the energy transfer from the larger to smaller scales, in spirit of searching for power-law behaviors in the spectra which is an indication of the energy cascade and wave-wave interaction. We perform a statistical analysis of energy spectra using the Venus Express spacecraft data in the Venusian magnetosheath. The fluxgate magnetometer data (VEXMAG) calibrated down to 1 Hz as well as plasma data from the ion mass analyzer (ASPERA) aboard the spacecraft are used in the years 2006-2009. Ten-minute intervals in the magnetosheath are selected, which is typical time length of observations of quasi-stationary fluctuations avoiding multiple boundaries crossings. The magnetic field data are transformed into the mean-field-aligned (MFA) coordinate system with respect to the large-scale magnetic field direction and the energy spectra are evaluated using a Welch algorithm in the frequency range between 0.008 Hz and 0.5 Hz for 105 time intervals. The averaged energy spectra show a power law upto 0.3 Hz with the approximate slope of -1, which is flatter than the Kolmogorov slope, -5/3. A slight hump in the spectra is found in the compressive component near 0.3 Hz, which could possibly be realization of mirror mode in the magnetosheath. A spectral break (sudden change in slope) accompanies the spectral hump at 0.4 Hz, above which the spectral curve becomes steeper. The overall spectral shape is reminiscent of turbulence. The low-frequency part with the slope -1 is interpreted as realization of the energy containing range, while the high-frequency part with the steepening is interpreted either as the beginning of energy cascade mediated by mirror mode or as the dissipation range due to wave-particle resonance processes. The present research work is fully supported by FP7/STORM (313038).

  20. Aerobraking at Venus: A science and technology enabler

    Science.gov (United States)

    Hibbard, Kenneth; Glaze, Lori; Prince, Jill

    2012-04-01

    Venus remains one of the great unexplored planets in our solar system, with key questions remaining on the evolution of its atmosphere and climate, its volatile cycles, and the thermal and magmatic evolution of its surface. One potential approach toward answering these questions is to fly a reconnaissance mission that uses a multi-mode radar in a near-circular, low-altitude orbit of ∼400 km and 60-70° inclination. This type of mission profile results in a total mission delta-V of ∼4.4 km/s. Aerobraking could provide a significant portion, potentially up to half, of this energy transfer, thereby permitting more mass to be allocated to the spacecraft and science payload or facilitating the use of smaller, cheaper launch vehicles.Aerobraking at Venus also provides additional science benefits through the measurement of upper atmospheric density (recovered from accelerometer data) and temperature values, especially near the terminator where temperature changes are abrupt and constant pressure levels drop dramatically in altitude from day to night.Scientifically rich, Venus is also an ideal location for implementing aerobraking techniques. Its thick lower atmosphere and slow planet rotation result in relatively more predictable atmospheric densities than Mars. The upper atmosphere (aerobraking altitudes) of Venus has a density variation of 8% compared to Mars' 30% variability. In general, most aerobraking missions try to minimize the duration of the aerobraking phase to keep costs down. These short phases have limited margin to account for contingencies. It is the stable and predictive nature of Venus' atmosphere that provides safer aerobraking opportunities.The nature of aerobraking at Venus provides ideal opportunities to demonstrate aerobraking enhancements and techniques yet to be used at Mars, such as flying a temperature corridor (versus a heat-rate corridor) and using a thermal-response surface algorithm and autonomous aerobraking, shifting many daily ground