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Sample records for retrans reactivity transients

  1. RETRANS, Reactivity Transients in LWR

    International Nuclear Information System (INIS)

    Kamelander, G.

    1989-01-01

    1 - Description of program or function: RETRANS is appropriate to calculate power excursions in light water reactors initiated by reactivity insertions due to withdrawal of control elements. As in the code TWIGL, the neutron physics model is based on the time-dependent two-group neutron diffusion equations. The equation of state of the coolant is approximated by a table built into the code. RETRANS solves the heat conduction equation and calculates the heat transfer coefficient for representative fuel rods at each time-step. 2 - Method of solution: The time-dependent neutron diffusion equations are modified by an exponential transformation and solved by means of a finite difference method. There is an option accelerating the inner iterations of the difference scheme by a coarse-mesh re-balancing method. The heat balance equations of the thermo- hydraulic model are discretized and converted into a tri-diagonal system of linear equations which is solved recursively. 3 - Restrictions on the complexity of the problem: r-z-geometry, one- phase-flow

  2. RETRAN experience with BWR transients at Yankee Atomic Electric Company

    International Nuclear Information System (INIS)

    Ansari, A.A.F.; Cronin, J.T.; Slifer, B.C.

    1981-01-01

    Yankee Atomic Electric Company is actively involved in the development of licensing methods for BWR's. The computer code chosen for analyzing system response under transient conditions is RETRAN. This paper describes the RETRAN model developed for Vermont Yankee, and the results of the RETRAN checkout and qualification that has been achieved at YAEC through comparison of RETRAN predictions to the startup test results performed at the plant as part of the 100% power startup test program. In addition, abnormal operational transients typically analyzed for licensing are also presented

  3. Analysis of cofrentes abnormal plant transients with RETRAN-02 and RETRAN-03

    International Nuclear Information System (INIS)

    Mata, P.; Sedano, P.G.; Serra, J.

    1992-01-01

    In this paper the applicability and usefulness of a complete and well-qualified plant transient code and model to support in-depth evaluation of anomalous plant transients are described. The qualified best-estimate RETRAN-02 model for the Cofrentes nuclear power plant (a boiling water reactor with an uprated power of 2952 MW) has been updated for RETRAN-03 using algebraic slip and one-dimensional kinetics. This model has been used in the analysis of recent abnormal plant transients at Cofrentes, including a partial control rod insertion at 92% power, a turbine trip at 67% power with reactor vessel overfill, and reactor instabilities during startup

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

    International Nuclear Information System (INIS)

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

    1977-06-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  6. RETRAN overview

    International Nuclear Information System (INIS)

    Agee, L.J.

    1985-01-01

    The RETRAN code has become the industry standard with respect to NSSS transient analysis. The objective of this paper is to present an overview of important RETRAN-related events since the second International meeting in April of 1982. This paper is divided into three parts. The first part addresses the current status of the code with emphasis on the design review of RETRAN-02 MOD002 and the goal of RETRAN-02 in the Reactor Analysis Support Package (RASP). These activities are being undertaken to simplify the use of RETRAN for safety analysis and reload application which may be part of an NRC submittal. The second part of the paper describes significant applications of RETRAN. In the analysis section, special emphasis is placed on validation analyses which compare the code to actual plant data or experimental facilities. The third section briefly describes the pre-release version of RETRAN and the developmental goals for the next version of RETRAN. One major limitation of all state-of-the-art thermal-hydraulic codes is the determination of the structure of the fluid. A brief description of research needs in this are indicated

  7. RETRAN-02: a program for transient thermal-hydraulic analysis of complex fluid-flow systems. Volume 4. Applications

    International Nuclear Information System (INIS)

    Peterson, C.E.; Gose, G.C.; McFadden, J.H.

    1983-01-01

    RETRAN-02 represents a significant achievement in the development of a versatile and reliable computer program for use in best estimate transient thermal-hydraulic analysis of light water reactor systems. The RETRAN-02 computer program is an extension of the RETRAN-01 program designed to provide analysis capabilities for 1) BWR and PWR transients, 2) small break loss of coolant accidents, 3) balance of plant modeling, and 4) anticipated transients without scram, while maintaining the analysis capabilities of the predecessor code. The RETRAN-02 computer code is constructed in a semimodular and dynamic dimensioned form where additions to the code can be easily carried out as new and improved models are developed. This report (the fourth of a five volume computer code manual) describes the verification and validation of RETRAN-02

  8. Peach Bottom Turbine Trip Simulations with RETRAN Using INER/TPC BWR Transient Analysis Method

    International Nuclear Information System (INIS)

    Kao Lainsu; Chiang, Show-Chyuan

    2005-01-01

    The work described in this paper is benchmark calculations of pressurization transient turbine trip tests performed at the Peach Bottom boiling water reactor (BWR). It is part of an overall effort in providing qualification basis for the INER/TPC BWR transient analysis method developed for the Kuosheng and Chinshan plants. The method primarily utilizes an advanced system thermal hydraulics code, RETRAN02/MOD5, for transient safety analyses. Since pressurization transients would result in a strong coupling effect between core neutronic and system thermal hydraulics responses, the INER/TPC method employs the one-dimensional kinetic model in RETRAN with a cross-section data library generated by the Studsvik-CMS code package for the transient calculations. The Peach Bottom Turbine Trip (PBTT) tests, including TT1, TT2, and TT3, have been successfully performed in the plant and assigned as standards commonly for licensing method qualifications for years. It is an essential requirement for licensing purposes to verify integral capabilities and accuracies of the codes and models of the INER/TPC method in simulating such pressurization transients. Specific Peach Bottom plant models, including both neutronics and thermal hydraulics, are developed using modeling approaches and experiences generally adopted in the INER/TPC method. Important model assumptions in RETRAN for the PBTT test simulations are described in this paper. Simulation calculations are performed with best-estimated initial and boundary conditions obtained from plant test measurements. The calculation results presented in this paper demonstrate that the INER/TPC method is capable of calculating accurately the core and system transient behaviors of the tests. Excellent agreement, both in trends and magnitudes between the RETRAN calculation results and the PBTT measurements, shows reliable qualifications of the codes/users/models involved in the method. The RETRAN calculated peak neutron fluxes of the PBTT

  9. Kuosheng BWR/6 recirculation pump trip transient analysis with the RETRAN02/MOD5 code

    International Nuclear Information System (INIS)

    Wang, J.R.; Shih, C.

    1992-01-01

    A recirculation pump trip (RPT) event results in a reduction in recirculation flow, which reduces the core coolant flow rate. A reduction in core flow results in an increase in core void fraction and hence a decrease in core power due to negative void reactivity feedback. Although this category of events is less severe than others and generally considered as nonlimiting, core instability still may occur such as that at LaSalle on March 9, 1988. This paper focuses on the RPT transient analysis of Kuosheng Nuclear Power Plant (KNPP), which has two units of General Electric-designed boiling water reactor (BWR)/6 with rated core thermal power of 2894 MW and rated core flow of 10645 kg/s (23472 lb m /s). The approach to investigating the RPT transient of KNPP consists of two steps. The first step is to develop a plant-specific model using the RETRAN02/MOD5 code. In this step, various plant-specific information, including design documentation, drawings, safety analysis reports, and other information supplied by vendors were collected for model development. The RPT startup test at 68% power was used for system model benchmarking to ensure the adequacy of this model and identify several sensitive parameters. The second step is to assess whether similar power oscillation phenomena may occur at KNPP because of an RPT with isolated feedwater heater event. Two transient analyses (with or without reactor scram) of the KNPP RPT with isolated feedwater heater were investigated

  10. Improvement of RETRAN-MASTER-TORC transient capability and coupling optimization

    International Nuclear Information System (INIS)

    Cho, J. Y.; Song, J. S.; Joo, H. G.; Seo, K. W.; Whang, D. H.; Lee, C. C.; Zee, S. Q.

    2003-11-01

    This work is to improve MASTER-TORC transient calculation capability by complementing the previously developed consolidated code system RETRAN- MASTER-TORC, and to reduce the computing time by coupling optimization. The coupling soundness and optimization performance of the consolidated code system are evaluated by solving a YGN3 control bank ejection accident and the OECD Main Steam Line Break(MSLB) benchmark problems. The YGN3 control bank ejection accident is analyzed by the MASTER-TORC system. Most of all results including the transient core power, peak power and time are similar with those from the MASTER-COBRA system. In the computing time, the MASTER- TORC system is proved to be same as the MASTER-COBRA system, which means the coupling is sound and well-optimized. In the analysis of the OECD MSLB benchmark problem, the RETRAN-MASTER-TORC system shows the very similar results with the RETRAN-MASTER-COBRA system. However, minor differences due to fuel conductivity and thermal capacity model are noticed. In TORC, these parameters are treated as constants, while they are modeled as temperature dependent functions in COBRA. Therefore, in the future, TORC need to complement the temperature dependent thermal properties for accurate fuel and cladding temperature calculation. In the computing time for this problem, RETRAN-MASTER-TORC system shows a little bit faster than COBRA case

  11. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  12. The limiting events transient analysis by RETRAN02 and VIPRE01 for an ABWR

    International Nuclear Information System (INIS)

    Tsai Chiungwen; Shih Chunkuan; Wang Jongrong; Lin Haotzu; Jin Jiunan; Cheng Suchin

    2009-01-01

    This paper describes the transient analysis of generator load rejection (LR) and One Turbine Control Valve Closure (OTCVC) events for Lungmen nuclear power plant (LMNPP). According to the Critical Power Ratio (CPR) criterion, the Preliminary Safety Analysis Report (PSAR) concluded that LR and OTCVC are the first and second limiting events respectively. In addition, the fuel type is changed from GE12 to GE14 now. It's necessary to re-analyze these two events for safety consideration. In this study, to quantify the impact to reactor, the difference of initial critical power ratio (ICPR) and minimum critical power ratio (MCPR), ie. ΔCPR is calculated. The ΔCPRs of the LR and OTCVC events are calculated with the combination of RETRAN02 and VIPRE01 codes. In RETRAN02 calculation, a thermal-hydraulic model was prepared for the transient analysis. The data including upper plenum pressure, core inlet flow, normalized power, and axial power shapes during transient are furthermore submitted into VIPRE01 for ΔCPR calculation. In VIPRE01 calculation, there was a hot channel model built to simulate the hottest fuel bundle. Based on the thermal-hydraulic data from RETRAN02, the ΔCPRs are calculated by VIPRE01 hot channel model. Additionally, the different TCV control modes are considered to study the influence of different TCV closure curves on transient behavior. Meanwhile, sensitivity studies including different initial system pressure and different initial power/flow conditions are also considered. Based on this analysis, the maximum ΔCPRs for LR and OTCVC are 0.162 and 0.191 respectively. According CPR criterion, the result shows that the impact caused by OTCVC event leads to be larger than LR event. (author)

  13. RETRAN nonequilibrium two-phase flow model for operational transient analyses

    International Nuclear Information System (INIS)

    Paulsen, M.P.; Hughes, E.D.

    1982-01-01

    The field balance equations, flow-field models, and equation of state for a nonequilibrium two-phase flow model for RETRAN are given. The differential field balance model equations are: (1) conservation of mixture mass; (2) conservation of vapor mass; (3) balance of mixture momentum; (4) a dynamic-slip model for the velocity difference; and (5) conservation of mixture energy. The equation of state is formulated such that the liquid phase may be subcooled, saturated, or superheated. The vapor phase is constrained to be at the saturation state. The dynamic-slip model includes wall-to-phase and interphase momentum exchanges. A mechanistic vapor generation model is used to describe vapor production under bulk subcooling conditions. The speed of sound for the mixture under nonequilibrium conditions is obtained from the equation of state formulation. The steady-state and transient solution methods are described

  14. Interface of RETRAN/MASTER Code System for APR1400

    International Nuclear Information System (INIS)

    Ku, Keuk Jong; Kang, Sang Hee; Kim, Han Gon

    2008-01-01

    MASTER(Multi-purpose Analyzer for Static and Transient Effects of Reactors), which was developed by KAERI, is a nuclear analysis and design code which can simulate the pressurized water reactor core or boiling water reactor core in 3-dimensional geometry. RETRAN is a best-estimate code for transient analysis of Non-LOCA. RETRAN code generates neutron number density in core using point kinetics model which includes feedback reactivities and converts the neutron number density into reactor power. It is conventional that RETRAN code for power generation is roughly to extrapolate feedback reactivities which are provided by MASTER code only one time before transient analysis. The purpose of this paper is to interface RETRAN code with MASTER code by real-time processing and to supply adequate feedback reactivities to RETRAN code. So, we develop interface code called MATRAN for real-time feedback reactivity processing. And for the application of MATRAN code, we compare the results of real-time MATRAN code with those of conventional RETRAN/MASTER code

  15. Applications of RETRAN-3D for nuclear power plant transient analyses

    International Nuclear Information System (INIS)

    Paulsen, M.P.; Gose, G.C.; McFadden, J.H.; Agee, L.J.

    1996-01-01

    The RETRAN-3D computer program has been developed to analyze reactor events for which nonequilibrium thermodynamics, multidimensional neutron kinetics, or the presence of noncondensable gases are important items for consideration. This paper summarizes the features of RETRAN-3D and the analyses that have been performed to provide the verification and validation of the program

  16. The RETRAN-03 computer code

    International Nuclear Information System (INIS)

    Paulsen, M.P.; McFadden, J.H.; Peterson, C.E.; McClure, J.A.; Gose, G.C.; Jensen, P.J.

    1991-01-01

    The RETRAN-03 code development effort is designed to overcome the major theoretical and practical limitations associated with the RETRAN-02 computer code. The major objectives of the development program are to extend the range of analyses that can be performed with RETRAN, to make the code more dependable and faster running, and to have a more transportable code. The first two objectives are accomplished by developing new models and adding other models to the RETRAN-02 base code. The major model additions for RETRAN-03 are as follows: implicit solution methods for the steady-state and transient forms of the field equations; additional options for the velocity difference equation; a new steady-state initialization option for computer low-power steam generator initial conditions; models for nonequilibrium thermodynamic conditions; and several special-purpose models. The source code and the environmental library for RETRAN-03 are written in standard FORTRAN 77, which allows the last objective to be fulfilled. Some models in RETRAN-02 have been deleted in RETRAN-03. In this paper the changes between RETRAN-02 and RETRAN-03 are reviewed

  17. Application of a qualified RETRAN model to plant transient evaluation support

    International Nuclear Information System (INIS)

    Sedano, P.G.; Mata, P.; Alcantud, F.; Serra, J.; Castrillo, F.

    1989-01-01

    This paper presents the applicability and usefulness of a complete and well qualified plant transient code and model to support in depth evaluation of anomalous plant transients. Analyses of several operational and abnormal transients that ocurred during the first three cycles of Cofrentes (BWR-6) NPP are presented. This application demonstrated the need of a very detailed and adjusted simulation of the control systems as well as the convenience of having as complete as possible data adquisition system. (orig.)

  18. Application of a qualified RETRAN model to plant transient evaluation support

    International Nuclear Information System (INIS)

    Sedano, P.G.; Mata, P.; Alcantud, F.; Serra, J.

    1989-01-01

    This paper presents the applicability and usefulness of a complete and well qualified plant transient code and model to support in depth evaluation of anomalous plant transients. Analyses of several operational and abnormal transients occurred during the first three cycles of Cofrentes (BWR-6) NPP are presented. This application remarked the need of a very detailed and adjusted simulation of the control systems as well as the convenience of having an as complete as possible data acquisition system

  19. One-dimensional reactor kinetics model for RETRAN

    International Nuclear Information System (INIS)

    Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

    1981-01-01

    Previous versions of RETRAN have had only a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. The principal assumption in deriving the point kinetics model is that the neutron flux may be separated into a time-dependent amplitude funtion and a time-independent shape function. Certain types of transients cannot be correctly analyzed under this assumption, since proper definitions for core average quantities such as reactivity or lifetime include the inner product of the adjoint flux with the perturbed flux. A one-dimensional neutronics model has been included in a preliminary version of RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects. This paper describes the neutronics model and discusses some of the analyses

  20. One-dimensional reactor kinetics model for RETRAN

    International Nuclear Information System (INIS)

    Gose, G.C.; Peterson, C.E.; Ellis, N.L.; McClure, J.A.

    1981-01-01

    This paper describes a one-dimensional spatial neutron kinetics model that was developed for the RETRAN code. The RETRAN -01 code has a point kinetics model to describe the reactor core behavior during thermal-hydraulic transients. A one-dimensional neutronics model has been developed for RETRAN-02. The ability to account for flux shape changes will permit an improved representation of the thermal and hydraulic feedback effects for many operational transients. 19 refs

  1. Physical models and numerical methods of the reactor dynamic computer program RETRAN

    International Nuclear Information System (INIS)

    Kamelander, G.; Woloch, F.; Sdouz, G.; Koinig, H.

    1984-03-01

    This report describes the physical models and the numerical methods of the reactor dynamic code RETRAN simulating reactivity transients in Light-Water-Reactors. The neutron-physical part of RETRAN bases on the two-group-diffusion equations which are solved by discretization similar to the TWIGL-method. An exponential transformation is applied and the inner iterations are accelerated by a coarse-mesh-rebalancing procedure. The thermo-hydraulic model approximates the equation of state by a built-in steam-water-table and disposes of options for the calculation of heat-conduction coefficients and heat transfer coefficients. (Author) [de

  2. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    International Nuclear Information System (INIS)

    Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun

    2014-01-01

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available

  3. Anticipated Transient Without SCRAM(ATWS) analysis using the RETRAN code

    Energy Technology Data Exchange (ETDEWEB)

    Youn, Bum soo; Lee, Jong beom; Song, Dong soo; Ha, Sang jun [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this study is to evaluate the Anticipated Transient Without Scram(ATWS) Loss of Load(LOL) and Loss of Normal Feedwater(LOFW) events for the OPR1000 reactor. The analysis calculates the peak RCS and secondary system pressure for the LOL and LOFW ATWS events. The main product of this study is the ATWS evaluation of the OPR1000 reactor LOL and LOFW events. The results include a sequence of events and plots of key output parameters.. This study includes results of Loss of Load and Loss of Feedwater ATWS. The LOL case results in a faster reactor trip than the LOFW since the LOFW does not have the turbine trip at time zero. In addition the LOFW event has the SBCS available and as secondary pressure increase, the steam releases from the SBCS valves provide extra cooling to the secondary system, which also cools the primary system. This additional cooling also delays the DSS trip. For the LOFW event, both the turbine and SBCS are providing additional cooling, hence the primary and secondary system heatups are slower and lower. Thus the RCS and steam generator pressure are higher for the LOL event than the LOFW event. The LOL also has a slower decrease in SG water level than the LOFW event. This is due to loss of condenser vacuum that trips and isolates the turbine and renders the SBCS unavailable for the LOL event. Hence the secondary cooling for the LOL event is due to the steam releases from the MSSVs; whereas the LOFW turbine remains online until a DTT occurs on the DSS. Also the SBCS is available because the condenser is available.

  4. RETRAN's role in the development of Northeast Utilities' analytical capabilities

    International Nuclear Information System (INIS)

    Bonaca, M.V.; Gharakhani, A.; Sterner, R.W.

    1983-01-01

    The RETRAN code plays an important role in Northeast Utilities' (NUSCO's) safety analysis capabilities. Its ability to predict plant response to operational transients has been shown by several benchmarking efforts. Additionally, it has been utilized in performing both design-basis and bestestimate analyses for NUSCO's operating plants. The decision to use RETRAN as a major contributor to reload analysis efforts at NUSCO was justified by the results of numerous analyses proving RETRAN's capabilities and credibility

  5. Analysis of a 12-Finger Rod Drop using RETRAN/MASTER Code System for APR1400

    International Nuclear Information System (INIS)

    Yu, Keuk Jong; You, Sung Chang; Kim, Han Gon

    2009-01-01

    The Optimized Power Reactor 1000 (OPR1000) has 4-finger and 12-finger Control Element Assemblies (CEAs). When the 12-finger CEA is dropped, Core Protection Calculator System (CPCS) shuts down the reactor to prevent fuel damage that could occur from the sudden reactor power peaking. By contrast, the improved CPCS of Advanced Power Reactor 1400 (APR1400), which has systems similar to those of the OPR1000, decreases reactor power rapidly using its Reactor Power Cutback System (RPCS) to avoid unwanted reactor trips caused by the CPCS during a 12- finger CEA drop event. RETRAN is a best-estimate code for transient analysis of Non-LOCA. The RETRAN control logic, which includes the function of reducing reactor power during a 12-Finger CEA drop, has been developed for the APR1400. A MATRAN program has also been developed. MATRAN is the interface program for realtime processing to connect RETRAN with MASTER code which is a nuclear analysis and design code. MATRAN supplies adequate feedback reactivities from the MASTER code to RETRAN code. The purpose of this study is to analyze the behavior of a nuclear reactor core and its primary system using conventional RETRAN analysis procedure and MATRAN program analysis procedure during a 12- finger CEA drop. In addition, the axial power distribution and Axial Shape Index (ASI) are produced by the MATRAN program and they are confirmed as within operation limits

  6. LMR steam generator blowdown with RETRAN

    International Nuclear Information System (INIS)

    Wei, T.Y.C.

    1985-01-01

    One of the transients being considered in the FSAR Chapter 15 analyses of anticipated LMR transients is the fast blowdown of a steam generator upon inadvertent actuation of the liquid metal/water reaction mitigation system. For the blowdown analysis, a stand-alone steam generator model for the IFR plant was constructed using RETRAN

  7. RETRAN-02 one-dimensional kinetics model: a review

    International Nuclear Information System (INIS)

    Gose, G.C.; McClure, J.A.

    1986-01-01

    RETRAN-02 is a modular code system that has been designed for one-dimensional, transient thermal-hydraulics analysis. In RETRAN-02, core power behavior may be treated using a one-dimensional reactor kinetics model. This model allows the user to investigate the interaction of time- and space-dependent effects in the reactor core on overall system behavior for specific LWR operational transients. The purpose of this paper is to review the recent analysis and development activities related to the one dimensional kinetics model in RETRAN-02

  8. RETRAN02/MOD02: an outside perspective

    International Nuclear Information System (INIS)

    Wei, T.Y.C.

    1984-03-01

    ANL recently participated in a review of the RETRAN02/MOD02 code to determine the range of accuracy, the reliability and the reproducibility of results obtained with the code for Chapter 15 non-LOCA system transients for both pressurized water reactors (PWRs) and boiling water reactors (BWRs). This paper summarizes the technical aspects of that review

  9. RETRAN dynamic slip model

    International Nuclear Information System (INIS)

    McFadden, J.H.; Paulsen, M.P.; Gose, G.C.

    1981-01-01

    A time dependent equation for the slip velocity in a two-phase flow condition has been incorporated into a developmental version of the RETRAN computer code. This model addition has been undertaken to remove a limitation in RETRAN-01 associated with the homogeneous equilibrium mixture model. In this paper, the development of the slip model is summarized and the corresponding constitutive equations are discussed. Comparisons of RETRAN analyses with steady-state void fraction data and data from the Semiscale S-02-6 small break test are also presented

  10. RETRAN-02 installation and verification for the CRAY computer

    International Nuclear Information System (INIS)

    1990-03-01

    The RETRAN-02 transient thermal-hydraulic analysis program developed by the Electric Power Research Institute (EPRI) has been selected as a tool for use in assessing the operation and safety of the SP-100 space reactor system being developed at Los Alamos National Laboratory (LANL). The released versions of RETRAN-02 are not operational on CRAY computer systems which are the primary mainframes in use at LANL requiring that the code be converted to the CRAY system. This document describes the code conversion, installation, and validation of the RETRAN-02/MOD004 code on the LANL CRAY computer system

  11. Reactivity transient calculatios in research reactor

    International Nuclear Information System (INIS)

    Santos, R.S. dos

    1986-01-01

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

  12. Development of a reactivity worth correction scheme for the one-dimensional transient analysis

    International Nuclear Information System (INIS)

    Cho, J. Y.; Song, J. S.; Joo, H. G.; Kim, H. Y.; Kim, K. S.; Lee, C. C.; Zee, S. Q.

    2003-11-01

    This work is to develop a reactivity worth correction scheme for the MASTER one-dimensional (1-D) calculation model. The 1-D cross section variations according to the core state in the MASTER input file, which are produced for 1-D calculation performed by the MASTER code, are incorrect in most of all the core states except for exactly the same core state where the variations are produced. Therefore this scheme performs the reactivity worth correction factor calculations before the main 1-D transient calculation, and generates correction factors for boron worth, Doppler and moderator temperature coefficients, and control rod worth, respectively. These correction factors force the one dimensional calculation to generate the same reactivity worths with the 3-dimensional calculation. This scheme is applied to the control bank withdrawal accident of Yonggwang unit 1 cycle 14, and the performance is examined by comparing the 1-D results with the 3-D results. This problem is analyzed by the RETRAN-MASTER consolidated code system. Most of all results of 1-D calculation including the transient power behavior, the peak power and time are very similar with the 3-D results. In the MASTER neutronics computing time, the 1-D calculation including the correction factor calculation requires the negligible time comparing with the 3-D case. Therefore, the reactivity worth correction scheme is concluded to be very good in that it enables the 1-D calculation to produce the very accurate results in a few computing time

  13. Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ferroukhi, H.; Coddington, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2001-07-01

    The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

  14. Analysis and sensitivity studies with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

    International Nuclear Information System (INIS)

    Ferroukhi, H.; Coddington, P.

    2001-01-01

    The OECD/NEA PWR rod ejection benchmark has been analysed using the 3-D nodal spatial-kinetic codes CORETRAN and RETRAN-3D. The following results were obtained. A) The agreement in 3-D solution between CORETRAN and RETRAN-3D was found to be very good both during steady-state and transient conditions. In particular at HZP (hot zero power), an excellent agreement in the initial steady-state 3-D power distribution and with regard to the core power excursion during the super-prompt critical phase of the transient (i.e. when the negative reactivity feedback is still very weak) was found. This illustrates the consistency in the neutronic solution between both codes. B) At both HZP and FP (full power) conditions, the CORETRAN and RETRAN-3D results lie well within the range of the previous benchmark solutions. In particular at HZP, both codes predict a power excursion and an increase in maximum pellet temperature that are among the closest results to those obtained with the benchmark reference solution. It must here be emphasised that these analyses are by no means a validation of the codes. However, the good agreement of both CORETRAN and RETRAN-3D with other 3-D solutions provides confidence in the ability of these codes to analyse LWR (light water reactor) core transients. In addition, it was found appropriate to perform, for this well-defined international benchmark problem, some sensitivity studies in order to assess the impact of modelling options on the CORETRAN and RETRAN-3D results. (authors)

  15. RETRAN dynamic slip model

    International Nuclear Information System (INIS)

    McFadden, J.H.; Paulsen, M.P.; Gose, G.C.

    1981-01-01

    Thermal-hydraulic codes in general use for system calculations are based on extensive analyses of loss-of-coolant accidents following the postulated rupture of a large coolant pipe. In this study, time-dependent equation for the slip velocity in a two-phase flow condition has been incorporated into the RETRAN-02 computer code. This model addition was undertaken to remove a limitation in RETRAN-01 associated with the homogeneous equilibrium mixture model. The dynamic slip equation was derived from a set of two-fluid conservation equations. 18 refs

  16. RETRAN-3D MOD003 Peach Bottom Turbine Trip 2 Multidimensional Kinetics Analysis Models and Results

    International Nuclear Information System (INIS)

    Mori, Michitsugu; Ogura, Katsunori; Gose, Garry C.; Wu, J.-Y.

    2003-01-01

    An analysis of the Peach Bottom Unit 2 Turbine Trip Test 2 (PB2/TT2) has been performed using RETRAN-3D MOD003. The purpose of the analysis was to investigate the PB2/TT2 overpressurization transient using the RETRAN-3D multidimensional kinetics model

  17. International RETRAN conference 1998: Proceedings

    International Nuclear Information System (INIS)

    Gose, G.; McFadden, J.

    1998-09-01

    The RETRAN computer code, developed by EPRI through its Contractors, EI International, Inc. (EI) and Computer Simulation and Analysis, Inc. (CSA), is now widely used by the international nuclear community for various types of safety analyses. In order to exchange information concerning the current use of the code by the various organizations, EPRI and CSA sponsored the Ninth International meeting in Monterey, California, on June 7--10, 1998. The opening session featured a panel discussion on Analysis Need for Supporting Nuclear Plants in the 21st Century by representatives of the US Nuclear Industry as well as a review of RETRAN activities since the last International RETRAN Meeting. During the three-day meeting, technical papers were presented by the various participants. The papers generally dealt with the following topics: (1) RETRAN-3D development, verification, and validation; (2) RETRAN-02/RETRAN-3D analysis of PWRs; (3) RETRAN-02/RETRAN-3D analysis of BWRs; and (4) CORETRAN development and analyses activities

  18. Qualification of RETRAN for simulator applications

    International Nuclear Information System (INIS)

    Harrison, J.F.

    1988-01-01

    The use of full-scope control room replica simulators increased substantially following the accident at Three Mile Island Unit 2. The technical capability required to represent severe events has been included, in varying degrees, in most simulators purchased since the TMI-2 accident. The ability of the instructor to create a large variety of combinations of malfunctions has also greatly expanded. The nuclear industry has developed a standard which establishes the minimum functional requirements for full-scope nuclear control room simulators used for operator training. This standard, ANSI/ANS-3.5, was first issued in 1981 and was reissued in 1985. A method for performing simulator qualification with best estimate analytical data has been proposed in EPRI NP-4243, Analytic Simulator Qualification Methodology. The idea presented there is to choose a set of transients which drive the simulator into all the system conditions (dynamic states) likely to be encountered during operator training. The key observable parameters for each state are compared to analyses performed with the best estimate analytical model The closeness of the comparison determines the fidelity of the simulator. The approach described in EPRI NP-4243 has been adapted for evaluating RETRAN's capability for use in simulator qualification. RETRAN analyses which compare the RETRAN results to plant or test facility data are evaluated with respect to the simulator test matrix documented in EPRI NP-4243

  19. Development of the RETRAN input model for Ulchin 3/4 visual system analyzer

    International Nuclear Information System (INIS)

    Lee, S. W.; Kim, K. D.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.; Hwang, M. K.

    2004-01-01

    As a part of the Long-Term Nuclear R and D program, KAERI has developed the so-called Visual System Analyzer (ViSA) based on best-estimate codes. The MARS and RETRAN codes are used as the best-estimate codes for ViSA. Between these two codes, the RETRAN code is used for realistic analysis of Non-LOCA transients and small-break loss-of-coolant accidents, of which break size is less than 3 inch diameter. So it is necessary to develop the RETRAN input model for Ulchin 3/4 plants (KSNP). In recognition of this, the RETRAN input model for Ulchin 3/4 plants has been developed. This report includes the input model requirements and the calculation note for the input data generation (see the Appendix). In order to confirm the validity of the input data, the calculations are performed for a steady state at 100 % power operation condition, inadvertent reactor trip and RCP trip. The results of the steady-state calculation agree well with the design data. The results of the other transient calculations seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the RETRAN input data can be used as a base input deck for the RETRAN transient analyzer for Ulchin 3/4. Moreover, it is found that Core Protection Calculator (CPC) module, which is modified by Korea Electric Power Research Institute (KEPRI), is well adapted to ViSA

  20. A novel feedwater system for the RETRAN model of the Palo Verde nuclear generating station

    International Nuclear Information System (INIS)

    Secker, P.A.; Webb, J.R.

    1988-01-01

    This paper presents a feedwater system model which supplies realistic boundary conditions to the RETRAN model of a Palo Verde Nuclear Generating Station reactor plant. The RETRAN thermal hydraulic code is used to analyze nuclear reactor system transients through a generalized thermal hydraulic volume/junction network. The feedwater system model is implemented using the control block modeling option available in the RETRAN code. The output of the control block model is coupled to the thermal hydraulic network by a fill junction. A forward Euler integration scheme is used by RETRAN for control block variables. The feedwater system model is formulated to allow implicit integration within the existing code framework. The potential need for small integration time steps is, therefore, alleviated. The model results are compared with test data

  1. Conceptual Design of GRIG (GUI Based RETRAN Input Generator)

    International Nuclear Information System (INIS)

    Lee, Gyung Jin; Hwang, Su Hyun; Hong, Soon Joon; Lee, Byung Chul; Jang, Chan Su; Um, Kil Sup

    2007-01-01

    For the development of high performance methodology using advanced transient analysis code, it is essential to generate the basic input of transient analysis code by rigorous QA procedures. There are various types of operating NPPs (Nuclear Power Plants) in Korea such as Westinghouse plants, KSNP(Korea Standard Nuclear Power Plant), APR1400 (Advance Power Reactor), etc. So there are some difficulties to generate and manage systematically the input of transient analysis code reflecting the inherent characteristics of various types of NPPs. To minimize the user faults and investment man power and to generate effectively and accurately the basic inputs of transient analysis code for all domestic NPPs, it is needed to develop the program that can automatically generate the basic input, which can be directly applied to the transient analysis, from the NPP design material. ViRRE (Visual RETRAN Running Environment) developed by KEPCO (Korea Electric Power Corporation) and KAERI (Korea Atomic Energy Research Institute) provides convenient working environment for Kori Unit 1/2. ViRRE shows the calculated results through on-line display but its capability is limited on the convenient execution of RETRAN. So it can not be used as input generator. ViSA (Visual System Analyzer) developed by KAERI is a NPA (Nuclear Plant Analyzer) using RETRAN and MARS code as thermal-hydraulic engine. ViSA contains both pre-processing and post-processing functions. In the pre-processing, only the trip data cards and boundary conditions can be changed through GUI mode based on pre-prepared text-input, so the capability of input generation is very limited. SNAP (Symbolic Nuclear Analysis Package) developed by Applied Programming Technology, Inc. and NRC (Nuclear Regulatory Commission) provides efficient working environment for the use of nuclear safety analysis codes such as RELAP5 and TRAC-M codes. SNAP covers wide aspects of thermal-hydraulic analysis from model creation through data analysis

  2. Analysis of steam line break of SMART using RETRAN-3D/INT

    International Nuclear Information System (INIS)

    Kim, Tae-Wan; Kim, Jong-Won; Park, Goon-Cherl

    2003-01-01

    RETRAN-3D has been modified to be suitable to safety analysis for integral type marine reactor with modular helical-coiled steam generator cassettes. The modified RETRAN-3D, RETRAN-3D/INT, has helical coil heat conductor model and heat transfer coefficient models for tube and shell sides of helical-coiled steam generator. In addition, moving models are added to simulate the effect of ship motions such as inclination, heaving, rolling and so on. RETRAN-3D/INT has been verified with natural circulation experiment conducted in Seoul National University and the analysis results for the first Japanese nuclear ship, MUTSU. In this study, the safety analysis for SMART, which has been developed by Korea Atomic Energy Research Institute, is performed to examine the applicability of RETRAN-3D/INT to the safety analysis of SMART. The steam line break is selected as reference case. The break type is assumed to the guillotine break. The loss of offsite power is considered as a coincident event and the failure of single train of passive residual heat removal system is assumed as single failure. From the results, it is found that RETRAN-3D/INT can appropriately simulate the transient of SMART and the improvement of non-condensable gas model is required. (author)

  3. Retran simulation of Oyster Creek generator trip startup test

    International Nuclear Information System (INIS)

    Alammar, M.A.

    1987-01-01

    RETRAN simulation of Oyster Creek generator trip startup test was carried out as part of Oyster Creek RETRAN model qualification program for reload licensing applications. The objective of the simulation was to qualify the turbine model and its interface with the control valve and bypass systems under severe transients. The test was carried out by opening the main breakers at rated power. The turbine speed governor closed the control valves and the pressure regulator opened the bypass valves within 0.5 sec. The stop valves closed by a no-load turbine trip, before the 10 percent overspeed trip was reached and the reactor scrammed on high APRM neutron flux. The simulation resulted in qualifying a normalized hydraulic torque for the turbine model and a 0.3 sec, delay block for the bypass model to account for the different delays in the hydraulic linkages present in the system. One-dimensional kinetics was used in this simulation

  4. Characteristics and use of the transient reactivity meter

    International Nuclear Information System (INIS)

    Yarbrough, W.M.

    1982-10-01

    At EG and G Idaho reactor facilities, reactivity measurements - an essential part of experimental reactor physics - are performed on line using an analog device known as the transient reactivity meter (TRM). The TRM has certain features that set it apart from most other instruments of its kind. This document describes these features and presents procedural information valuable to those who set up and use the TRM in a reactor measurement system

  5. RETRAN safety analyses of the nuclear-powered ship Mutsu

    International Nuclear Information System (INIS)

    Yoshinori, N.; Ishida, T.; Tanaka, Y.; Yoshiaki, F.

    1983-01-01

    A number of operational transient analyses of the nuclear-powered ship Mutsu have been performed in response to Japanese nuclear safety regulatory concerns. The RETRAN and COBRA-IV computer codes were used to provide a quantitative basis for the safety evaluation of the plant. This evaluation includes a complete loss of load without reactor scram, an excessive load increase incident, and an accidental depressurization of the primary system. The minimum departure from nucleate boiling ratio remained in excess of 1.53 for these three transients. Hence, the integrity of the core was shown to be maintained during these transients. Comparing the transient behaviors with those of land-based pressurized water reactors, the characteristic features of the Mutsu reactor were presented and the safety of the plant under the operational transient conditions was confirmed

  6. Development of a GUI-based RETRAN running environment and its application

    International Nuclear Information System (INIS)

    Kim, K.D.; Jeong, J.J.; Mo, S.Y.; Lee, Y.G.; Lee, C.B.

    2001-01-01

    In order to assist RETRAN users in their input preparation, code execution, and output interpretation, a visual interactive RETRAN running environment (ViRRE) has been developed. ViRRE provides dialog boxes and graphical modules for base input data generation and transient initiation on a user-friendly basis, and special graphical displays to provide an in-depth understanding of the major thermal-hydraulic phenomena during normal and accident conditions for nuclear power plants. This paper presents the main features of ViRRE and an example of its application. (authors)

  7. Development of a model for Retran-3D for pressure losses at T-junctions

    International Nuclear Information System (INIS)

    Barten, W.; Coddington, P.; Sullivan, J.

    2001-01-01

    For Nuclear Power Plants, both for PWRs and BWRs, there are many instances in normal operation, accidents and transients when it is important to know the pressure drop and quality of the flow, at a flow junction. In this paper two-phase pressure drops in a horizontal T-junction with equal areas are assessed in the case of separating flow using the RETRAN-3D code. Therefore applying the RETRAN-3D code first recoverable pressure drops are calculated for different flow rate ratios, inlet qualities and system pressures for steam-water mixtures. These pressure drops are then compared to analytical expressions by Soliman and Ebadian (1994) developed from the analysis of a wide range of two-phase experimental pressure drop data for equal-sided junctions with horizontal inlet and side branches both for steam-water and air-water flow. With these comparisons the experimental pressure loss is separated into a recoverable part (i.e. that calculated by RETRAN-3D) and an irrecoverable. A model for the irrecoverable pressure losses is derived for the RETRAN-3D code by comparing the RETRAN-3D total momentum equation with the expressions generally used for pressure changes at T-junctions. The results of this model are compared to the experimental data through the expressions of Soliman and Ebadian and are shown to produce very good comparisons particularly for the range of conditions consistent with the experimental data. (author)

  8. Development of a model for Retran-3D for pressure losses at T-junctions

    Energy Technology Data Exchange (ETDEWEB)

    Barten, W.; Coddington, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Sullivan, J. [Penn State University, Nuclear Engineering Program (United States)

    2001-07-01

    For Nuclear Power Plants, both for PWRs and BWRs, there are many instances in normal operation, accidents and transients when it is important to know the pressure drop and quality of the flow, at a flow junction. In this paper two-phase pressure drops in a horizontal T-junction with equal areas are assessed in the case of separating flow using the RETRAN-3D code. Therefore applying the RETRAN-3D code first recoverable pressure drops are calculated for different flow rate ratios, inlet qualities and system pressures for steam-water mixtures. These pressure drops are then compared to analytical expressions by Soliman and Ebadian (1994) developed from the analysis of a wide range of two-phase experimental pressure drop data for equal-sided junctions with horizontal inlet and side branches both for steam-water and air-water flow. With these comparisons the experimental pressure loss is separated into a recoverable part (i.e. that calculated by RETRAN-3D) and an irrecoverable. A model for the irrecoverable pressure losses is derived for the RETRAN-3D code by comparing the RETRAN-3D total momentum equation with the expressions generally used for pressure changes at T-junctions. The results of this model are compared to the experimental data through the expressions of Soliman and Ebadian and are shown to produce very good comparisons particularly for the range of conditions consistent with the experimental data. (author)

  9. RETRAN applications in pressurized thermal shock analysis of turkey point units 3 and 4

    International Nuclear Information System (INIS)

    Arpa, J.; Fatemi, A.S.; Mathavan, S.K.

    1985-01-01

    A methodology to assess the impact of overcooling transients on vessel wall integrity with respect to pressurized thermal shock conditions has been developed at Florida Power and Light Company for the Turkey Point Nuclear Units. Small break loss-of-coolant and small steamline break events have been simulated with the RETRAN code. Highly conservative assumptions, such as engineered safeguards with minimum temperature and maximum flow, have been made to maximize cooldown and thermal stress in the vessel wall. Temperatures, pressures, and flows obtained with RETRAN provide input for stress and fracture mechanics analyses that evaluate reactor vessel integrity. The results of the RETRAN analyses compare well with generic calculations performed by the Westinghouse Owners Group for a similar type of plant

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

    International Nuclear Information System (INIS)

    Baggoura, B.; Mazrou, H.

    2001-01-01

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

  11. Proceedings of a specialist meeting on boron reactivity transients

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The CSNI Specialist Meeting on Boron Dilution Reactivity Transients was hosted by the Penn State University in collaboration with the US Nuclear Regulatory Commission and the TRAC Users Group. More than 70 experts from 12 OECD countries, as well as experts from Russia and other non-OECD countries attended the meeting. Thirty papers were presented in five technical sessions. The purpose of the meeting was to bring together experts involved in the different activities related to boron dilution transients. The experts came from all involved parties, including research organizations, regulatory authorities, vendors and utilities. Information was openly shared and discussed on the experimental results, plant and systems analysis, numerical analysis of mixing and probability and consequences of these transients. Regulatory background and licensing implications were also included to provide the proper frame work for the technical discussion. Each of these areas corresponded to a separate session. The meeting focused on the thermal-hydraulic aspects because of the current interest in that subject and the significant amount of new technical information being generated. Three papers of the same conference are already available in INIS as individual reports: Potential for boron dilution during small-break LOCAs in PWRs (Ref. number: 27029412); Analysis of boron dilution in a four-loop PWR (Ref. number: 27051651); Probability and consequences of a rapid boron dilution sequence in a PWR (Ref. number: 27029411)

  12. Vector models in RETRAN-02 MOD 2

    International Nuclear Information System (INIS)

    Kinnersly, S.R.

    1985-06-01

    The vector momentum model in RETRAN-02 allows momentum flux to be modelled in two dimensions. Vector models in RETRAN-2 are described, including both the actual implementation in the code and the specification given in the code manual. The vector momentum model is described in detail. Other models which use vector quantities include models for volume average flow, volume average slip velocity, volume average phase velocities and fill junction flows. Both code implementations and code manual descriptions are described and inconsistencies noted. The differences between the standard RETRA-02 Mod 2 version and the Winfrith version RETN2204 are noted. (U.K.)

  13. RETRAN-3D Analysis Of The OECD/NRC Peach Bottom 2 Turbine Trip Benchmark

    International Nuclear Information System (INIS)

    Barten, W.; Coddington, P.

    2003-01-01

    This paper presents the PSI results on the different Phases of the Peach Bottom BWR Turbine Trip Benchmark using the RETRAN-3D code. In the first part of the paper, the analysis of Phase 1 is presented, in which the system pressure is predicted based on a pre-defined core power distribution. These calculations demonstrate the importance of accurate modelling of the non-equilibrium effects within the steam separator region. In the second part, a selection of the RETRAN-3D results for Phase 2 are given, where the power is predicted using a 3-D core with pre-defined core flow and pressure boundary conditions. A comparison of calculations using the different (Benchmark-specified) boundary conditions illustrates the sensitivity of the power maximum on the various resultant system parameters. In the third part of the paper, the results of the Phase 3 calculation are presented. This phase, which is a combination of the analytical work of Phases 1 and 2, gives good agreement with the measured data. The coupling of the pressure and flow oscillations in the steam line, the mass balance in the core, the (void) reactivity and the core power are all discussed. It is shown that the reactivity effects resulting from the change in the core void can explain the overall behaviour of the transient prior to the reactor scram. The time-dependent, normalized power for different thermal-hydraulic channels in the core is discussed in some detail. Up to the time of reactor scram, the power change was similar in all channels, with differences of the order of only a few percent. The axial shape of the channel powers at the time of maximum (overall) power increased in the core centre (compared with the shape at time zero). These changes occur as a consequence of the relative change in the channel void, which is largest in the region of the onset of boiling, and the influence on the different fuel assemblies of the complex ring pattern of the control rods. (author)

  14. RETRAN-3D Analysis Of The OECD/NRC Peach Bottom 2 Turbine Trip Benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Barten, W.; Coddington, P

    2003-03-01

    This paper presents the PSI results on the different Phases of the Peach Bottom BWR Turbine Trip Benchmark using the RETRAN-3D code. In the first part of the paper, the analysis of Phase 1 is presented, in which the system pressure is predicted based on a pre-defined core power distribution. These calculations demonstrate the importance of accurate modelling of the non-equilibrium effects within the steam separator region. In the second part, a selection of the RETRAN-3D results for Phase 2 are given, where the power is predicted using a 3-D core with pre-defined core flow and pressure boundary conditions. A comparison of calculations using the different (Benchmark-specified) boundary conditions illustrates the sensitivity of the power maximum on the various resultant system parameters. In the third part of the paper, the results of the Phase 3 calculation are presented. This phase, which is a combination of the analytical work of Phases 1 and 2, gives good agreement with the measured data. The coupling of the pressure and flow oscillations in the steam line, the mass balance in the core, the (void) reactivity and the core power are all discussed. It is shown that the reactivity effects resulting from the change in the core void can explain the overall behaviour of the transient prior to the reactor scram. The time-dependent, normalized power for different thermal-hydraulic channels in the core is discussed in some detail. Up to the time of reactor scram, the power change was similar in all channels, with differences of the order of only a few percent. The axial shape of the channel powers at the time of maximum (overall) power increased in the core centre (compared with the shape at time zero). These changes occur as a consequence of the relative change in the channel void, which is largest in the region of the onset of boiling, and the influence on the different fuel assemblies of the complex ring pattern of the control rods. (author)

  15. Tenth International RETRAN Conference Overview: RETRAN's Role in Supporting the Nuclear Industry's Vision

    International Nuclear Information System (INIS)

    Agee, Lance J.

    2003-01-01

    The nuclear industry's current 'vision' for 2020 is for growth in U.S. nuclear to a 23% share of generation in 2020. To support this vision, the Electric Power Research Institute's Nuclear Power Division has developed a strategic bridge plan. The major objectives of the plan are discussed. Of key importance is the U.S. Nuclear Regulatory Commission (NRC) staff's proposed framework for risk-informed regulations. RETRAN-3D will undoubtedly be used by the industry to support Risk-Informed Regulation, specifically option 3.The reason that RETRAN-3D is the most logical tool to support Risk-Informed Regulation is that in January 2001 the NRC issued a safety evaluation report (SER) on RETRAN-3D. The significance of the SER to the RETRAN community is described, and a list of the most important SER conditions provided.Next, the new and unique applications of RETRAN-3D are referenced. Finally, discussion of the future direction of safety software indicates what the author feels is needed to adequately support both existing plant upgrades and future plant designs

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

    International Nuclear Information System (INIS)

    Kawai, H.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Yang Shunhai

    1995-10-01

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

  18. Implementation of a new interfacial mass and energy transfer model in RETRAN-3D

    International Nuclear Information System (INIS)

    Macian, R.; Cebulh, P.; Coddington, P.; Paulsen, M.

    1999-01-01

    The RETRAN-3D MOD002.0 best estimate code includes a five-equation flow field model developed to deal with situations in which thermodynamic non-equilibrium phenomena are important. Several applications of this model to depressurization and pressurization transients showed serious convergence problems. An analysis of the causes for the numerical instabilities identified the models for interfacial heat and mass transfer as the source of the problems. A new interfacial mass and energy transfer model has thus been developed and implemented in RETRAN-3D. The heat transfer for each phase is equal to the product of the interfacial area density, a heat transfer coefficient and the temperature difference between the interface at saturation and the bulk temperature of the respective phase. However, in the context of RETRAN-3D, the vapor remains saturated in a two-phase volume, and no vapor heat transfer is thus calculated. The values of interfacial area density and heat transfer coefficient are obtained based on correlations appropriate for different flow regimes. A flow regime map, based on the work of Taitel and Dukler, with void fraction and mixture mass flux as map coordinates, is used to identify the flow regime present in a given volume. The new model has performed well when assessed against data from four experimental facilities covering depressurization, condensation and steady state void distribution. The results also demonstrate the viability of the approach followed to develop the new model for a five-equation based code. (author)

  19. Safety analyses of the nuclear-powered ship Mutsu with RETRAN

    International Nuclear Information System (INIS)

    Naruko, Y.; Ishida, T.; Tanaka, Y.; Futamura, Y.

    1982-01-01

    To provide a quantitative basis for the safety evaluation of the N.S. Mutsu, a number of safety analyses were performed in the course of reexamination. With respect to operational transient analyses, the RETRAN computer code was used to predict plant performances on the basis of postulated transient scenarios. The COBRA-IV computer code was also used to obtain a value of the minimum DNBR for each transient, which is necessary to predict detailed thermal-hydraulic performances in the core region of the reactor. In the present paper, the following three operational transients, which were calculated as a part of the safety analyses, are being dealt with: a complete loss of load without reactor scram; an excessive load increase incident, which is followed by a 30 percent stepwise load increase in the steam dump flow; and an accidental depressurization of the primary system, which is followed by a sudden full opening of the pressurizer spray valve. A Mutsu two-loop RETRAN model and simulation results were described. The results being compared with those of land-based PWRs, the characteristic features of the Mutsu reactor were presented and the safety of the plant under the operational transient conditions was confirmed

  20. Three-dimensional space-time kinetic analysis with CORETRAN and RETRAN-3D of the NEACRP PWR rod ejection benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ferroukhi, H.; Coddington, P

    2001-03-01

    One of the activities within the STARS project, in the Laboratory for Reactor Physics and System Behaviour; is the development of a coupling methodology between the three-dimensional, space-time kinetics codes CORETRAN and RETRAN-3D in order to perform core and plant transient analyses of the Swiss LWRs. The CORETRAN code is a 3-D full-core simulator, intended to be used for core-related analyses, while RETRAN-3D is the three-dimensional kinetics version of the plant system code RETRAN, and can therefore be used for best-estimate analyses of a wide range of transients in both PWRs and BWRs. Because the neutronics solver in both codes is based on the same kinetics model, one important advantage is that the codes can be coupled so that the initial conditions for a RETRAN-3D plant analysis are generated by a detailed-core, steady-state calculation using CORETRAN. As a first step towards using CORETRAN and RETRAN-3D for kinetic applications, the NEACRP PWR rod ejection benchmark has been analyzed with both codes, and is presented in this paper. The first objective is to verify the consistency between the static and kinetic solutions of the two codes, and so gain confidence in the coupling methodology. The second objective is to assess the CORETRAN and RETRAN-3D solutions for a well-defined RIA transient, comparing with previously published results. In parallel, several sensitivity studies have been performed in an attempt to identify models and calculational options important for a correct analysis of an RIA event in a LWR using these two codes. (author)

  1. Natural Circulation High Pressure Loop Dynamics Around Operating Point, Tests and Modelling With Retran 02

    International Nuclear Information System (INIS)

    Masriera, N.A; Doval, A.S; Mazufri, C.M

    2000-01-01

    The Natural Circulation High Pressure Loop (CAPCN) reproduces in scale all the one-dimensional thermal-hydraulic phenomena occurring in the primary loop of CAREM-25 reactor.It plays an important role in the qualification process of calculating computer codes.This facility demanded to develop several technological solutions in order to achieve the measuring and control quality required by that process.This engineering and experimental development allowed completing the first stage of dynamic tests during 1998.The trends of recorded data were systematically evaluated in terms of the deviations of main variables in response to different perturbations.By this analysis a group of eight transients was selected, providing a Minimum Representative Set (MRS) of dynamic tests, allowing the evaluation of all dynamic phenomena.Each of these transients was simulated with RETRAN-02, using a spreadsheet to facilitate the consistent elaboration and modification of input files.Comparing measured data and computer simulations, it may be concluded that it is possible to reproduce the dynamic response of all the transients with a level of approximation quite homogeneous and generally acceptable.It is possible to identify the detailed physical models that fit better the dynamic phenomena, and which of the limitations of RETRAN code are more relevant

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

    International Nuclear Information System (INIS)

    Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

    2003-01-01

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

  3. Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

    International Nuclear Information System (INIS)

    Kim, Y. H.; Lee, C. S.

    2003-01-01

    KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters

  4. Neutron and thermo - hydraulic model of a reactivity transient in a nuclear power plant fuel element

    International Nuclear Information System (INIS)

    Oliva, Jose de Jesus Rivero

    2012-01-01

    A reactivity transient without reactor scram was modeled and calculated using analytical expressions for the space distributions of the temperature fields, combined with discrete numerical calculations for the time dependences of thermal power and temperatures. The transient analysis covered the time dependencies of reactivity, global thermal power, fuel heat flux and temperatures in fuel, cladding and cooling water. The model was implemented in Microsoft Office Excel, dividing the Excel file in several separated worksheets for input data, initial steady-state calculations, calculation of parameters non-depending on eigenvalues, eigenvalues determination, calculation of parameters depending on eigenvalues, transient calculation and graphical representation of intermediate and final results. The results show how the thermal power reaches a new equilibrium state due to the negative reactivity feedback derived from the fuel temperature increment. Nevertheless, the reactor mean power increases 40% during the first second and, in the hottest channel, the maximum fuel temperature goes to a significantly high value, slightly above 2100 deg C, after 8 seconds of transient. Consequently, the results confirm that certain degree of fuel damage could be expected in case of a reactor scram failure. Once the basic model has being established the scope of accidents for future analyses can be extended, modifying the nuclear power behavior (reactivity) during transient and the boundary conditions for coolant temperature. A more complex model is underway for an annular fuel element. (author)

  5. Development of RETRAN-03/MOV code for thermal-hydraulic analysis of nuclear reactor under moving conditions

    International Nuclear Information System (INIS)

    Kim, Hak Jae; Park, Goon Cherl

    1996-01-01

    Nuclear ship reactors have several; features different from land-based PWR's. Especially, effects of ship motions on reactor thermal-hydraulics and good load following capability for abrupt load changes are essential characteristics of nuclear ship reactors. This study modified the RETRAN-03 to analyze the thermal-hydraulic transients under three-dimensional ship motions, named RETRAN-03/MOV in order to apply to future marine reactors. First Japanese nuclear ship MUTSU reactor have been analyzed under various ship motions to verify this code. Calculations have been performed under rolling,heaving and stationary inclination conditions during normal operation. Also, the natural circulation has been analyzed, which can provide the decay heat removed to ensure the passive safety of marine reactors. As results, typical thermal-hydraulic characteristics of marine reactors such as flow rate oscillations and S/G water level oscillations have been successfully simulated at various conditions. 7 refs., 11 figs. (author)

  6. The Evaluation of Steam Generator Level Measurement Model for OPR1000 Using RETRAN-3D

    International Nuclear Information System (INIS)

    Doo Yong Lee; Soon Joon Hong; Byung Chul Lee; Heok Soon Lim

    2006-01-01

    Steam generator level measurement is important factor for plant transient analyses using best estimate thermal hydraulic computer codes since the value of steam generator level is used for steam generator level control system and plant protection system. Because steam generator is in the saturation condition which includes steam and liquid together and is the place that heat exchange occurs from primary side to secondary side, computer codes are hard to calculate steam generator level realistically without appropriate level measurement model. In this paper, we prepare the steam generator models using RETRAN-3D that include geometry models, full range feedwater control system and five types of steam generator level measurement model. Five types of steam generator level measurement model consist of level measurement model using elevation difference in downcomer, 1D level measurement model using fluid mass, 1D level measurement model using fluid volume, 2D level measurement model using power and fluid mass, and 2D level measurement model using power and fluid volume. And we perform the evaluation of the capability of each steam generator level measurement model by simulating the real plant transient condition, the title is 'Reactor Trip by The Failure of The Deaerator Level Control Card of Ulchin Unit 3'. The comparison results between real plant data and RETRAN-3D analyses for each steam generator level measurement model show that 2D level measurement model using power and fluid mass or fluid volume has more realistic prediction capability compared with other level measurement models. (authors)

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

    International Nuclear Information System (INIS)

    Freels, J.D.

    1993-01-01

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

  8. Transient debris freezing and potential wall melting during a severe reactivity initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Moore, R.L.

    1981-01-01

    It is important to light water reactor (LWR) safety analysis to understand the transient freezing of molten core debris on cold structures following a hypothetical core meltdown accident. The purpose of this paper is to (a) present the results of a severe reactivity initiated accident (RIA) in-pile experiment with regard to molten debris distribution and freezing following test fuel rod failure, (b) analyze the transient freezing of molten debris (primarily a mixture of UO/sub 2/ fuel and Zircaloy cladding) deposited on the inner surface of the test shroud wall upon rod failure, and (c) assess the potential for wall melting upon being contacted by the molten debris. 26 refs

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

    International Nuclear Information System (INIS)

    Kawai, Hiroshi

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Xu Li; Shi Gong; Ma Dayuan; Yu Hong

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. A RETRAN-02 model of the Sizewell B PCSR design - the Winfrith one-loop model, version 3.0

    International Nuclear Information System (INIS)

    Kinnersly, S.R.

    1983-11-01

    A one-loop RETRAN-02 model of the Sizewell B Pre Construction Safety Report (PCSR) design, set up at Winfrith, is described and documented. The model is suitable for symmetrical pressurised transients. Comparison with data from the Sizewell B PCSR shows that the model is a good representation of that design. Known errors, limitations and deficiencies are described. The mode of storage and maintenance at Winfrith using PROMUS (Program Maintenance and Update System) is noted. It is recommended that users modify the standard data by adding replacement cards to the end so as to aid in identification, use and maintenance of local versions. (author)

  13. Reactive power generation in high speed induction machines by continuously occurring space-transients

    Science.gov (United States)

    Laithwaite, E. R.; Kuznetsov, S. B.

    1980-09-01

    A new technique of continuously generating reactive power from the stator of a brushless induction machine is conceived and tested on a 10-kw linear machine and on 35 and 150 rotary cage motors. An auxiliary magnetic wave traveling at rotor speed is artificially created by the space-transient attributable to the asymmetrical stator winding. At least two distinct windings of different pole-pitch must be incorporated. This rotor wave drifts in and out of phase repeatedly with the stator MMF wave proper and the resulting modulation of the airgap flux is used to generate reactive VA apart from that required for magnetization or leakage flux. The VAR generation effect increases with machine size, and leading power factor operation of the entire machine is viable for large industrial motors and power system induction generators.

  14. 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)

  15. Summary of transient analysis

    International Nuclear Information System (INIS)

    Saha, P.

    1984-01-01

    This chapter reviews the papers on the pressurized water reactor (PWR) and boiling water reactor (BWR) transient analyses given at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Most of the papers were based on the systems calculations performed using the TRAC-PWR, RELAP5 and RETRAN codes. The status of the nuclear industry in the code applications area is discussed. It is concluded that even though comprehensive computer codes are available for plant transient analysis, there is still a need to exercise engineering judgment, simpler tools and even hand calculations to supplement these codes

  16. Plant data comparisons for Comanche Peak 1/2 main feedwater pump trip transient

    Energy Technology Data Exchange (ETDEWEB)

    Boatwright, W.J.; Choe, W.G; Hiltbrand, D.W. [TU Electric, Dallas, TX (United States)] [and others

    1995-09-01

    A RETRAN-02 MOD5 model of Comanche Peak Steam Electric Station was developed by TU Electric for the purpose of performing core reload safety analyses. In order to qualify this model, comparisons against plant transient data from a partial loss of main feedwater flow were performed. These comparisons demonstrated that good representations of the plant response could be obtained with RETRAN-02 and the user-developed models of the primary-to-secondary heat transfer and plant control systems.

  17. RETRAN analysis of San Onofre Unit 2 turbine trip from 100% power

    International Nuclear Information System (INIS)

    Ting, Y.P.

    1985-01-01

    During the San Onofre Nuclear Generating Station Unit (SONGS 2) startup test, the plant experienced a turbine trip from 100% power on June 16, 1983. The trip was initiated by the condenser pressure switch malfunctioning. The plant computers were operating and recorded many plant key parameters. The resulting trip behaved as if it has been manually initiated and it was considered equivalent to a preplanned turbine trip test. A RETRAN-02 model was developed to simulate the SONGS 2 June 16 turbine trip event. The RETRAN analysis of the trip is a continuing effort of in-house SONGS 2 RETRAN model development to benchmark the calculations against the plant startup test data. The overall agreement between measured data and the RETRAN calculations was very good, providing confidence in the capability of the model and the RETRAN program. Comparative data are presented

  18. Development of a NSSS T/H Module for the YGN 1/2 NPP Simulator Using a Best-Estimate Code, RETRAN

    International Nuclear Information System (INIS)

    Seo, I. Y.; Lee, Y. K.; Jeun, G. D.; Suh, J. S.

    2005-01-01

    KEPRI(Korea Electric Power Research Institute) developed a realistic nuclear steam supply system thermal-hydraulic module, named ARTS code, based on the best-estimate code RETRAN for the improvement of the KNPEC(Korea Nuclear Plant Education Center) unit 2 full-scope simulator. In this work, we make a nuclear steam supply system thermal-hydraulic module for the YGN 1/2 nuclear power plant simulator using a practical application of a experience of ARTS code development. The ARTS code was developed based on RETRAN, which is a best estimate code developed by EPRI(Electric Power Research Institute) for various transient analyses of NPP(Nuclear Power Plants). Robustness and the real time calculation capability have been improved by simplifications, removing of discontinuities of the physical correlations of the RETRAN code and some other modifications. And its scope for the simulation has been extended by supplementation of new calculation modules such as a dedicated pressurizer relief tank model and a backup model. The supplement is developed so that users cannot recognize the model change from the main ARTS module

  19. Propagation of void fraction uncertainty measures in the RETRAN-3D simulation of the Peach Bottom turbine trip

    International Nuclear Information System (INIS)

    Vinai, Paolo; Macian-Juan, Rafael; Chawla, Rakesh

    2011-01-01

    The paper describes the propagation of void fraction uncertainty, as quantified by employing a novel methodology developed at Paul Scherrer Institut, in the RETRAN-3D simulation of the Peach Bottom turbine trip test. Since the transient considered is characterized by a strong coupling between thermal-hydraulics and neutronics, the accuracy in the void fraction model has a very important influence on the prediction of the power history and, in particular, of the maximum power reached. It has been shown that the objective measures used for the void fraction uncertainty, based on the direct comparison between experimental and predicted values extracted from a database of appropriate separate-effect tests, provides power uncertainty bands that are narrower and more realistic than those based, for example, on expert opinion. The applicability of such an approach to best estimate, nuclear power plant transient analysis has thus been demonstrated.

  20. Cerebral vasomotor reactivity: steady-state versus transient changes in carbon dioxide tension.

    Science.gov (United States)

    Brothers, R Matthew; Lucas, Rebekah A I; Zhu, Yong-Sheng; Crandall, Craig G; Zhang, Rong

    2014-11-01

    Cerebral vasomotor reactivity (CVMR) to changes in arterial carbon dioxide tension (P aCO 2) is assessed during steady-state or transient changes in P aCO 2. This study tested the following two hypotheses: (i) that CVMR during steady-state changes differs from that during transient changes in P aCO 2; and (ii) that CVMR during rebreathing-induced hypercapnia would be blunted when preceded by a period of hyperventilation. For each hypothesis, end-tidal carbon dioxide tension (P ET , CO 2) middle cerebral artery blood velocity (CBFV), cerebrovascular conductance index (CVCI; CBFV/mean arterial pressure) and CVMR (slope of the linear regression between changes in CBFV and CVCI versus P ET , CO 2) were assessed in eight individuals. To address the first hypothesis, measurements were made during the following two conditions (randomized): (i) steady-state increases in P ET , CO 2 of 5 and 10 Torr above baseline; and (ii) rebreathing-induced transient breath-by-breath increases in P ET , CO 2. The linear regression for CBFV versus P ET , CO 2 (P = 0.65) and CVCI versus P ET , CO 2 (P = 0.44) was similar between methods; however, individual variability in CBFV or CVCI responses existed among subjects. To address the second hypothesis, the same measurements were made during the following two conditions (randomized): (i) immediately following a brief period of hypocapnia induced by hyperventilation for 1 min followed by rebreathing; and (ii) during rebreathing only. The slope of the linear regression for CBFV versus P ET , CO 2 (P < 0.01) and CVCI versus P ET , CO 2 (P < 0.01) was reduced during hyperventilation plus rebreathing relative to rebreathing only. These results indicate that cerebral vasomotor reactivity to changes in P aCO 2 is similar regardless of the employed methodology to induce changes in P aCO 2 and that hyperventilation-induced hypocapnia attenuates the cerebral vasodilatory responses during a subsequent period of rebreathing

  1. Level-Swell Prediction With RETRAN-3D And Its Application To A BWR Steam-Line-Break Analysis

    International Nuclear Information System (INIS)

    Aounallah, Y.; Hofer, K.

    2003-01-01

    Level-swell experiments have often been simulated using system codes, such as TRAC and RELAP, but only cursory assessments have been performed with the operational-transient code RETRAN-3D, the main system code used within the STARS project. The present study, initiated in the framework of a BWR Steam-Line-Break (SLB) accident scenario, addresses this lacuna by performing RETRAN simulations of the General Electric Level-Swell experiments, and by investigating their implications on power plant accident analyses. Parameters to which the predicted level swell is sensitive have been identified, and recommendations on code options are made. The SLB analysis objective was to determine the amount of steam and liquid discharged through the break under specified boundary conditions, and to gauge the results against reference values. The impact of the nodalization of the upper part of the reactor pressure vessel was investigated and found to play an important role, whereas the level swell induced from flashing was found not to be the predominant factor for these simulations. (author)

  2. Rapid and transient stimulation of intracellular reactive oxygen species by melatonin in normal and tumor leukocytes

    International Nuclear Information System (INIS)

    Radogna, Flavia; Paternoster, Laura; De Nicola, Milena; Cerella, Claudia; Ammendola, Sergio; Bedini, Annalida; Tarzia, Giorgio; Aquilano, Katia; Ciriolo, Maria; Ghibelli, Lina

    2009-01-01

    Melatonin is a modified tryptophan with potent biological activity, exerted by stimulation of specific plasma membrane (MT1/MT2) receptors, by lower affinity intracellular enzymatic targets (quinone reductase, calmodulin), or through its strong anti-oxidant ability. Scattered studies also report a perplexing pro-oxidant activity, showing that melatonin is able to stimulate production of intracellular reactive oxygen species (ROS). Here we show that on U937 human monocytes melatonin promotes intracellular ROS in a fast (< 1 min) and transient (up to 5-6 h) way. Melatonin equally elicits its pro-radical effect on a set of normal or tumor leukocytes; intriguingly, ROS production does not lead to oxidative stress, as shown by absence of protein carbonylation, maintenance of free thiols, preservation of viability and regular proliferation rate. ROS production is independent from MT1/MT2 receptor interaction, since a) requires micromolar (as opposed to nanomolar) doses of melatonin; b) is not contrasted by the specific MT1/MT2 antagonist luzindole; c) is not mimicked by a set of MT1/MT2 high affinity melatonin analogues. Instead, chlorpromazine, the calmodulin inhibitor shown to prevent melatonin-calmodulin interaction, also prevents melatonin pro-radical effect, suggesting that the low affinity binding to calmodulin (in the micromolar range) may promote ROS production.

  3. KIVA3, Transient Multicomponent 2-D and 3-D Reactive Flows with Fuel Sprays

    International Nuclear Information System (INIS)

    Amsden, A.A.

    2001-01-01

    1 - Description of program or function: KIVA3VRELEASE2 is a computer program for the numerical calculation of transient, two and three-dimensional, chemically reactive flows with sprays. It is a newer version of the earlier KIVA3 (1993) that has now been extended to model vertical of canted valves in the cylinder head of a gasoline or diesel engine. KIVA3, in turn, was based on the earlier KIVA2 (1989) and uses the same numerical solution procedure and solves the same sort of equations. KIVA3VRELEASE2 uses a block-structured mesh with connectivity defined through indirect addressing. The departure from a single rectangular structure in logical space allows complex geometries to be modeled with significantly greater efficiency because large regions of deactivated cells are no longer necessary. Cell-face boundary conditions permit greater flexibility and simplification in the application of boundary conditions. KIVA3VRELEASE2 contains a number of significant changes. New features enhance the robustness, efficiency, and usefulness of the overall program for engine modeling. Automatic restart of the cycle with a reduced time-step in case of iteration limit or temperature overflow will reduce code crashes. A new option provides automatic deactivation of a port region when it is closed from the cylinder and reactivation when it communicates with the cylinder. Corrections in the code improve accuracy; extensions to the particle-based liquid wall film model makes the model more complete and a spli injection option has been added. A new subroutine monitors the liquid and gaseous fuel phases and energy balance data and emissions are monitored and printed. New features have been added to the grid generator K3PREP and the graphics post processor, K3POST. 2 - Method of solution: KIVA3VRELEASE2 solves the unsteady equations of motion of a turbulent, chemically reactive mixture of ideal gases, coupled to the equations for a single-component vaporizing fuel spray. The gas

  4. Transient voltage control of a DFIG-based wind power plant for suppressing overvoltage using a reactive current reduction loop

    Directory of Open Access Journals (Sweden)

    Geon Park

    2016-01-01

    Full Text Available This paper proposes a transient voltage control scheme of a doubly fed induction generator (DFIG-based wind power plant (WPP using a reactive current reduction loop to suppress the overvoltage at a point of interconnection (POI and DFIG terminal after a fault clearance. The change of terminal voltage of a DFIG is monitored at every predefined time period to detect the fault clearance. If the voltage change exceeds a set value, then the reactive current reduction loop reduces the reactive current reference in the DFIG controller using the step function. The reactive current injection of DFIGs in a WPP is rapidly reduced, and a WPP can rapidly suppress the overvoltage at a fault clearance because the reactive current reference is reduced. Using an electromagnetic transients program–released version (EMTP–RV simulator, the performance of the proposed scheme was validated for a model system comprising 20 units of a 5-MW DFIG considering various scenarios, such as fault and wind conditions. Test results show that the proposed scheme enables a WPP to suppress the overvoltage at the POI and DFIG terminal within a short time under grid fault conditions.

  5. RETRAN analysis of inter-system LOCA within the primary coolant pump

    International Nuclear Information System (INIS)

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

  6. Development of the RETRAN hot spot model for KSNP

    International Nuclear Information System (INIS)

    Kim, Yo Han; Kim, Yong Deog; Lee, Chang Sup

    2004-01-01

    Under the funding of Ministry of Commerce, Industry and Energy, Korea Electric Power Research Institute (KEPRI), the research center of Korea Electric Power Corporation (KEPCO), has been developed the in-house non-loss-of-coolant accident (non-LOCA) analysis methodology for Korea Standard Nuclear Power Plants (KSNP). To develop the methodology, the related documents of EPRI and vendors were examined and the methodologies of some foreign utilities were reviewed also to compensate for lack of capabilities. In fact, one of the major goals of the project is to build the code and methodology systems to replace the restricted codes by U. S. Government mentioned in the Technical Transfer Agreement between KEPCO and ABB-CE. To achieve the goal, KEPRI has developed the methodology based on general-purpose system codes, such as RETRAN, RELAP and MASTER. Despite the efforts, some functional weaks were raised from the users. So, KEPRI has developed the RETRAN hot spot model (HSM) to compensate some functions used for the estimation of the fuel temperature and enthalpy, cladding surface temperature, etc. In current methodology for KSNP, the parameters are calculated with STRIKIN-II code, which is one of the restricted codes. In this paper the development of HSM is described. And to estimate the feasibility of the model, the rod ejection accident (REA) was analyzed and the results were compared with those calculated by STRIKIN-II code. Through the feasibility study, it was concluded that the developed model showed the acceptable results and could be used further works

  7. ANO-2 turbine trip transient test analysis using MMS

    International Nuclear Information System (INIS)

    Jain, P.K.; Divakaruni, S.M.

    1984-01-01

    The data from the turbine trip transient tests conducted at the Arkansas Nuclear One-Unit 2 was used as one of the benchmark cases for validating the Modular Modeling System (MMS) Code, developed by the Electric Power Research Institute (EPRI). The data was used first to validate the modules in stand-alone simulation tests and then in a Nuclear Steam Supply system integral tests. This paper presents the results from the MMS simulation effort and compares the code generated results with the plant data as well as RETRAN results. In general, MMS simulation results compare very well with the plant data. The code calculations for the hot and cold leg temperatures, primary system pressure and the pressurizer level are very good compared to RETRAN; however, MMS results for steam generator level compare reasonably well only with RETRAN calculations

  8. Transient Control of Synchronous Machine Active and Reactive Power in Micro-grid Power Systems

    Science.gov (United States)

    Weber, Luke G.

    There are two main topics associated with this dissertation. The first is to investigate phase-to-neutral fault current magnitude occurring in generators with multiple zero-sequence current sources. The second is to design, model, and tune a linear control system for operating a micro-grid in the event of a separation from the electric power system. In the former case, detailed generator, AC8B excitation system, and four-wire electric power system models are constructed. Where available, manufacturers data is used to validate the generator and exciter models. A gain-delay with frequency droop control is used to model an internal combustion engine and governor. The four wire system is connected through a transformer impedance to an infinite bus. Phase-to-neutral faults are imposed on the system, and fault magnitudes analyzed against three-phase faults to gauge their severity. In the latter case, a balanced three-phase system is assumed. The model structure from the former case - but using data for a different generator - is incorporated with a model for an energy storage device and a net load model to form a micro-grid. The primary control model for the energy storage device has a high level of detail, as does the energy storage device plant model in describing the LC filter and transformer. A gain-delay battery and inverter model is used at the front end. The net load model is intended to be the difference between renewable energy sources and load within a micro-grid system that has separated from the grid. Given the variability of both renewable generation and load, frequency and voltage stability are not guaranteed. This work is an attempt to model components of a proposed micro-grid system at the University of Wisconsin Milwaukee, and design, model, and tune a linear control system for operation in the event of a separation from the electric power system. The control module is responsible for management of frequency and active power, and voltage and reactive

  9. Pre- and post-test analyses of a KKL turbine trip test at 109% power using RETRAN-3D

    Energy Technology Data Exchange (ETDEWEB)

    Coddington, P

    2001-03-01

    As part of the PSI/HSK STARS project, pre-test calculations have been performed for a KKL turbine trip test at 109% power using the RETRAN-3D code. In this paper, we first present the results of these calculations, together with a description of the test and a comparison of the results with the measured plant data, and then discuss in more detail the differences between the pre-test results and the plant measurements, including the differences in the initial and boundary conditions, and how these differences influenced the calculated results. Finally, we comment on a series of post-test and sensitivity analyses, which were performed to resolve some of the discrepancies. The results of the pre-test (blind) calculations show good overall agreement with the experimental data, particularly for the maximum in the steam-line mass flow rate following the opening of the turbine bypass valves. This is of critical importance, since the steam-line flow has the least margin to the reactor scram limit. The agreement is especially good since the control rod banks used for the selected rod insertion were changed from those given in the test specification. Following a review of the comparison of the pre-test calculations with the measured data, several deficiencies in the RETRAN-3D model for KKL were identified and corrected as part of the post-test analysis. This allowed for both an improvement in the calculated results, and a deeper understanding of the behaviour of the turbine trip transient. (author)

  10. Pre- and post-test analyses of a KKL turbine trip test at 109% power using RETRAN-3D

    International Nuclear Information System (INIS)

    Coddington, P.

    2001-01-01

    As part of the PSI/HSK STARS project, pre-test calculations have been performed for a KKL turbine trip test at 109% power using the RETRAN-3D code. In this paper, we first present the results of these calculations, together with a description of the test and a comparison of the results with the measured plant data, and then discuss in more detail the differences between the pre-test results and the plant measurements, including the differences in the initial and boundary conditions, and how these differences influenced the calculated results. Finally, we comment on a series of post-test and sensitivity analyses, which were performed to resolve some of the discrepancies. The results of the pre-test (blind) calculations show good overall agreement with the experimental data, particularly for the maximum in the steam-line mass flow rate following the opening of the turbine bypass valves. This is of critical importance, since the steam-line flow has the least margin to the reactor scram limit. The agreement is especially good since the control rod banks used for the selected rod insertion were changed from those given in the test specification. Following a review of the comparison of the pre-test calculations with the measured data, several deficiencies in the RETRAN-3D model for KKL were identified and corrected as part of the post-test analysis. This allowed for both an improvement in the calculated results, and a deeper understanding of the behaviour of the turbine trip transient. (author)

  11. Periodic transients linked to a variation in reactivity; Transitoires de periode lies a une variation de reactivite

    Energy Technology Data Exchange (ETDEWEB)

    Furet, J; Weil, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    We study here the influence of the transient, linked to a variation in reactivity, on the measurement of the period, this measurement being made from the logarithmic differential of the power and being defined by 1/T 1/p(dp/dt). We show that the adjustment of the thresholds of period safety is often incompatible with the velocities of liberation of reactivity. A compromise is then necessary between the speed of response of the periodimeter and the speed with which the reactivity is liberated. This makes it necessary to have rapid security devices for the power levels in the piles in which the speeds of liberation of the reactivity are high. (author) [French] On etudie ici l'influence du transitoire lie a une variation de la reactivite sur la mesure de la periode, cette mesure etant faite a partir de la derivee logarithmique de la puissance et etant definie par 1/T 1/p(dp/dt). On montre que le reglage des seuils de securite periode est souvent incompatible avec les vitesses de liberation de reactivite. Il y a alors un compromis a faire entre la vitesse de reponse du periodemetre et la vitesse de liberation de reactivite. Ceci impose de disposer de securites rapides sur les niveaux de puissance, dans les piles ou les vitesses de liberation de reactivite sont importantes. (auteur)

  12. Development of a GUI based RETRAN running environment for Kori NPP units 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Doo

    2000-09-01

    RETRAN was developed by EPRI and introduced for domestic use. RETRAN, which is a best-estimate system code approved by USNRC and used by most utilities in US, can be used in various plant support activities such as licensing calculations for plant design changes, EOP validation, and training. RETRAN, however, has been limited to only a few groups of specialists because of the difficulty involved in its usage. The aim of this project is to develop a graphic user interface (GUI) based code running environment for RETRAN named PRE (RETRAN Running Environment) in order to assist ordinary users in their input preparation, code execution, and output interpretation. TRIP and CONTROL BLOCK and VOLUME/JUNCTION input cards from base input are designed to be able to modify the existing input cards and add a new input cards through dialog boxes for users who have not much expertise in use of RETRAN. The RRE is designed to provide the calculated results though on-line X-Y graphs, plant mimics, indicators, nodalization window for easy interpretation of its output. It also provides the replay function using pre-calculated results saved in files. The RRE was developed for Kori NPP units 1 and 2 using Delphi 4.0 and Visual Fortran 6.0 and it runs on personal computers to increase the accessibility. The RRE developed in this study for Kori units 1 and 2 can be used in various plant support activities which require thermal-hydraulic analysis of the NSSS (Nuclear Steam Supply System) such as licensing calculations for plant design change, validation of EOP improvement, and operator training. The RRE developed can be expanded its application to other nuclear plants with low expense.

  13. Development of a GUI based RETRAN running environment for Kori NPP units 1 and 2

    International Nuclear Information System (INIS)

    Kim, Kyung Doo

    2000-09-01

    RETRAN was developed by EPRI and introduced for domestic use. RETRAN, which is a best-estimate system code approved by USNRC and used by most utilities in US, can be used in various plant support activities such as licensing calculations for plant design changes, EOP validation, and training. RETRAN, however, has been limited to only a few groups of specialists because of the difficulty involved in its usage. The aim of this project is to develop a graphic user interface (GUI) based code running environment for RETRAN named PRE (RETRAN Running Environment) in order to assist ordinary users in their input preparation, code execution, and output interpretation. TRIP and CONTROL BLOCK and VOLUME/JUNCTION input cards from base input are designed to be able to modify the existing input cards and add a new input cards through dialog boxes for users who have not much expertise in use of RETRAN. The RRE is designed to provide the calculated results though on-line X-Y graphs, plant mimics, indicators, nodalization window for easy interpretation of its output. It also provides the replay function using pre-calculated results saved in files. The RRE was developed for Kori NPP units 1 and 2 using Delphi 4.0 and Visual Fortran 6.0 and it runs on personal computers to increase the accessibility. The RRE developed in this study for Kori units 1 and 2 can be used in various plant support activities which require thermal-hydraulic analysis of the NSSS (Nuclear Steam Supply System) such as licensing calculations for plant design change, validation of EOP improvement, and operator training. The RRE developed can be expanded its application to other nuclear plants with low expense

  14. Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

    International Nuclear Information System (INIS)

    Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

    1997-01-01

    Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

  15. Classification of transient processes with a jumplike change in the reactivity

    International Nuclear Information System (INIS)

    Sabaeva, T.A.

    1989-01-01

    The problem of the change in the neutron flux density accompanying a jumplike (instantaneous) change in the reactivity is classical and is studied in most textbooks and monographs devoted to the regulation of nuclear reactors, where in constructing the response only the feedback on delayed neutrons is taken into account. The use of a linear feedback of a general form permits describing reactors of different types. A classification of feedbacks on reactivity was presented by Sabaeva, where a parabolic region in phase space is separated. A peak in the neutron flux corresponds to the image point falling into this region. In this paper the conditions making it possible to find the change in the neutrons flux immediately after an instantaneous change in the reactivity are derived, and the feedbacks are classified based on this

  16. Proceedings of the OECD/CSNI specialists meeting on boron dilution reactivity transients

    International Nuclear Information System (INIS)

    1997-06-01

    The purpose of the meeting was to bring together experts involved in the different activities related to boron dilution transients. The experts came from all involved parties, including research organizations, regulatory authorities, vendors and utilities. Information was openly shared and discussed on the experimental results, plant and systems analysis, numerical analysis of mixing and probability and consequences of these transients. Regulatory background and licensing implications were also included to provide the proper frame work for the technical discussion. Each of these areas corresponded to a separate session. The meeting focused on the thermal-hydraulic aspects because of the current interest in that subject and the significant amount of new technical information being generated

  17. RETRAN-3D analysis of the base case and the four extreme cases of the OECD/NRC Peach Bottom 2 Turbine Trip benchmark

    International Nuclear Information System (INIS)

    Barten, Werner; Coddington, Paul; Ferroukhi, Hakim

    2006-01-01

    This paper presents the results of RETRAN-3D calculations of the base case and the four extreme cases of phase 3 of the Peach Bottom 2 OECD/NRC Turbine Trip benchmark for coupled thermal-hydraulic and neutronic codes. The PSI-RETRAN-3D model gives good agreement with the measured data of the base case. In addition to the base case, the analysis of the extreme cases provides a further understanding of the reactor behaviour, which is the result of the dynamic coupling of the whole system, i.e., the interaction between the steam line and vessel flows, the pressure, the Doppler, void and control reactivity and power. For the extreme cases without scram the bank of safety relief valves is able to mitigate the effects of the turbine trip for short times. The 3-D nature of the core power distribution has been investigated by analysing the power density of the different thermal-hydraulic channels. In all cases prior to the reactor scram the course of the power is similar in all the channels with differences of the order of a few percent showing that, by and large, the core acts in a coherent manner. At the time of maximum power, the axial power distribution in the different channels is increased at the core centre with respect to the distribution at time zero, by an amount, which is different for the different channels

  18. Guanosine radical reactivity explored by pulse radiolysis coupled with transient electrochemistry.

    Science.gov (United States)

    Latus, A; Alam, M S; Mostafavi, M; Marignier, J-L; Maisonhaute, E

    2015-06-04

    We follow the reactivity of a guanosine radical created by a radiolytic electron pulse both by spectroscopic and electrochemical methods. This original approach allows us to demonstrate that there is a competition between oxidation and reduction of these intermediates, an important result to further analyse the degradation or repair pathways of DNA bases.

  19. Assessment of void fraction prediction using the RETRAN-3d and CORETRAN-01/VIPRE-02 codes

    International Nuclear Information System (INIS)

    Aounallah, Y.; Coddington, P.; Gantner, U.

    2000-01-01

    A review of wide-range void fraction correlations against an extensive database has been undertaken to identify the correlations best suited for nuclear safety applications. Only those based on the drift-flux model have been considered. The survey confirmed the application range of the Chexal-Lellouche correlation, and the database was also used to obtain new parameters for the Inoue drift-flux correlation, which was also found suitable. A void fraction validation study has also been undertaken for the codes RETRAN-3D and CORETRAN-01/VIPRE-02 at the assembly and sub-assembly levels. The study showed the impact of the RETRAN-03 user options on the predicted void fraction, and the RETRAN-3D limitation at very low fluid velocity. At the sub-assembly level, CORETRAN-01/VIPRE-02 substantially underestimates the void in regions with low power-to-flow ratios. Otherwise, a generally good predictive performance was obtained with both RETRAN-3D and CORETRAN-01/VIPRE-02. (authors)

  20. Assessment of void fraction prediction using the RETRAN-3d and CORETRAN-01/VIPRE-02 codes

    Energy Technology Data Exchange (ETDEWEB)

    Aounallah, Y.; Coddington, P.; Gantner, U

    2000-07-01

    A review of wide-range void fraction correlations against an extensive database has been undertaken to identify the correlations best suited for nuclear safety applications. Only those based on the drift-flux model have been considered. The survey confirmed the application range of the Chexal-Lellouche correlation, and the database was also used to obtain new parameters for the Inoue drift-flux correlation, which was also found suitable. A void fraction validation study has also been undertaken for the codes RETRAN-3D and CORETRAN-01/VIPRE-02 at the assembly and sub-assembly levels. The study showed the impact of the RETRAN-03 user options on the predicted void fraction, and the RETRAN-3D limitation at very low fluid velocity. At the sub-assembly level, CORETRAN-01/VIPRE-02 substantially underestimates the void in regions with low power-to-flow ratios. Otherwise, a generally good predictive performance was obtained with both RETRAN-3D and CORETRAN-01/VIPRE-02. (authors)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  3. CYLFUX, Fast Reactor Reactivity Transients Simulation in LWR by 2-D 2 Group Diffusion

    International Nuclear Information System (INIS)

    Schmidt, A.

    1973-01-01

    1 - Nature of physical problem solved: A 2-dimensional calculation of the 2-group, space-dependent neutron diffusion equations is performed in r-z geometry using an arbitrary number of groups of delayed neutron precursors. The program is designed to simulate fast reactivity excursions in light water reactors taking into account Doppler feedback via adiabatic heatup of fuel. Axial motions of control rods may be considered including scram action on option. 2 - Method of solution: The differential equations are solved at each time step by an explicit finite difference method using two time levels. The stationary distributions are obtained by using the same algorithm. 3 - Restrictions on the complexity of the problem: No restriction to the number of space points and delayed neutron energy groups besides the computer size

  4. SPECT study of cerebral blood flow reactivity after acetazolamide in patients with transient ischemic attacks

    International Nuclear Information System (INIS)

    Chollet, F.; Celsis, P.; Clanet, M.; Guiraud-Chaumeil, B.; Rascol, A.; Marc-Vergnes, J.P.

    1989-01-01

    We investigated 15 patients with one or more transient ischemic attacks (TIAs) in the internal carotid artery territory within the month following the most recent TIA. Cerebral blood flow (CBF) was measured by single-photon emission computed tomography, using intravenous xenon-133 before and after injection of 1 g acetazolamide. Six patients had severe carotid stenosis or occlusion; the other nine patients had no significant carotid lesions. Twenty age-matched volunteers free of neurologic symptoms or history were used as controls. Mean CBF in the sylvian region was not significantly different between patients and controls. Seven patients exhibited a focal hypoperfusion at rest in the symptomatic hemisphere, and their hypoperfused areas were hyporeactive after administration of acetazolamide. Seven other patients exhibited hyporeactive areas after acetazolamide administration while their CBF tomograms at rest were normal. Thus, CBF abnormalities were detected in 14 of the 15 patients. Our findings suggest that CBF measured early after acetazolamide administration could be useful to confirm the clinical diagnosis of TIA. In the nine patients with no significant lesion of the internal carotid artery, the areas of hypoperfusion were small and were probably related to the focal ischemic event. In the six patients with severe lesions of the internal carotid artery, abnormalities were of variable size and intensity but were often large and pronounced. The discrepancy between these two subgroups of patients could be ascribed to the hemodynamic influence of the internal carotid artery lesions. Moreover, our findings may provide some insight into the pathophysiology of TIAs

  5. SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR

    International Nuclear Information System (INIS)

    Halsall, M.J.; Course, A.F.; Sidell, J.

    1979-09-01

    SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)

  6. Development and verification of a coupled code system RETRAN-MASTER-TORC

    International Nuclear Information System (INIS)

    Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

    2004-01-01

    Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

  7. Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

    2013-01-01

    Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

  8. RETRAN simulation of Oyster Creek MSIV closure and bypass valve tests

    International Nuclear Information System (INIS)

    Alammar, M.A.

    1987-01-01

    A series of benchmarks against start-up tests have been performed on the Oyster Creek boiling water reactor unit 2 RETRAN model in support of developing an in-house reload capability. The liquid and the pressure regulator models have been benchmarked against level and pressure setpoint changes, where small setpoint perturbations were made at rated power. The purpose of the present benchmark is to check the liquid level behavior during a severe level drop as during void collapse following a scram and to size the bypass valves by benchmarking the valves' contraction coefficient. The main steam isolation valves (MSIVs) closure start-up test was chosen for the former, while the bypass valve test was chosen for the latter. The two benchmarks complete the qualification of the upper downcomer liquid level for small and large level changes and the pressure regulator system for the Oyster Creek RETRAN model

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  10. A Comparative analysis for control rod drop accident in RETRAN DNB and CETOP DNB Model

    International Nuclear Information System (INIS)

    Yang, Chang Keun; Kim, Yo Han; Ha, Sang Jun

    2009-01-01

    In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method

  11. Validation of RETRAN-03 by simulating a peach bottom turbine trip and boiloff at the full integral simulation test facility

    International Nuclear Information System (INIS)

    Westacott, J.L.; Peterson, C.E.

    1992-01-01

    This paper reports that the RETRAN-03 computer code is validated by simulating two tests that were performed at the Full Integral Simulation Test (FIST) facility. The RETRAN-03 results of a turbine trip (test 4PTT1) and failure to maintain water level at decay power (test T1QUV) are compared with the FIST test data. The RETRAN-03 analysis of test 4PTT1 is compared with a previous TRAC-BWR analysis of the test. Sensitivity to various model nodalizations and RETRAN-03 slip options are studied by comparing results of test T1QUV. The predicted thermal-hydraulic responses of both tests agree well with the test data. The pressure response of test 4PTT1 and the boiloff rate for test T1QUV are accurately predicted. Core uncovery time is found to be sensitive to the upper downcomer and upper plenum nodalization. The RETRAN-03 algebraic and dynamic slip options produce similar results for test T1QUV

  12. RETRAN-02 analysis of upper head cooling during controlled natural circulation cooldown of Yankee Nuclear Power Station

    International Nuclear Information System (INIS)

    Fujita, N.; Helrich, R.E.; Bergeron, P.A.

    1982-01-01

    RETRAN-02 is particularly well-suited for investigating the fluid conditions in the upper head during a natural circulation cooldown. The RETRAN input model was developed with four basic objectives: (1) accurate description of the upper head cooling mechanisms; (2) proper simulation of natural circulation; (3) respresentations of operator actions required to proceed from full-power to shutdown-cooling-system conditions using both automatic and manual controls; and (4) reduction of the computer cost of simulating this evolution of approximately 10-hour duration. The response of the upper head fluid temperature calculated by RETRAN was in close agreement with measured data obtained from a natural circulation cooldown experiment performed for the Connecticut Yankee Plant, whose design is very similar to the Yankee Nuclear Power Station

  13. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  14. Using Rising Limb Analysis to Estimate Uptake of Reactive Solutes in Advective and Transient Storage Sub-compartments of Stream Ecosystems

    Science.gov (United States)

    Thomas, S. A.; Valett, H.; Webster, J. R.; Mulholland, P. J.; Dahm, C. N.

    2001-12-01

    Identifying the locations and controls governing solute uptake is a recent area of focus in studies of stream biogeochemistry. We introduce a technique, rising limb analysis (RLA), to estimate areal nitrate uptake in the advective and transient storage (TS) zones of streams. RLA is an inverse approach that combines nutrient spiraling and transient storage modeling to calculate total uptake of reactive solutes and the fraction of uptake occurring within the advective sub-compartment of streams. The contribution of the transient storage zones to solute loss is determined by difference. Twelve-hour coinjections of conservative (Cl-) and reactive (15NO3) tracers were conducted seasonally in several headwater streams among which AS/A ranged from 0.01 - 2.0. TS characteristics were determined using an advection-dispersion model modified to include hydrologic exchange with a transient storage compartment. Whole-system uptake was determined by fitting the longitudinal pattern of NO3 to first-order, exponential decay model. Uptake in the advective sub-compartment was determined by collecting a temporal sequence of samples from a single location beginning with the arrival of the solute front and concluding with the onset of plateau conditions (i.e. the rising limb). Across the rising limb, 15NO3:Cl was regressed against the percentage of water that had resided in the transient storage zone (calculated from the TS modeling). The y-intercept thus provides an estimate of the plateau 15NO3:Cl ratio in the absence of NO3 uptake within the transient storage zone. Algebraic expressions were used to calculate the percentage of NO3 uptake occurring in the advective and transient storage sub-compartments. Application of RLA successfully estimated uptake coefficients for NO3 in the subsurface when the physical dimensions of that habitat were substantial (AS/A > 0.2) and when plateau conditions at the sampling location consisted of waters in which at least 25% had resided in the

  15. RETRANS - A tool to verify the functional equivalence of automatically generated source code with its specification

    International Nuclear Information System (INIS)

    Miedl, H.

    1998-01-01

    Following the competent technical standards (e.g. IEC 880) it is necessary to verify each step in the development process of safety critical software. This holds also for the verification of automatically generated source code. To avoid human errors during this verification step and to limit the cost effort a tool should be used which is developed independently from the development of the code generator. For this purpose ISTec has developed the tool RETRANS which demonstrates the functional equivalence of automatically generated source code with its underlying specification. (author)

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

    International Nuclear Information System (INIS)

    Yoshida, K.; Naser, J.A.

    1988-05-01

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

  17. The development of high performance numerical simulation code for transient groundwater flow and reactive solute transport problems based on local discontinuous Galerkin method

    International Nuclear Information System (INIS)

    Suzuki, Shunichi; Motoshima, Takayuki; Naemura, Yumi; Kubo, Shin; Kanie, Shunji

    2009-01-01

    The authors develop a numerical code based on Local Discontinuous Galerkin Method for transient groundwater flow and reactive solute transport problems in order to make it possible to do three dimensional performance assessment on radioactive waste repositories at the earliest stage possible. Local discontinuous Galerkin Method is one of mixed finite element methods which are more accurate ones than standard finite element methods. In this paper, the developed numerical code is applied to several problems which are provided analytical solutions in order to examine its accuracy and flexibility. The results of the simulations show the new code gives highly accurate numeric solutions. (author)

  18. The influence of the reactivity ramp on the course of the power transient in the MARK 1A core of the SNR 300

    International Nuclear Information System (INIS)

    Froehlich, R.; Schmuck, P.

    1976-01-01

    The course of a hypothetic transient overpower accident caused by the onset of a not further specified reactivity ramp accompanied by the simultaneous failure of both shutdown systems must be analyzed in the SNR 300 Mark 1A core licensing procedure. The present study is limited to the discussion of the starting and shutdown phases of such accidents for the fresh core. Depending on the operational state of the reactor, the core geometry is still intact during the starting phase. In the following shutdown phase (core disassembly phase), large-scale mass transfer leads to the nuclear shutdown of the reactor. (orig./AK) [de

  19. Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

    Directory of Open Access Journals (Sweden)

    Isabelle Guénot-Delahaie

    2018-03-01

    Full Text Available The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs, power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs. As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions.This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur.Areas of improvement are finally discussed with a view to simulating and

  20. Preemptive, but not reactive, spinal cord stimulation mitigates transient ischemia-induced myocardial infarction via cardiac adrenergic neurons

    NARCIS (Netherlands)

    Southerland, E. M.; Milhorn, D. M.; Foreman, R. D.; Linderoth, B.; DeJongste, M. J. L.; Armour, J. A.; Subramanian, V.; Singh, M.; Singh, K.; Ardell, J. L.

    2007-01-01

    Our objective was to determine whether electrical neuromodulation using spinal cord stimulation ( SCS) mitigates transient ischemia-induced ventricular infarction and, if so, whether adrenergic neurons are involved in such cardioprotection. The hearts of anesthetized rabbits, subjected to 30 min of

  1. Disruption of the blood-brain interface in neonatal rat neocortex induces a transient expression of metallothionein in reactive astrocytes

    DEFF Research Database (Denmark)

    Penkowa, M; Moos, T

    1995-01-01

    rats were subjected to a localized freeze lesion of the neocortex of the right temporal cortex. This lesion results in a disrupted blood-brain interface, leading to extravasation of plasma proteins. From 16 h, reactive astrocytosis, defined as an increase in the number and size of cells expressing GFAP...

  2. Development of a Nuclear Steam Supply System Thermal-Hydraulic Module for the Nuclear Power Plant Simulator Using a Best-Estimate Code, RETRAN

    International Nuclear Information System (INIS)

    Suh, Jae Seung

    2004-08-01

    The NSSS (Nuclear Steam Supply System) thermal-hydraulic programs adopted in the domestic full-scope power plant simulators were provided in early 1980s by foreign vendors. Because of limited computational capability at that time, they usually used very simplified physical models for the real-time simulation of Ness thermal-hydraulic transients, which entails inaccurate results and, thus, the possibility of so-called 'negative training', especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, a realistic NSSS thermal-hydraulic program ARTS has been developed, it was based on the RETRAN code for the improvement of the Nuclear Power Plant full-scope simulator. Since ARTS is a generalized code to solve a simultaneous equation system, the smaller time-step size should be used if converged solution could not obtain even in a single volume. Therefore, dedicated models which do not force to reduce the time-step size are sometimes more suitable in terms of a real-time calculation and robustness. The PRT(Pressurizer Relief Tank) is a good example, which requires a dedicated model. The PRT consists of subcooled water in bottom and non-condensable gas in top. The sparger merged under subcooled water enhances condensation. The complicated thermal-hydraulic phenomena such as condensation, phase separation with existence of non-condensable gas makes difficult to simulate. Therefore, the PRT volume may limit the time-step size if it is modeled with a general control volume. To mitigate the time-step size reduction due to convergence failure at this component using RETRAN, the PRT model was developed as a dedicated model. The dedicated model was expected to provide reasonable results without convergence problem in the analysis of the system transients. The ARTS code guarantees the real-time calculations of almost all transients and ensures the robustness of simulations. However, there are some possibilities of calculation failure in the

  3. APR1400 Locked Rotor Transient Analysis using KNAP

    International Nuclear Information System (INIS)

    Lee, Dong-Hyuk; Kim, Yo-Han; Ha, Sang Jun

    2007-01-01

    KEPRI (Korea Electric Power Research Institute) has developed safety analysis methodology for non-LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code for most transients. For locked rotor transient DNBR analysis, UNICORN-TM code is used. UNICORN-TM is the unified code of RETRAN, MASTER and TORC. The UNICORN-TM has 1-D and 3-D neutron kinetics calculation capability. For locked rotor DNBR analysis, 1-D neutron kinetics is used. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400) locked rotor analysis and compare the results with those in the APR1400 SSAR(Standard Safety Analysis Report). The locked rotor transient is one of the 'decrease in reactor coolant system flow rate' events and the results are typically described in the chapter 15.3.3 of SAR (Safety Analysis Report). In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, locked rotor transient is analyzed using UNICORN-TM code and the results are compared with those from APR1400 SSAR

  4. APR1400 Locked Rotor Transient Analysis using KNAP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong-Hyuk; Kim, Yo-Han; Ha, Sang Jun [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    KEPRI (Korea Electric Power Research Institute) has developed safety analysis methodology for non-LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code for most transients. For locked rotor transient DNBR analysis, UNICORN-TM code is used. UNICORN-TM is the unified code of RETRAN, MASTER and TORC. The UNICORN-TM has 1-D and 3-D neutron kinetics calculation capability. For locked rotor DNBR analysis, 1-D neutron kinetics is used. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400) locked rotor analysis and compare the results with those in the APR1400 SSAR(Standard Safety Analysis Report). The locked rotor transient is one of the 'decrease in reactor coolant system flow rate' events and the results are typically described in the chapter 15.3.3 of SAR (Safety Analysis Report). In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, locked rotor transient is analyzed using UNICORN-TM code and the results are compared with those from APR1400 SSAR.

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

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-10-01

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

  6. Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

    International Nuclear Information System (INIS)

    Sankaranarayanan, S.

    2004-01-01

    Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  8. Cerebral blood flow and CO2 reactivity in transient ischemic attacks: comparison between TIAs due to the ICA occlusion and ICA mild stenosis

    International Nuclear Information System (INIS)

    Tsuda, Y.; Kimura, K.; Yoneda, S.; Etani, H.; Asai, T.; Nakamura, M.; Abe, H.

    1983-01-01

    Hemispheric mean cerebral blood flow (CBF), together with its CO2 reactivity in response to hyperventilation, was investigated in 18 patients with transient ischemic attacks (TIAs) by intraarterial 133Xe injection method in a subacute-chronic stage of the clinical course. In 8 patients, the lesion responsible for symptoms was regarded as unilateral internal carotid artery (ICA) occlusion, and in 10 patients, it was regarded as unilateral ICA mild stenosis (less than 50% stenosis in diameter). Resting flow values were significantly decreased in the affected hemisphere of TIA due to the ICA occlusion as compared with the unaffected hemisphere of the same patient, regarded as the relative control. It was not decreased in the affected hemisphere of TIA due to the ICA mild stenosis as compared with the control. With respect to the responsiveness of CBF to changes in PaCO2, it was preserved in both TIAs, due to the ICA occlusion and ICA mild stenosis. Vasoparalysis was not observed in either types of TIAs in the subacute-chronic stage. However, in the relationship of blood pressure and CO2 reactivity, expressed as delta CBF(%)/delta PaCO2, pressure-dependent CO2 reactivity as a group was observed with significance in 8 cases of TIA due to the ICA occlusion, while no such relationship was noted in 10 cases of TIA due to the ICA mild stenosis. Moreover, clinical features were different between TIAs due to the ICA occlusion and ICA mild stenosis, i.e., more typical, repeatable TIA (6.3 +/- 3.7 times) with shorter duration (less than 30 minutes) was observed in TIAs due to the ICA mild stenosis, while more prolonged, less repeatable TIA (2.4 +/- 1.4 times) was observed in TIAs due to fixed obstruction of the ICA. From these observations, two different possible mechanisms as to the pathogenesis of TIA might be expected

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

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Sunagawa, Takeyoshi

    2014-01-01

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

  10. Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

    International Nuclear Information System (INIS)

    Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

    2004-01-01

    Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

  11. Poison 1 - a programme for calculation of reactivity transients due to fission product poisoning and its application in continuous determination of xenon and samarium poisoning in reactor KS-150

    International Nuclear Information System (INIS)

    Rana, S.B.

    1973-12-01

    The report contains a user's description of the 3-dimensional programme POISON 1 for calculating reactivity transients due to fission-product poisoning during changes of reactor power. The chapter dealing with Xe poisoning contains a description of Xe tables, the method of operational determination of Xe poisoning, use of Xe transients for calibrating control rods and means of shutting down the reactor without being overriden by Xe poisoning. Sm poisoning is determined continuously on the basis of the power diagram of reactor operation. In conclusion a possibility of using the programme in a process computer in combination with self-powered detectors as local power sensors is indicated. (author)

  12. C-reactive protein predicts further ischemic events in first-ever transient ischemic attack or stroke patients with intracranial large-artery occlusive disease.

    Science.gov (United States)

    Arenillas, Juan F; Alvarez-Sabín, José; Molina, Carlos A; Chacón, Pilar; Montaner, Joan; Rovira, Alex; Ibarra, Bernardo; Quintana, Manuel

    2003-10-01

    The role of inflammation in intracranial large-artery occlusive disease is unclear. We sought to investigate the relationship between high-sensitivity C-reactive protein (CRP) levels and the risk of further ischemic events in first-ever transient ischemic attack (TIA) or stroke patients with intracranial large-artery occlusive disease. Of a total of 127 consecutive first-ever TIA or ischemic stroke patients with intracranial stenoses detected by transcranial Doppler ultrasonography, 71 fulfilled all inclusion criteria, which included angiographic confirmation. Serum high-sensitivity CRP level was determined a minimum of 3 months after the qualifying event. Patients were followed up during 1 year after blood sampling. Thirteen patients (18.3%) with intracranial large-artery occlusive disease experienced an end point event: 9 cerebral ischemic events, 7 of which were attributable to intracranial large-artery occlusive disease, and 4 myocardial infarctions. Patients in the highest quintile of high-sensitivity CRP level had a significantly higher adjusted odds ratio for new events compared with those in the first quintile (odds ratio, 8.66; 95% CI, 1.39 to 53.84; P=0.01). A high-sensitivity CRP level above the receiver operating characteristic curve cutoff value of 1.41 mg/dL emerged as an independent predictor of new end point events (hazard ratio, 7.14; 95% CI, 1.77 to 28.73; P=0.005) and of further intracranial large-artery occlusive disease-related ischemic events (hazard ratio, 30.67; 95% CI, 3.6 to 255.5; P=0.0015), after adjustment for age, sex, and risk factors. Kaplan-Meier curves showed that a significantly lower proportion of patients with a high-sensitivity CRP >1.41 mg/dL remained free of a new ischemic event (P<0.0001). High-sensitivity CRP serum level predicts further intracranial large-artery occlusive disease-related and any major ischemic events in patients with first-ever TIA or stroke with intracranial large-artery occlusive disease. These findings

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

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Asaka, Hideaki; Ishida, Toshihisa

    1986-01-01

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

  14. Study on the reactive transient α-λ3-iodanyl-acetophenone complex in the iodine(III)/PhI(I) catalytic cycle of iodobenzene-catalyzed α-acetoxylation reaction of acetophenone by electrospray ionization tandem mass spectrometry.

    Science.gov (United States)

    Wang, Hao-Yang; Zhou, Juan; Guo, Yin-Long

    2012-03-30

    Hypervalent iodine compounds are important and widely used oxidants in organic chemistry. In 2005, Ochiai reported the PhI-catalyzed α-acetoxylation reaction of acetophenone by the oxidation of PhI with m-chloroperbenzoic acid (m-CPBA) in acetic acid. However, until now, the most critical reactive α-λ(3)-iodine alkyl acetophenone intermediate (3) had not been isolated or directly detected. Electrospray ionization tandem mass spectrometry (ESI-MS/MS) was used to intercept and characterize the transient reactive α-λ(3)-iodine alkyl acetophenone intermediate in the reaction solution. The trivalent iodine species was detected when PhI and m-CPBA in acetic acid were mixed, which indicated the facile oxidation of a catalytic amount of PhI(I) to the iodine(III) species by m-CPBA. Most importantly, 3·H(+) was observed at m/z 383 from the reaction solution and this ion gave the protonated α-acetoxylation product 4·H(+) at m/z 179 in MS/MS by an intramolecular reductive elimination of PhI. These ESI-MS/MS studies showed the existence of the reactive α-λ(3)-iodine alkyl acetophenone intermediate 3 in the catalytic cycle. Moreover, the gas-phase reactivity of 3·H(+) was consistent with the proposed solution-phase reactivity of the α-λ(3)-iodine alkyl acetophenone intermediate 3, thus confirming the reaction mechanism proposed by Ochiai. Copyright © 2012 John Wiley & Sons, Ltd.

  15. Cerebrovascular endothelin-1 hyper-reactivity is associated with transient receptor potential canonical channels 1 and 6 activation and delayed cerebral hypoperfusion after forebrain ischaemia in rats

    DEFF Research Database (Denmark)

    Johansson, S E; Andersen, X E D R; Hansen, R H

    2015-01-01

    . METHODS: Experimental forebrain ischaemia was induced in Wistar male rats by a two-vessel occlusion model, and the cerebral blood flow was measured by magnetic resonance imaging two days after reperfusion. In vitro vasoreactivity studies, immunofluorescence and quantitative PCR were performed on cerebral...... in the vascular smooth muscle cells was enhanced and correlated with decreased cerebral blood flow two days after forebrain ischaemia. Furthermore, under conditions when voltage-dependent calcium channels were inhibited, endothelin-1-induced cerebrovascular contraction was enhanced and this enhancement...... was presumably mediated by Ca(2+) influx via upregulated transient receptor potential canonical channels 1 and 6. CONCLUSIONS: Our data demonstrates that endothelin-1-mediated influx of extracellular Ca(2+) activates transient receptor potential canonical channels 1 and 6 in cerebral vascular smooth muscle cells...

  16. Photogeneration of reactive transient species upon irradiation of natural water samples: Formation quantum yields in different spectral intervals, and implications for the photochemistry of surface waters.

    Science.gov (United States)

    Marchisio, Andrea; Minella, Marco; Maurino, Valter; Minero, Claudio; Vione, Davide

    2015-04-15

    Chromophoric dissolved organic matter (CDOM) in surface waters is a photochemical source of several transient species such as CDOM triplet states ((3)CDOM*), singlet oxygen ((1)O2) and the hydroxyl radical (OH). By irradiation of lake water samples, it is shown here that the quantum yields for the formation of these transients by CDOM vary depending on the irradiation wavelength range, in the order UVB > UVA > blue. A possible explanation is that radiation at longer wavelengths is preferentially absorbed by the larger CDOM fractions, which show lesser photoactivity compared to smaller CDOM moieties. The quantum yield variations in different spectral ranges were definitely more marked for (3)CDOM* and OH compared to (1)O2. The decrease of the quantum yields with increasing wavelength has important implications for the photochemistry of surface waters, because long-wavelength radiation penetrates deeper in water columns compared to short-wavelength radiation. The average steady-state concentrations of the transients ((3)CDOM*, (1)O2 and OH) were modelled in water columns of different depths, based on the experimentally determined wavelength trends of the formation quantum yields. Important differences were found between such modelling results and those obtained in a wavelength-independent quantum yield scenario. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Huang, Pinghue

    1998-01-01

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

  18. What makes ecological systems reactive?

    Science.gov (United States)

    Snyder, Robin E

    2010-06-01

    Although perturbations from a stable equilibrium must ultimately vanish, they can grow initially, and the maximum initial growth rate is called reactivity. Reactivity thus identifies systems that may undergo transient population surges or drops in response to perturbations; however, we lack biological and mathematical intuition about what makes a system reactive. This paper presents upper and lower bounds on reactivity for an arbitrary linearized model, explores their strictness, and discusses their biological implications. I find that less stable systems (i.e. systems with long transients) have a smaller possible range of reactivities for which no perturbations grow. Systems with more species have a higher capacity to be reactive, assuming species interactions do not weaken too rapidly as the number of species increases. Finally, I find that in discrete time, reactivity is determined largely by mean interaction strength and neither discrete nor continuous time reactivity are sensitive to food web topology. 2010 Elsevier Inc. All rights reserved.

  19. Model for calculating the boron concentration in PWR type reactors

    International Nuclear Information System (INIS)

    Reis Martins Junior, L.L. dos; Vanni, E.A.

    1986-01-01

    A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

  20. Prevalence of Ex Vivo High On-treatment Platelet Reactivity on Antiplatelet Therapy after Transient Ischemic Attack or Ischemic Stroke on the PFA-100(®) and VerifyNow(®).

    LENUS (Irish Health Repository)

    Kinsella, Justin A

    2012-09-12

    BACKGROUND: The prevalence of ex vivo high on-treatment platelet reactivity (HTPR) to commonly prescribed antiplatelet regimens after transient ischemic attack (TIA) or ischemic stroke is uncertain. METHODS: Platelet function inhibition was simultaneously assessed with modified light transmission aggregometry (VerifyNow; Accumetrics Inc, San Diego, CA) and with a moderately high shear stress platelet function analyzer (PFA-100; Siemens Medical Solutions USA, Inc, Malvern, PA) in a pilot, cross-sectional study of TIA or ischemic stroke patients. Patients were assessed on aspirin-dipyridamole combination therapy (n = 51) or clopidogrel monotherapy (n = 25). RESULTS: On the VerifyNow, HTPR on aspirin was identified in 4 of 51 patients (8%) on aspirin-dipyridamole combination therapy (≥550 aspirin reaction units on the aspirin cartridge). Eleven of 25 (44%) patients had HTPR on clopidogrel (≥194 P2Y12 reaction units on the P2Y12 cartridge). On the PFA-100, 21 of 51 patients (41%) on aspirin-dipyridamole combination therapy had HTPR on the collagen-epinephrine (C-EPI) cartridge. Twenty-three of 25 patients (92%) on clopidogrel had HTPR on the collagen-adenosine diphosphate (C-ADP) cartridge. The proportion of patients with antiplatelet HTPR was lower on the VerifyNow than PFA-100 in patients on both regimens (P < .001). CONCLUSIONS: The prevalence of ex vivo antiplatelet HTPR after TIA or ischemic stroke is markedly influenced by the method used to assess platelet reactivity. The PFA-100 C-ADP cartridge is not sensitive at detecting the antiplatelet effects of clopidogrel ex vivo. Larger prospective studies with the VerifyNow and with the PFA-100 C-EPI and recently released Innovance PFA P2Y cartridges (Siemens Medical Solutions USA, Inc) in addition to newer tests of platelet function are warranted to assess whether platelet function monitoring predicts clinical outcome in ischemic cerebrovascular disease.

  1. Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Tzang, Y.-C., E-mail: yctzang@aec.gov.t [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China); Chiang, R.-F.; Ferng, Y.-M.; Pei, B.-S. [National Tsing Hua University, Department of Engineering and System Science, Hsinchu 30013, Taiwan (China)

    2009-12-15

    In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the DELTACPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.

  2. Transient analyzer

    International Nuclear Information System (INIS)

    Muir, M.D.

    1975-01-01

    The design and design philosophy of a high performance, extremely versatile transient analyzer is described. This sub-system was designed to be controlled through the data acquisition computer system which allows hands off operation. Thus it may be placed on the experiment side of the high voltage safety break between the experimental device and the control room. This analyzer provides control features which are extremely useful for data acquisition from PPPL diagnostics. These include dynamic sample rate changing, which may be intermixed with multiple post trigger operations with variable length blocks using normal, peak to peak or integrate modes. Included in the discussion are general remarks on the advantages of adding intelligence to transient analyzers, a detailed description of the characteristics of the PPPL transient analyzer, a description of the hardware, firmware, control language and operation of the PPPL transient analyzer, and general remarks on future trends in this type of instrumentation both at PPPL and in general

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Chernobyl reactor transient simulation study

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  5. Transient analysis for resolving safety issues

    International Nuclear Information System (INIS)

    Chao, J.; Layman, W.

    1987-01-01

    The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

  6. Analysis of steady state and transient two-phase flows in downwardly inclined lines

    International Nuclear Information System (INIS)

    Crawford, T.J.

    1983-01-01

    A study of steady-state and transient two-phase flows in downwardly inclined lines is described. Steady-state flow patterns maps are presented using Freon-113 as the working fluid to provide new high density vapors. These flow maps with high density vapor serve to significantly extend the investigations of steady-state downward two-phase flow patterns. Physical models developed which successfully predicted the onset or location of various flow pattern transitions. A new simplified criterion that would be useful to designers and experimenters is offered for the onset of dispersed flow. A new empirical holdup correlation and a new bubble diameter/flow rate correlation are also proposed. Flow transients in vertical downward lines were studied to investigate the possible formation of intermediate or spurious flow patterns that would not be seen at steady-state conditions. Void fraction behavior during the transients was modeled by using the dynamic slip equation from the transient analysis code RETRAN. Physical models of interfacial area were developed and compared with models and data from literature. There was satisfactory agreement between the models of the present study and the literature models and data. The concentration parameter of the drift flux model was evaluated for vertical downward flow. These new values of the flow dependent parameter were different from those previously proposed in the literature for use in upward flows, and made the drift flux model suitable for use in upward or downward flow lines

  7. Development of the MARS input model for Ulchin 1/2 transient analyzer

    International Nuclear Information System (INIS)

    Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Chung, B. D.; Hwang, M.

    2003-03-01

    KAERI has been developing the NSSS transient analyzer based on best-estimate codes for Ulchin 1/2 plants. The MARS and RETRAN code are used as the best-estimate codes for the NSSS transient analyzer. Among the two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the input model requirements and the calculation note for the Ulchin 1/2 MARS input data generation (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 1/2

  8. Development of the MARS input model for Ulchin 3/4 transient analyzer

    International Nuclear Information System (INIS)

    Jeong, J. J.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Hwang, M. G.

    2003-12-01

    KAERI has been developing the NSSS transient analyzer based on best-estimate codes.The MARS and RETRAN code are adopted as the best-estimate codes for the NSSS transient analyzer. Among these two codes, the MARS code is to be used for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. This report includes the MARS input model requirements and the calculation note for the MARS input data generation (see the Appendix) for Ulchin 3/4 plant analyzer. In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Ulchin 3/4

  9. Development of the MARS input model for Kori nuclear units 1 transient analyzer

    International Nuclear Information System (INIS)

    Hwang, M.; Kim, K. D.; Lee, S. W.; Lee, Y. J.; Lee, W. J.; Chung, B. D.; Jeong, J. J.

    2004-11-01

    KAERI has been developing the 'NSSS transient analyzer' based on best-estimate codes for Kori Nuclear Units 1 plants. The MARS and RETRAN codes have been used as the best-estimate codes for the NSSS transient analyzer. Among these codes, the MARS code is adopted for realistic analysis of small- and large-break loss-of-coolant accidents, of which break size is greater than 2 inch diameter. So it is necessary to develop the MARS input model for Kori Nuclear Units 1 plants. This report includes the input model (hydrodynamic component and heat structure models) requirements and the calculation note for the MARS input data generation for Kori Nuclear Units 1 plant analyzer (see the Appendix). In order to confirm the validity of the input data, we performed the calculations for a steady state at 100 % power operation condition and a double-ended cold leg break LOCA. The results of the steady-state calculation agree well with the design data. The results of the LOCA calculation seem to be reasonable and consistent with those of other best-estimate calculations. Therefore, the MARS input data can be used as a base input deck for the MARS transient analyzer for Kori Nuclear Units 1

  10. Transient analysis models for nuclear power plants

    International Nuclear Information System (INIS)

    Agapito, J.R.

    1981-01-01

    The modelling used for the simulation of the Angra-1 start-up reactor tests, using the RETRAN computer code is presented. Three tests are simulated: a)nuclear power plant trip from 100% of power; b)great power excursions tests and c)'load swing' tests.(E.G.) [pt

  11. Transient pseudohypoaldosteronism

    Directory of Open Access Journals (Sweden)

    Stajić Nataša

    2011-01-01

    Full Text Available Introduction. Infants with urinary tract malformations (UTM presenting with urinary tract infection (UTI are prone to develop transient type 1 pseudohypoaldosteronism (THPA1. Objective. Report on patient series with characteristics of THPA1, UTM and/or UTI and suggestions for the diagnosis and therapy. Methods. Patients underwent blood and urine electrolyte and acid-base analysis, serum aldosterosterone levels and plasma rennin activity measuring; urinalysis, urinoculture and renal ultrasound were done and medical and/or surgical therapy was instituted. Results. Hyponatraemia (120.9±5.8 mmol/L, hyperkalaemia (6.9±0.9 mmol/L, metabolic acidosis (plasma bicarbonate, 11±1.4 mmol/L, and a rise in serum creatinine levels (145±101 μmol/L were associated with inappropriately high urinary sodium (51.3±17.5 mmol/L and low potassium (14.1±5.9 mmol/L excretion. Elevated plasma aldosterone concentrations (170.4±100.5 ng/dL and the very high levels of the plasma aldosterone to potassium ratio (25.2±15.6 together with diminished urinary K/Na values (0.31±0.19 indicated tubular resistance to aldosterone. After institution of appropriate medical and/or surgical therapy, serum electrolytes, creatinine, and acid-base balance were normalized. Imaging studies showed ureteropyelic or ureterovesical junction obstruction in 3 and 2 patients, respectively, posterior urethral valves in 3, and normal UT in 1 patient. According to our knowledge, this is the first report on THPA1 in the Serbian literature. Conclusion. Male infants with hyponatraemia, hyperkalaemia and metabolic acidosis have to have their urine examined and the renal ultrasound has to be done in order to avoid both, the underdiagnosis of THPA1 and the inappropriate medication.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  14. Additional 5 kWe thermoelectric system temperature transients

    International Nuclear Information System (INIS)

    Halfen, F.J.

    1972-01-01

    Several additional system transients have been calculated for the 5 kW(e) TE system and are reported in this document. They include a startup transient with a reactivity rate of 0.005 cents/sec, several startup accidents, a step reactivity insertion at full power and a loss of electrical load. These data are intended for input to system design analyses and for possible use in the protected accident section of the safety report. (U.S.)

  15. Current interruption transients calculation

    CERN Document Server

    Peelo, David F

    2014-01-01

    Provides an original, detailed and practical description of current interruption transients, origins, and the circuits involved, and how they can be calculated Current Interruption Transients Calculationis a comprehensive resource for the understanding, calculation and analysis of the transient recovery voltages (TRVs) and related re-ignition or re-striking transients associated with fault current interruption and the switching of inductive and capacitive load currents in circuits. This book provides an original, detailed and practical description of current interruption transients, origins,

  16. Design criteria of integrated reactors based on transients

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. Sensitivity analysis of reactive ecological dynamics.

    Science.gov (United States)

    Verdy, Ariane; Caswell, Hal

    2008-08-01

    Ecological systems with asymptotically stable equilibria may exhibit significant transient dynamics following perturbations. In some cases, these transient dynamics include the possibility of excursions away from the equilibrium before the eventual return; systems that exhibit such amplification of perturbations are called reactive. Reactivity is a common property of ecological systems, and the amplification can be large and long-lasting. The transient response of a reactive ecosystem depends on the parameters of the underlying model. To investigate this dependence, we develop sensitivity analyses for indices of transient dynamics (reactivity, the amplification envelope, and the optimal perturbation) in both continuous- and discrete-time models written in matrix form. The sensitivity calculations require expressions, some of them new, for the derivatives of equilibria, eigenvalues, singular values, and singular vectors, obtained using matrix calculus. Sensitivity analysis provides a quantitative framework for investigating the mechanisms leading to transient growth. We apply the methodology to a predator-prey model and a size-structured food web model. The results suggest predator-driven and prey-driven mechanisms for transient amplification resulting from multispecies interactions.

  18. Reactive Arthritis

    Directory of Open Access Journals (Sweden)

    Eren Erken

    2013-06-01

    Full Text Available Reactive arthritis is an acute, sterile, non-suppurative and inflammatory arthropaty which has occured as a result of an infectious processes, mostly after gastrointestinal and genitourinary tract infections. Reiter syndrome is a frequent type of reactive arthritis. Both reactive arthritis and Reiter syndrome belong to the group of seronegative spondyloarthropathies, associated with HLA-B27 positivity and characterized by ongoing inflammation after an infectious episode. The classical triad of Reiter syndrome is defined as arthritis, conjuctivitis and urethritis and is seen only in one third of patients with Reiter syndrome. Recently, seronegative asymmetric arthritis and typical extraarticular involvement are thought to be adequate for the diagnosis. However, there is no established criteria for the diagnosis of reactive arthritis and the number of randomized and controlled studies about the therapy is not enough. [Archives Medical Review Journal 2013; 22(3.000: 283-299

  19. Transient drainage summary report

    International Nuclear Information System (INIS)

    1996-09-01

    This report summarizes the history of transient drainage issues on the Uranium Mill Tailings Remedial Action (UMTRA) Project. It defines and describes the UMTRA Project disposal cell transient drainage process and chronicles UMTRA Project treatment of the transient drainage phenomenon. Section 4.0 includes a conceptual cross section of each UMTRA Project disposal site and summarizes design and construction information, the ground water protection strategy, and the potential for transient drainage

  20. PSH Transient Simulation Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Muljadi, Eduard [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-12-21

    PSH Transient Simulation Modeling presentation from the WPTO FY14 - FY16 Peer Review. Transient effects are an important consideration when designing a PSH system, yet numerical techniques for hydraulic transient analysis still need improvements for adjustable-speed (AS) reversible pump-turbine applications.

  1. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    International Nuclear Information System (INIS)

    Meng Lin; Rui Hu; Yun Su; Ronghua Zhang; Yanhua Yang

    2005-01-01

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  2. TRANSIENT ELECTRONICS CATEGORIZATION

    Science.gov (United States)

    2017-08-24

    AFRL-RY-WP-TR-2017-0169 TRANSIENT ELECTRONICS CATEGORIZATION Dr. Burhan Bayraktaroglu Devices for Sensing Branch Aerospace Components & Subsystems...SUBTITLE TRANSIENT ELECTRONICS CATEGORIZATION 5a. CONTRACT NUMBER In-house 5b. GRANT NUMBER N/A 5c. PROGRAM ELEMENT NUMBER N/A 6. AUTHOR(S) Dr. Burhan...88ABW-2017-3747, Clearance Date 31 July 2017. Paper contains color. 14. ABSTRACT Transient electronics is an emerging technology area that lacks proper

  3. Positive void reactivity

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1992-09-01

    This report is a review of some of the important aspects of the analysis of large loss-of-coolant accidents (LOCAs). One important aspect is the calculation of positive void reactivity. To study this subject the lattice physics codes used for void worth calculations and the coupled neutronic and thermal-hydraulic codes used for the transient analysis are reviewed. Also reviewed are the measurements used to help validate the codes. The application of these codes to large LOCAs is studied with attention focused on the uncertainty factor for the void worth used to bias the results. Another aspect of the subject dealt with in the report is the acceptance criteria that are applied. This includes the criterion for peak fuel enthalpy and the question of whether prompt criticality should also be a criterion. To study the former, fuel behavior measurements and calculations are reviewed. (Author) (49 refs., 2 figs., tab.)

  4. Reactive Systems

    DEFF Research Database (Denmark)

    Aceto, Luca; Ingolfsdottir, Anna; Larsen, Kim Guldstrand

    A reactive system comprises networks of computing components, achieving their goals through interaction among themselves and their environment. Thus even relatively small systems may exhibit unexpectedly complex behaviours. As moreover reactive systems are often used in safety critical systems......, the need for mathematically based formal methodology is increasingly important. There are many books that look at particular methodologies for such systems. This book offers a more balanced introduction for graduate students and describes the various approaches, their strengths and weaknesses, and when...... they are best used. Milner's CCS and its operational semantics are introduced, together with the notions of behavioural equivalences based on bisimulation techniques and with recursive extensions of Hennessy-Milner logic. In the second part of the book, the presented theories are extended to take timing issues...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  6. Spectroscopic classification of transients

    DEFF Research Database (Denmark)

    Stritzinger, M. D.; Fraser, M.; Hummelmose, N. N.

    2017-01-01

    We report the spectroscopic classification of several transients based on observations taken with the Nordic Optical Telescope (NOT) equipped with ALFOSC, over the nights 23-25 August 2017.......We report the spectroscopic classification of several transients based on observations taken with the Nordic Optical Telescope (NOT) equipped with ALFOSC, over the nights 23-25 August 2017....

  7. The transient analysis of single turbine control valve closure for Lungmen ABWR

    International Nuclear Information System (INIS)

    Ma Shaoshih; Yuann Yngruey; Shih Chunkuan

    2012-01-01

    Highlights: ► The LRM was used to evaluate the single control valve closure event. ► The purpose is to offer an updated analysis about the MCFL under the partial arc mode instead of FSAR’s result. ► It is concluded that the 112% MCFL setting is the most limiting case. ► The MCFL setting actually used in SBPCS must be kept between 112% to 114% to gain the operational margin. ► The HFF index defined by the normalized heat flux can be used to predict the CPR change. - Abstract: The single control valve closure in fast (SCVCF) event is the most limiting transient in terms of delta critical power ratio (ΔCPR) for the Lungmen Plant, which is a basis to determine the operating limit minimum critical power ratio value. The partial arc mode is adopted in Lungmen Plant to control the position of the turbine control valve. However, the transient analyses presented in the Lungmen Final Safety Analysis Report (FSAR) assume that the TCVs are in the full arc mode. In this study, the Lungmen RETRAM model with partial arc mode is used to analyze the SCVCF event to offer more realistic results than the FSAR. It is concluded that the most limiting maximum combined flow limiter (MCFL) setting in RETRAN analysis is different from that of FSAR. An optimum operating range for the MCFL is suggested to gain the margin against the operating drift. Additionally, a Heat Flux Factor index is defined to appropriately determine the ranking of these cases in terms of ΔCPR.

  8. Modelling of power-reactivity coefficient measurement

    International Nuclear Information System (INIS)

    Strmensky, C.; Petenyi, V.; Jagrik, J.; Minarcin, M.; Hascik, R.; Toth, L.

    2005-01-01

    Report describes results of modeling of power-reactivity coefficient analysis on power-level. In paper we calculate values of discrepancies arisen during transient process. These discrepancies can be arisen as result of experiment evaluation and can be caused by disregard of 3D effects on neutron distribution. The results are critically discussed (Authors)

  9. Advanced Instrumentation for Transient Reactor Testing

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-01-31

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

  10. Pius, self-protective thermohydraulics transient without safety system intervention

    International Nuclear Information System (INIS)

    Fredell, J.; Bredolt, V.

    1989-01-01

    In this paper, the self-protective thermohydraulic feedback of the PIUS reactor system is illustrated by an in-depth discussion of one typical transient. The selected transient is an undetected total loss of feedwater in the complete absence of conventional safety system intervention. The reactor shuts itself down to residual power in two steps. First, the power decreases due to the strongly negative moderator temperature reactivity coefficient, and then a complete shutdown occurs by ingress of cold, highly borated water from the reactor pool. The transient is terminated without any harm to the fuel or paint systems

  11. 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

  12. Transients: The regulator's view

    International Nuclear Information System (INIS)

    Sheron, B.W.; Speis, T.P.

    1984-01-01

    This chapter attempts to clarify the basis for the regulator's concerns for transient events. Transients are defined as both anticipated operational occurrences and postulated accidents. Recent operational experience, supplemented by improved probabilistic risk analysis methods, has demonstrated that non-LOCA transient events can be significant contributors to overall risk. Topics considered include lessons learned from events and issues, the regulations governing plant transients, multiple failures, different failure frequencies, operator errors, and public pressure. It is concluded that the formation of Owners Groups and Regulatory Response Groups within the owners groups are positive signs of the industry's concern for safety and responsible dealing with the issues affecting both the US NRC and the industry

  13. Transient multivariable sensor evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, Richard B.; Heifetz, Alexander

    2017-02-21

    A method and system for performing transient multivariable sensor evaluation. The method and system includes a computer system for identifying a model form, providing training measurement data, generating a basis vector, monitoring system data from sensor, loading the system data in a non-transient memory, performing an estimation to provide desired data and comparing the system data to the desired data and outputting an alarm for a defective sensor.

  14. Predictable anomalies of process parameters on failure mode of internal structures in RPV by transient thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Maki, Akira; Mori, Michitsugu; Kanemoto, Shigeru; Konishi, Hideo

    1997-01-01

    A study has been conducted to evaluate how process parameters will exhibit the change in the event of the troubles related to reactor internal by using transient thermal-hydraulic analysis codes (RETRAN3D-MOD002, etc.). In the present study, the following six events are analytically investigated: 1) a leak from the upper plenum; 2) a leak from the middle part of a shroud; 3) a leak from the lower plenum; 4) a leak from the riser pipe for the jet-pump; 5) the blockage of the jet-pump nozzle; and 6) a leak from the jet-pump diffuser. The results by analyses indicated that the leak from the upper plenum resulted in increasing in the inlet temperature of primary loop recirculation (PLR) and in the differential pressure at the core support plate, and decreasing in the neutron flux (reactor power). Similar analyses were made for the five other events to identify the pattern of relevant process parameter variation in each event. (author)

  15. Transient flow combustion

    Science.gov (United States)

    Tacina, R. R.

    1984-01-01

    Non-steady combustion problems can result from engine sources such as accelerations, decelerations, nozzle adjustments, augmentor ignition, and air perturbations into and out of the compressor. Also non-steady combustion can be generated internally from combustion instability or self-induced oscillations. A premixed-prevaporized combustor would be particularly sensitive to flow transients because of its susceptability to flashback-autoignition and blowout. An experimental program, the Transient Flow Combustion Study is in progress to study the effects of air and fuel flow transients on a premixed-prevaporized combustor. Preliminary tests performed at an inlet air temperature of 600 K, a reference velocity of 30 m/s, and a pressure of 700 kPa. The airflow was reduced to 1/3 of its original value in a 40 ms ramp before flashback occurred. Ramping the airflow up has shown that blowout is more sensitive than flashback to flow transients. Blowout occurred with a 25 percent increase in airflow (at a constant fuel-air ratio) in a 20 ms ramp. Combustion resonance was found at some conditions and may be important in determining the effects of flow transients.

  16. Experimental validation of Pu-Sm evolution model for CANDU-6 power transients

    International Nuclear Information System (INIS)

    Coutsiers, Eduardo E.; Pomerantz, Marcelo E.; Moreno, Carlos A.

    2000-01-01

    Development of a methodology to evaluate the reactivity produced by Pu-Sm transient, effect displayed after power transients. This methodology allows to predict the behavior of liquid zones with which the fine control of CANDU reactor power is made. With this information, it is easier to foresee the refueling demand after power movements. The comparison with experimental results showed good agreement. (author)

  17. Analysis of transients in the SRP test pile

    International Nuclear Information System (INIS)

    Church, J.P.

    1976-11-01

    Analysis of the hypothetical upper limit accident in the Savannah River Test Pile showed that the offsite thyroid dose from fission product release would be -3 of the 10-CFR-100 guideline dose for 95 percent of measured meteorological conditions. Offsite whole body dose would be negligible. The Test Pile was modified to limit the length of test piece that can be charged to the pile. These modifications reduce the potential offsite dose to -5 of the regulatory guidelines. Assessment of Test Pile safety included calculations of transients initiated by a variety of reactivity additions that were either terminated or not terminated by safety systems. Reactivity addition mechanisms considered were abnormally driving control rods out of the pile and charging abnormal test pieces into the pile. The transients were evaluated in the adiabatic approximation in which three-dimensional calculations of static flux shapes and reactivity were superimposed on point reactor kinetics calculations. Negative reactivity feedback effects appropriate for the pile and the temperature dependence of material properties, such as specific heat and thermal conductivity, were included. The results show that, for the worst initiators, safety systems can prevent the temperature rise from exceeding 1 0 C anywhere in the Test Pile. If the safety systems do not function, the pile temperatures will increase until the transient is ended by the inherent negative reactivity effects, including the melting of some fuel

  18. A technique for computing bowing reactivity feedback in LMFBR's

    International Nuclear Information System (INIS)

    Finck, P.J.

    1987-01-01

    During normal or accidental transients occurring in a LMFBR core, the assemblies and their support structure are subjected to important thermal gradients which induce differential thermal expansions of the walls of the hexcans and differential displacement of the assembly support structure. These displacements, combined with the creep and swelling of structural materials, remain quite small, but the resulting reactivity changes constitute a significant component of the reactivity feedback coefficients used in safety analyses. It would be prohibitive to compute the reactivity changes due to all transients. Thus, the usual practice is to generate reactivity gradient tables. The purpose of the work presented here is twofold: develop and validate an efficient and accurate scheme for computing these reactivity tables; and to qualify this scheme

  19. Transient hardened power FETs

    International Nuclear Information System (INIS)

    Dawes, W.R. Jr.; Fischer, T.A.; Huang, C.C.C.; Meyer, W.J.; Smith, C.S.; Blanchard, R.A.; Fortier, T.J.

    1986-01-01

    N-channel power FETs offer significant advantages in power conditioning circuits. Similiarily to all MOS technologies, power FET devices are vulnerable to ionizing radiation, and are particularily susceptible to burn-out in high dose rate irradiations (>1E10 rads(Si)/sec.), which precludes their use in many military environments. This paper will summarize the physical mechanisms responsible for burn-out, and discuss various fabrication techniques designed to improve the transient hardness of power FETs. Power FET devices were fabricated with several of these techniques, and data will be presented which demonstrates that transient hardness levels in excess of 1E12 rads(Si)/sec. are easily achievable

  20. Transients in the Vivitron

    International Nuclear Information System (INIS)

    Cooke, C.M.; Frick, G.; Roumie, M.

    1993-01-01

    Electrical measurements are presented for the construction of a model for the study of transients in the Vivitron. Observation of the transmission of electrical pulses in the porticos clearly shows transmission-line behaviour. Measurements of the vector impedance of the outer porticos show the same transmission-line properties, but also gives a description of the modification from a pure transmission line due to the circular electrodes. The results of this investigation should allow the construction of a computer model which predicts the evolution of the transients in the case of a spark in the Vivitron. (orig.)

  1. Two-frequency, one-detector reactivity system (TFODRS)

    International Nuclear Information System (INIS)

    Sachs, R.D.; Woodall, D.M.

    1985-01-01

    A two-frequency, one-detector reactivity system (TFODRS) was experimentally verified on the University of New Mexico (UNM) AGN-201M thermal reactor. That system was used to obtain the absolute steady-state reactivity and to demonstrate the feasibility of acquiring the transient reactivity. A detailed description of TFODRS hardware and software is given in this paper. The TFODRS obtains the absolute and net reactivity by computing the frequency spectrum of the reactor neutron-detector signal. The ratio of the high-frequency to the low-frequency components about an empirical break point is used to determine the reactivity. The TFODRS was successfully used to measure a known AGN-201M steady-state reactivity, with a relative error of 18%. TFODRS transient curves as a function of reactivity were shown to be different from the steady-state curves. The transient curves appear to be a function of the rate of reactivity insertion. The authors speculate that a modified TFODRS, using state-of-the-art microprocessors, could be used for fast reactors. The TFODRS is not presently a practicable reactimeter. However, with more research and development, it is felt it could be used in near-term nuclear industry applications, such as monitoring fuel storage pools

  2. The reactivity meter and core reactivity

    International Nuclear Information System (INIS)

    Siltanen, P.

    1999-01-01

    This paper discussed in depth the point kinetic equations and the characteristics of the point kinetic reactivity meter, particularly for large negative reactivities. From a given input signal representing the neutron flux seen by a detector, the meter computes a value of reactivity in dollars (ρ/β), based on inverse point kinetics. The prompt jump point of view is emphasised. (Author)

  3. Transient Heat Conduction

    DEFF Research Database (Denmark)

    Rode, Carsten

    1998-01-01

    Analytical theory of transient heat conduction.Fourier's law. General heat conducation equation. Thermal diffusivity. Biot and Fourier numbers. Lumped analysis and time constant. Semi-infinite body: fixed surface temperature, convective heat transfer at the surface, or constant surface heat flux...

  4. Transient cavitation in pipelines

    NARCIS (Netherlands)

    Kranenburg, C.

    1974-01-01

    The aim of the present study is to set up a one-dimensional mathematical model, which describes the transient flow in pipelines, taking into account the influence of cavitation and free gas. The flow will be conceived of as a three-phase flow of the liquid, its vapour and non-condensible gas. The

  5. Compressive Transient Imaging

    KAUST Repository

    Sun, Qilin

    2017-04-01

    High resolution transient/3D imaging technology is of high interest in both scientific research and commercial application. Nowadays, all of the transient imaging methods suffer from low resolution or time consuming mechanical scanning. We proposed a new method based on TCSPC and Compressive Sensing to achieve a high resolution transient imaging with a several seconds capturing process. Picosecond laser sends a serious of equal interval pulse while synchronized SPAD camera\\'s detecting gate window has a precise phase delay at each cycle. After capturing enough points, we are able to make up a whole signal. By inserting a DMD device into the system, we are able to modulate all the frames of data using binary random patterns to reconstruct a super resolution transient/3D image later. Because the low fill factor of SPAD sensor will make a compressive sensing scenario ill-conditioned, We designed and fabricated a diffractive microlens array. We proposed a new CS reconstruction algorithm which is able to denoise at the same time for the measurements suffering from Poisson noise. Instead of a single SPAD senor, we chose a SPAD array because it can drastically reduce the requirement for the number of measurements and its reconstruction time. Further more, it not easy to reconstruct a high resolution image with only one single sensor while for an array, it just needs to reconstruct small patches and a few measurements. In this thesis, we evaluated the reconstruction methods using both clean measurements and the version corrupted by Poisson noise. The results show how the integration over the layers influence the image quality and our algorithm works well while the measurements suffer from non-trival Poisson noise. It\\'s a breakthrough in the areas of both transient imaging and compressive sensing.

  6. Heat shock and herpes virus: enhanced reactivation without untargeted mutagenesis

    International Nuclear Information System (INIS)

    Lytle, C.D.; Carney, P.G.

    1988-01-01

    Enhanced reactivation of Ultraviolet-irradiated virus has been reported to occur in heat-shocked host cells. Since enhanced virus reactivation is often accompanied by untargeted mutagenesis, we investigated whether such mutagenesis would occur for herpes simplex virus (HSV) in CV-1 monkey kidney cells subjected to heat shock. In addition to expressing enhanced reactivation, the treated cells were transiently more susceptible to infection by unirradiated HSV. No mutagenesis of unirradiated HSV was found whether infection occurred at the time of increased susceptibility to infection or during expression of enhanced viral reactivation

  7. Actinide and Xenon reactivity effects in ATW high flux systems

    International Nuclear Information System (INIS)

    Woosley, M.; Olson, K.; Henderson, D.L.

    1995-01-01

    In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

  8. Actinide and xenon reactivity effects in ATW high flux systems

    International Nuclear Information System (INIS)

    Woosley, M.; Olson, K.; Henderson, D. L.; Sailor, W. C.

    1995-01-01

    In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides

  9. Actinide and Xenon reactivity effects in ATW high flux systems

    Energy Technology Data Exchange (ETDEWEB)

    Woosley, M. [Univ. of Virginia, Charlottesville, VA (United States); Olson, K.; Henderson, D.L. [Univ. of Wisconsin, Madison, WI (United States)] [and others

    1995-10-01

    In this paper, initial system reactivity response to flux changes caused by the actinides and xenon are investigated separately for a high flux ATW system. The maximum change in reactivity after a flux change due to the effect of the changing quantities of actinides is generally at least two orders of magnitude smaller than either the positive or negative reactivity effect associated with xenon after a shutdown or start-up. In any transient flux event, the reactivity response of the system to xenon will generally occlude the response due to the actinides.

  10. Point kinetics improvements to evaluate three-dimensional effects in transients calculation

    International Nuclear Information System (INIS)

    Castellotti, U.

    1987-01-01

    A calculation method, which considers the flux axial perturbations in the parameters related to the reactivity within a point kinetics model, is described. The method considered uses axial factors of consideration which act on the thermohydraulic variables included in the reactivity calculation. The PUMA three-dimensional code as reference model for the comparisons, is used. The limitations inherent to the reactivity balance of the point models used in the transients calculation, are given. (Author)

  11. Results of LWR core transient benchmarks

    International Nuclear Information System (INIS)

    Finnemann, H.; Bauer, H.; Galati, A.; Martinelli, R.

    1993-10-01

    LWR core transient (LWRCT) benchmarks, based on well defined problems with a complete set of input data, are used to assess the discrepancies between three-dimensional space-time kinetics codes in transient calculations. The PWR problem chosen is the ejection of a control assembly from an initially critical core at hot zero power or at full power, each for three different geometrical configurations. The set of problems offers a variety of reactivity excursions which efficiently test the coupled neutronic/thermal - hydraulic models of the codes. The 63 sets of submitted solutions are analyzed by comparison with a nodal reference solution defined by using a finer spatial and temporal resolution than in standard calculations. The BWR problems considered are reactivity excursions caused by cold water injection and pressurization events. In the present paper, only the cold water injection event is discussed and evaluated in some detail. Lacking a reference solution the evaluation of the 8 sets of BWR contributions relies on a synthetic comparative discussion. The results of this first phase of LWRCT benchmark calculations are quite satisfactory, though there remain some unresolved issues. It is therefore concluded that even more challenging problems can be successfully tackled in a suggested second test phase. (authors). 46 figs., 21 tabs., 3 refs

  12. Quadratic reactivity fuel cycle model

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1985-01-01

    For educational purposes it is highly desirable to provide simple yet realistic models for fuel cycle and fuel economy. In particular, a lumped model without recourse to detailed spatial calculations would be very helpful in providing the student with a proper understanding of the purposes of fuel cycle calculations. A teaching model for fuel cycle studies based on a lumped model assuming the summability of partial reactivities with a linear dependence of reactivity usefully illustrates fuel utilization concepts. The linear burnup model does not satisfactorily represent natural enrichment reactors. A better model, showing the trend of initial plutonium production before subsequent fuel burnup and fission product generation, is a quadratic fit. The study of M-batch cycles, reloading 1/Mth of the core at end of cycle, is now complicated by nonlinear equations. A complete account of the asymptotic cycle for any order of M-batch refueling can be given and compared with the linear model. A complete account of the transient cycle can be obtained readily in the two-batch model and this exact solution would be useful in verifying numerical marching models. It is convenient to treat the parabolic fit rho = 1 - tau 2 as a special case of the general quadratic fit rho = 1 - C/sub tau/ - (1 - C)tau 2 in suitably normalized reactivity and cycle time units. The parabolic results are given in this paper

  13. Transient osteoporosis of hip

    Directory of Open Access Journals (Sweden)

    Mahesh M Choudhary

    2015-01-01

    Full Text Available We report a case of transient osteoporosis of the hip (TOH in a 50-year-old man including the clinical presentation, diagnostic studies, management, and clinical progress. TOH is a rare self-limiting condition that typically affects middle-aged men or, less frequently, women in the third trimester of pregnancy. Affected individuals present clinically with acute hip pain, limping gait, and limited ranges of hip motion. TOH may begin spontaneously or after a minor trauma. Radiographs are typically unremarkable but magnetic resonance (MR imaging studies yield findings consistent with bone marrow edema. TOH is referred to as regional migratory osteoporosis (RMO if it travels to other joints or the contralateral hip. TOH often resembles osteonecrosis but the two conditions must be differentiated due to different prognoses and management approaches. The term TOH is often used interchangeably and synonymously with transient bone marrow edema (TBME.

  14. Stability of Ignition Transients

    OpenAIRE

    V.E. Zarko

    1991-01-01

    The problem of ignition stability arises in the case of the action of intense external heat stimuli when, resulting from the cut-off of solid substance heating, momentary ignition is followed by extinction. Physical pattern of solid propellant ignition is considered and ignition criteria available in the literature are discussed. It is shown that the above mentioned problem amounts to transient burning at a given arbitrary temperature distribution in the condensed phase. A brief survey...

  15. Transient FDTD simulation validation

    OpenAIRE

    Jauregui Tellería, Ricardo; Riu Costa, Pere Joan; Silva Martínez, Fernando

    2010-01-01

    In computational electromagnetic simulations, most validation methods have been developed until now to be used in the frequency domain. However, the EMC analysis of the systems in the frequency domain many times is not enough to evaluate the immunity of current communication devices. Based on several studies, in this paper we propose an alternative method of validation of the transients in time domain allowing a rapid and objective quantification of the simulations results.

  16. MHD aspects of coronal transients

    International Nuclear Information System (INIS)

    Anzer, U.

    1979-10-01

    If one defines coronal transients as events which occur in the solar corona on rapid time scales (< approx. several hours) then one would have to include a large variety of solar phenomena: flares, sprays, erupting prominences, X-ray transients, white light transients, etc. Here we shall focus our attention on the latter two phenomena. (orig.) 891 WL/orig. 892 RDG

  17. Void effects on BWR Doppler and void reactivity feedback

    International Nuclear Information System (INIS)

    Hsiang-Shou Cheng; Diamond, D.J.

    1978-01-01

    The significance of steam voids and control rods on the Doppler feedback in a gadolinia shimmed BWR is demonstrated. The importance of bypass voids when determining void feedback is also shown. Calculations were done using a point model, i.e., feedback was expressed in terms of reactivity coefficients which were determined for individual four-bundle configurations and then appropriately combined to yield reactor results. For overpower transients the inclusion of the void effect of control rods is to reduce Doppler feedback. For overpressurization transients the inclusion of the effect of bypass void wil increase the reactivity due to void collapse. (author)

  18. FFTF fuel pin design procedure verification for transient operation

    International Nuclear Information System (INIS)

    Baars, R.E.

    1975-05-01

    The FFTF design procedures for evaluating fuel pin transient performance are briefly reviewed, and data where available are compared with design procedure predictions. Specifically, burst conditions derived from Fuel Cladding Transient Tester (FCTT) tests and from ANL loss-of-flow tests are compared with burst pressures computed using the design procedure upon which the cladding integrity limit was based. Failure times are predicted using the design procedure for evaluation of rapid reactivity insertion accidents, for five unterminated TREAT experiments in which well characterized fuel failures were deliberately incurred. (U.S.)

  19. Experimental study and modelling of transient boiling

    International Nuclear Information System (INIS)

    Baudin, Nicolas

    2015-01-01

    A failure in the control system of the power of a nuclear reactor can lead to a Reactivity Initiated Accident in a nuclear power plant. Then, a power peak occurs in some fuel rods, high enough to lead to the coolant film boiling. It leads to an important increase of the temperature of the rod. The possible risk of the clad failure is a matter of interest for the Institut de Radioprotection et de Securite Nucleaire. The transient boiling heat transfer is not yet understood and modelled. An experimental set-up has been built at the Institut de Mecanique des Fluides de Toulouse (IMFT). Subcooled HFE-7000 flows vertically upward in a semi annulus test section. The inner half cylinder simulates the clad and is made of a stainless steel foil, heated by Joule effect. Its temperature is measured by an infrared camera, coupled with a high speed camera for the visualization of the flow topology. The whole boiling curve is studied in steady state and transient regimes: convection, onset of boiling, nucleate boiling, critical heat flux, film boiling and rewetting. The steady state heat transfers are well modelled by literature correlations. Models are suggested for the transient heat flux: the convection and nucleate boiling evolutions are self-similar during a power step. This observation allows to model more complex evolutions, as temperature ramps. The transient Hsu model well represents the onset of nucleate boiling. When the intensity of the power step increases, the film boiling begins at the same temperature but with an increasing heat flux. For power ramps, the critical heat flux decreases while the corresponding temperature increases with the heating rate. When the wall is heated, the film boiling heat transfer is higher than in steady state but it is not understood. A two-fluid model well simulates the cooling film boiling and the rewetting. (author)

  20. Reactive Kripke semantics

    CERN Document Server

    Gabbay, Dov M

    2013-01-01

    This text offers an extension to the traditional Kripke semantics for non-classical logics by adding the notion of reactivity. Reactive Kripke models change their accessibility relation as we progress in the evaluation process of formulas in the model. This feature makes the reactive Kripke semantics strictly stronger and more applicable than the traditional one. Here we investigate the properties and axiomatisations of this new and most effective semantics, and we offer a wide landscape of applications of the idea of reactivity. Applied topics include reactive automata, reactive grammars, rea

  1. Magnetic transients in flares

    International Nuclear Information System (INIS)

    Zirin, H.; Tanaka, K.

    1981-01-01

    We present data on magnetic transients (mgtr's) observed in flares on 1980 July 1 and 5 with Big Bear videomagnetograph (VMG). The 1980 July 1 event was a white light flare in which a strong bipolar mgtr was observed, and a definite change in the sunspots occurred at the time of the flare. In the 1980 July 5 flare, a mgtr was observed in only one polarity, and, although no sunspot changes occurred simultaneous with the flare, major spot changes occurred in a period of hours

  2. Familial Transient Global Amnesia

    Directory of Open Access Journals (Sweden)

    R.Rhys Davies

    2012-12-01

    Full Text Available Following an episode of typical transient global amnesia (TGA, a female patient reported similar clinical attacks in 2 maternal aunts. Prior reports of familial TGA are few, and no previous account of affected relatives more distant than siblings or parents was discovered in a literature survey. The aetiology of familial TGA is unknown. A pathophysiological mechanism akin to that in migraine attacks, comorbidity reported in a number of the examples of familial TGA, is one possibility. The study of familial TGA cases might facilitate the understanding of TGA aetiology.

  3. A fast reactor transient analysis methodology for personal computers

    International Nuclear Information System (INIS)

    Ott, K.O.

    1993-01-01

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

  4. Measurand transient signal suppressor

    Science.gov (United States)

    Bozeman, Richard J., Jr. (Inventor)

    1994-01-01

    A transient signal suppressor for use in a controls system which is adapted to respond to a change in a physical parameter whenever it crosses a predetermined threshold value in a selected direction of increasing or decreasing values with respect to the threshold value and is sustained for a selected discrete time interval is presented. The suppressor includes a sensor transducer for sensing the physical parameter and generating an electrical input signal whenever the sensed physical parameter crosses the threshold level in the selected direction. A manually operated switch is provided for adapting the suppressor to produce an output drive signal whenever the physical parameter crosses the threshold value in the selected direction of increasing or decreasing values. A time delay circuit is selectively adjustable for suppressing the transducer input signal for a preselected one of a plurality of available discrete suppression time and producing an output signal only if the input signal is sustained for a time greater than the selected suppression time. An electronic gate is coupled to receive the transducer input signal and the timer output signal and produce an output drive signal for energizing a control relay whenever the transducer input is a non-transient signal which is sustained beyond the selected time interval.

  5. Transient regional osteoporosis

    Directory of Open Access Journals (Sweden)

    F. Trotta

    2011-09-01

    Full Text Available Transient osteoporosis of the hip and regional migratory osteoporosis are uncommon and probably underdiagnosed bone diseases characterized by pain and functional limitation mainly affecting weight-bearing joints of the lower limbs. These conditions are usually self-limiting and symptoms tend to abate within a few months without sequelae. Routine laboratory investigations are unremarkable. Middle aged men and women during the last months of pregnancy or in the immediate post-partum period are principally affected. Osteopenia with preservation of articular space and transitory edema of the bone marrow provided by magnetic resonance imaging are common to these two conditions, so they are also known by the term regional transitory osteoporosis. The appearance of bone marrow edema is not specific to regional transitory osteoporosis but can be observed in several diseases, i.e. trauma, reflex sympathetic dystrophy, avascular osteonecrosis, infections, tumors from which it must be differentiated. The etiology of this condition is unknown. Pathogenesis is still debated in particular the relationship with reflex sympathetic dystrophy, with which regional transitory osteoporosis is often identified. The purpose of the present review is to remark on the relationship between transient osteoporosis of the hip and regional migratory osteoporosis with particular attention to the bone marrow edema pattern and relative differential diagnosis.

  6. Reactive perforating collagenosis

    Directory of Open Access Journals (Sweden)

    Yadav Mukesh

    2009-01-01

    Full Text Available Reactive perforating collagenosis is a rare cutaneous disorder of unknown etiology. We hereby describe a case of acquired reactive perforating collagenosis in a patient of diabetes and chronic renal failure.

  7. Preliminary analysis of typical transients in fusion driven subcritical system (FDS-I)

    International Nuclear Information System (INIS)

    Bai Yunqing; Ke Yan; Wu Yican

    2007-01-01

    The potential safety characteristic is expected as one of the advantages of fusion-driven subcritical system (FDS-I) for the transmutation and incineration of nuclear waste compared with the critical reactor. Transients of the FDS-I may occur due to the perturbation of external neutron source, the failure of functional device, and the occurrence of the uncontrolled event. As typical transient scenarios, the following cases were analyzed: unprotected plasma overpower (UPOP), unprotected loss of flow (ULOF), unprotected transient overpower (UTOP). The transient analyses for the FDS-I were performed with a coupled two-dimensional thermal-hydraulics and neutronics transient analysis code NTC2D. The negative feedback of reactivity is the interesting safety feature of FDS-I as temperature increase, due to the fuel form of the circulating particle. The present simulation results showed that the current FDS-I design has a resistance against severe transient scenarios. (author)

  8. Reactivity on the Web

    OpenAIRE

    Bailey, James; Bry, François; Eckert, Michael; Patrânjan, Paula Lavinia

    2005-01-01

    Reactivity, the ability to detect simple and composite events and respond in a timely manner, is an essential requirement in many present-day information systems. With the emergence of new, dynamic Web applications, reactivity on the Web is receiving increasing attention. Reactive Web-based systems need to detect and react not only to simple events but also to complex, real-life situations. This paper introduces XChange, a language for programming reactive behaviour on the Web,...

  9. Development and application of objective uncertainty measures for nuclear power plant transient analysis[Dissertation 3897

    Energy Technology Data Exchange (ETDEWEB)

    Vinai, P

    2007-10-15

    For the development, design and licensing of a nuclear power plant (NPP), a sound safety analysis is necessary to study the diverse physical phenomena involved in the system behaviour under operational and transient conditions. Such studies are based on detailed computer simulations. With the progresses achieved in computer technology and the greater availability of experimental and plant data, the use of best estimate codes for safety evaluations has gained increasing acceptance. The application of best estimate safety analysis has raised new problems that need to be addressed: it has become more crucial to assess as to how reliable code predictions are, especially when they need to be compared against safety limits that must not be crossed. It becomes necessary to identify and quantify the various possible sources of uncertainty that affect the reliability of the results. Currently, such uncertainty evaluations are generally based on experts' opinion. In the present research, a novel methodology based on a non-parametric statistical approach has been developed for objective quantification of best-estimate code uncertainties related to the physical models used in the code. The basis is an evaluation of the accuracy of a given physical model achieved by comparing its predictions with experimental data from an appropriate set of separate-effect tests. The differences between measurements and predictions can be considered stochastically distributed, and thus a statistical approach can be employed. The first step was the development of a procedure for investigating the dependence of a given physical model's accuracy on the experimental conditions. Each separate-effect test effectively provides a random sample of discrepancies between measurements and predictions, corresponding to a location in the state space defined by a certain number of independent system variables. As a consequence, the samples of 'errors', achieved from analysis of the entire

  10. Development and application of objective uncertainty measures for nuclear power plant transient analysis

    International Nuclear Information System (INIS)

    Vinai, P.

    2007-10-01

    associated to various individual points over the state space. By applying a novel multi-dimensional clustering technique, based on the non-parametric statistical Kruskal-Wallis test, it has been possible to achieve a partitioning of the state space into regions differing in terms of the quality of the physical model's predictions. The second step has been the quantification of the model's uncertainty, for each of the identified state space regions, by applying a probability density function (pdf) estimator. This is a kernel-type estimator, modelled on a universal orthogonal series estimator, such that its behaviour takes advantage of the good features of both estimator types and yields reasonable pdfs, even with samples of small size and not very compact distributions. The pdfs provide a reliable basis for sampling 'error values' for use in Monte-Carlo-type uncertainty propagation studies, aimed at quantifying the impact of the physical model's uncertainty on the code's output variables of interest. The effectiveness of the developed methodology was demonstrated by applying it to the quantification of the uncertainty related to thermal-hydraulic (drift-flux) models implemented in the best-estimate safety analysis code RETRAN-3D. This has been done via the usage of a wide database of void-fraction experiments for saturated and sub-cooled conditions. Appropriate pdfs were generated for quantification of the physical model's uncertainty in a 2-dimensional (pressure/mass-flux) state space, partitioned into 3 separate regions. The impact of the RETRAN-3D drift-flux model uncertainties has been assessed at three different levels of the code's application: (a) Achilles Experiment No. 2, a separate effect experiment not included in the original assessment database; (b) Omega Rod Bundle Test No. 9, an integral experiment simulating a PWR loss-of-coolant accident (LOCA); and (c) the Peach Bottom turbine trip test, a NPP (BWR) plant transient in which the void feedback mechanism plays

  11. The reactive solid-gas flow of a fluidized bed for UO2 conversion

    International Nuclear Information System (INIS)

    Juanico, L.E.

    1991-01-01

    The reactive solid-gas flow of a fluidized bed for UO 2 conversion was modeled. The sedimentation-reaction process was treated using the drift-flux equations. Also, the associated pressure transient due to the reaction gas release was analyzed. An experiment was carried out to compare the results, and pressure transient was numerically simulated, reaching interesting conclusions. (Author) [es

  12. Monadic Functional Reactive Programming

    NARCIS (Netherlands)

    A.J. van der Ploeg (Atze); C Shan

    2013-01-01

    htmlabstractFunctional Reactive Programming (FRP) is a way to program reactive systems in functional style, eliminating many of the problems that arise from imperative techniques. In this paper, we present an alternative FRP formulation that is based on the notion of a reactive computation: a

  13. The joy of transient chaos

    Energy Technology Data Exchange (ETDEWEB)

    Tél, Tamás [Institute for Theoretical Physics, Eötvös University, and MTA-ELTE Theoretical Physics Research Group, Pázmány P. s. 1/A, Budapest H-1117 (Hungary)

    2015-09-15

    We intend to show that transient chaos is a very appealing, but still not widely appreciated, subfield of nonlinear dynamics. Besides flashing its basic properties and giving a brief overview of the many applications, a few recent transient-chaos-related subjects are introduced in some detail. These include the dynamics of decision making, dispersion, and sedimentation of volcanic ash, doubly transient chaos of undriven autonomous mechanical systems, and a dynamical systems approach to energy absorption or explosion.

  14. Transient osteoporosis of the hip

    International Nuclear Information System (INIS)

    McWalter, Patricia; Hassan Ahmed

    2007-01-01

    Transient osteoporosis of the hip is an uncommon cause of hip pain, mostly affecting healthy middle-aged men and also women in the third trimester of pregnancy. We present a case of transient osteoporosis of the hip in a 33-year-old non-pregnant female patient. This case highlights the importance of considering a diagnosis of transient osteoporosis of the hip in patients who present with hip pain. (author)

  15. The ZTF Bright Transient Survey

    Science.gov (United States)

    Fremling, C.; Sharma, Y.; Kulkarni, S. R.; Miller, A. A.; Taggart, K.; Perley, D. A.; Gooba, A.

    2018-06-01

    As a supplement to the Zwicky Transient Facility (ZTF; ATel #11266) public alerts (ATel #11685) we plan to report (following ATel #11615) bright probable supernovae identified in the raw alert stream from the ZTF Northern Sky Survey ("Celestial Cinematography"; see Bellm & Kulkarni, 2017, Nature Astronomy 1, 71) to the Transient Name Server (https://wis-tns.weizmann.ac.il) on a daily basis; the ZTF Bright Transient Survey (BTS; see Kulkarni et al., 2018; arXiv:1710.04223).

  16. The joy of transient chaos.

    Science.gov (United States)

    Tél, Tamás

    2015-09-01

    We intend to show that transient chaos is a very appealing, but still not widely appreciated, subfield of nonlinear dynamics. Besides flashing its basic properties and giving a brief overview of the many applications, a few recent transient-chaos-related subjects are introduced in some detail. These include the dynamics of decision making, dispersion, and sedimentation of volcanic ash, doubly transient chaos of undriven autonomous mechanical systems, and a dynamical systems approach to energy absorption or explosion.

  17. Transient Infrared Emission Spectroscopy

    Science.gov (United States)

    Jones, Roger W.; McClelland, John F.

    1989-12-01

    Transient Infrared Emission Spectroscopy (TIRES) is a new technique that reduces the occurrence of self-absorption in optically thick solid samples so that analytically useful emission spectra may be observed. Conventional emission spectroscopy, in which the sample is held at an elevated, uniform temperature, is practical only for optically thin samples. In thick samples the emission from deep layers of the material is partially absorbed by overlying layers.1 This self-absorption results in emission spectra from most optically thick samples that closely resemble black-body spectra. The characteristic discrete emission bands are severely truncated and altered in shape. TIRES bypasses this difficulty by using a laser to heat only an optically thin surface layer. The increased temperature of the layer is transient since the layer will rapidly cool and thicken by thermal diffusion; hence the emission collection must be correlated with the laser heating. TIRES may be done with both pulsed and cw lasers.2,3 When a pulsed laser is used, the spectrometer sampling must be synchronized with the laser pulsing so that only emission during and immediately after each laser pulse is observed.3 If a cw laser is used, the sample must move rapidly through the beam. The hot, transient layer is then in the beam track on the sample at and immediately behind the beam position, so the spectrometer field of view must be limited to this region near the beam position.2 How much self-absorption the observed emission suffers depends on how thick the heated layer has grown by thermal diffusion when the spectrometer samples the emission. Use of a pulsed laser synchronized with the spectrometer sampling readily permits reduction of the time available for heat diffusion to about 100 acs .3 When a cw laser is used, the heat-diffusion time is controlled by how small the spectrometer field of view is and by how rapidly the sample moves past within this field. Both a very small field of view and a

  18. Anticipated transients without scram

    International Nuclear Information System (INIS)

    Lellouche, G.S.

    1980-01-01

    This article discusses in various degrees of depth the publications WASH-1270, WASH-1400, and NUREG-0460, and has as its purpose a description of the technical work done by Electric Power Research Institute (EPRI) personnel and its contractors on the subject of anticipated transients without scram (ATWS). It demonstrates the close relation between the probability of scram failure derived from historical scram data and that derived from the use of component data in a model of a system (the so-called synthesis method), such as was done in WASH-1400. The inherent conservatism of these models is demonstrated by showing that they predict significantly more events than have in fact occurred and that such models still predict scram failure probabilities low enough to make ATWS an insignificant contributor to accident risk

  19. Transient fuel melting

    International Nuclear Information System (INIS)

    Roche, L.; Schmitz, F.

    1982-10-01

    The observation of micrographic documents from fuel after a CABRI test leads to postulate a specific mode of transient fuel melting during a rapid nuclear power excursion. When reaching the melt threshold, the bands which are characteristic for the solid state are broken statistically over a macroscopic region. The time of maintaining the fuel at the critical enthalpy level between solid and liquid is too short to lead to a phase separation. A significant life-time (approximately 1 second) of this intermediate ''unsolide'' state would have consequences on the variation of physical properties linked to the phase transition solid/liquid: viscosity, specific volume and (for the irradiated fuel) fission gas release [fr

  20. Transient osteoporosis of pregnancy.

    Science.gov (United States)

    Maliha, George; Morgan, Jordan; Vrahas, Mark

    2012-08-01

    Transient osteoporosis of pregnancy (TOP) is a rare yet perhaps under-reported condition that has affected otherwise healthy pregnancies throughout the world. The condition presents suddenly in the third trimester of a usually uneventful pregnancy and progressively immobilizes the mother. Radiographic studies detect drastic loss of bone mass, elevated rates of turnover in the bone, and oedema in the affected portion. Weakness of the bone can lead to fractures during delivery and other complications for the mother. Then, within weeks of labour, symptoms and radiological findings resolve. Aetiology is currently unknown, although neural, vascular, haematological, endocrine, nutrient-deficiency, and other etiologies have been proposed. Several treatments have also been explored, including simple bed rest, steroids, bisphosphonates, calcitonin, induced termination of pregnancy, and surgical intervention. The orthopedist plays an essential role in monitoring the condition (and potential complications) as well as ensuring satisfactory outcomes for both the mother and newborn. Copyright © 2012 Elsevier Ltd. All rights reserved.

  1. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

  2. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  3. Digital reactivity meter

    International Nuclear Information System (INIS)

    Akkus, B.; Anac, H.; Alsan, S.; Erk, S.

    1991-01-01

    Nowadays, various digital methods making use of microcomputers for neutron detector signals and determining the reactivity by numerical calculations are used in reactor control systems in place of classical reactivity meters. In this work, a calculation based on the ''The Time Dependent Transport Equation'' has been developed for determining the reactivity numerically. The reactivity values have been obtained utilizing a computer-based data acquisition and control system and compared with the analog reactivity meter values as well as the values calculated from the ''Inhour Equation''

  4. Transients in reactors for power systems compensation

    Science.gov (United States)

    Abdul Hamid, Haziah

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

  5. SCANAIR: A transient fuel performance code

    International Nuclear Information System (INIS)

    Moal, Alain; Georgenthum, Vincent; Marchand, Olivier

    2014-01-01

    Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

  6. SCANAIR: A transient fuel performance code

    Energy Technology Data Exchange (ETDEWEB)

    Moal, Alain, E-mail: alain.moal@irsn.fr; Georgenthum, Vincent; Marchand, Olivier

    2014-12-15

    Highlights: • Since the early 1990s, the code SCANAIR is developed at IRSN. • The software focuses on studying fast transients such as RIA in light water reactors. • The fuel rod modelling is based on a 1.5D approach. • Thermal and thermal-hydraulics, mechanical and gas behaviour resolutions are coupled. • The code is used for safety assessment and integral tests analysis. - Abstract: Since the early 1990s, the French “Institut de Radioprotection et de Sûreté Nucléaire” (IRSN) has developed the SCANAIR computer code with the view to analysing pressurised water reactor (PWR) safety. This software specifically focuses on studying fast transients such as reactivity-initiated accidents (RIA) caused by possible ejection of control rods. The code aims at improving the global understanding of the physical mechanisms governing the thermal-mechanical behaviour of a single rod. It is currently used to analyse integral tests performed in CABRI and NSRR experimental reactors. The resulting validated code is used to carry out studies required to evaluate margins in relation to criteria for different types of fuel rods used in nuclear power plants. Because phenomena occurring during fast power transients are complex, the simulation in SCANAIR is based on a close coupling between several modules aimed at modelling thermal, thermal-hydraulics, mechanical and gas behaviour. During the first stage of fast power transients, clad deformation is mainly governed by the pellet–clad mechanical interaction (PCMI). At the later stage, heat transfers from pellet to clad bring the cladding material to such high temperatures that the boiling crisis might occurs. The significant over-pressurisation of the rod and the fact of maintaining the cladding material at elevated temperatures during a fairly long period can lead to ballooning and possible clad failure. A brief introduction describes the context, the historical background and recalls the main phenomena involved under

  7. Transient Go: A Mobile App for Transient Astronomy Outreach

    Science.gov (United States)

    Crichton, D.; Mahabal, A.; Djorgovski, S. G.; Drake, A.; Early, J.; Ivezic, Z.; Jacoby, S.; Kanbur, S.

    2016-12-01

    Augmented Reality (AR) is set to revolutionize human interaction with the real world as demonstrated by the phenomenal success of `Pokemon Go'. That very technology can be used to rekindle the interest in science at the school level. We are in the process of developing a prototype app based on sky maps that will use AR to introduce different classes of astronomical transients to students as they are discovered i.e. in real-time. This will involve transient streams from surveys such as the Catalina Real-time Transient Survey (CRTS) today and the Large Synoptic Survey Telescope (LSST) in the near future. The transient streams will be combined with archival and latest image cut-outs and other auxiliary data as well as historical and statistical perspectives on each of the transient types being served. Such an app could easily be adapted to work with various NASA missions and NSF projects to enrich the student experience.

  8. Method of controlling reactivity

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi.

    1982-01-01

    Purpose: To improve the reactivity controlling characteristics by artificially controlling the leakage of neutron from a reactor and providing a controller for controlling the reactivity. Method: A reactor core is divided into several water gaps to increase the leakage of neutron, its reactivity is reduced, a gas-filled control rod or a fuel assembly is inserted into the gap as required, the entire core is coupled in a system to reduce the leakage of the neutron, and the reactivity is increased. The reactor shutdown is conducted by the conventional control rod, and to maintain critical state, boron density varying system is used together. Futher, a control rod drive is used with that similar to the conventional one, thereby enabling fast reactivity variation, and the positive reactivity can be obtained by the insertion, thereby improving the reactivity controlling characteristics. (Yoshihara, H.)

  9. Experimental study on transient boiling heat transfer

    International Nuclear Information System (INIS)

    Visentini, R.

    2012-01-01

    Boiling phenomena can be found in the everyday life, thus a lot of studies are devoted to them, especially in steady state conditions. Transient boiling is less known but still interesting as it is involved in the nuclear safety prevention. In this context, the present work was supported by the French Institute of Nuclear Safety (IRSN). In fact, the IRSN wanted to clarify what happens during a Reactivity-initiated Accident (RIA). This accident occurs when the bars that control the nuclear reactions break down and a high power peak is passed from the nuclear fuel bar to the surrounding fluid. The temperature of the nuclear fuel bar wall increases and the fluid vaporises instantaneously. Previous studies on a fuel bar or on a metal tube heated by Joule effect were done in the past in order to understand the rapid boiling phenomena during a RIA. However, the measurements were not really accurate because the measurement techniques were not able to follow rapid phenomena. The main goal of this work was to create an experimental facility able to simulate the RIA boiling conditions but at small scale in order to better understand the boiling characteristics when the heated-wall temperature increases rapidly. Moreover, the experimental set-up was meant to be able to produce less-rapid transients as well, in order to give information on transient boiling in general. The facility was built at the Fluid-Mechanics Institute of Toulouse. The core consists of a metal half-cylinder heated by Joule effect, placed in a half-annulus section. The inner half cylinder is made of a 50 microns thick stainless steel foil. Its diameter is 8 mm, and its length 200 mm. The outer part is a 34 mm internal diameter glass half cylinder. The semi-annular section is filled with a coolant, named HFE7000. The configuration allows to work in similarity conditions. The heated part can be place inside a loop in order to study the flow effect. The fluid temperature influence is taken into account as

  10. Analysis of reactivity feedback effects of void and temperature in the MAPLE-X10 reactor

    International Nuclear Information System (INIS)

    Carlson, P.A.; Heeds, W.; Shim, S.Y.; King, S.G.

    1992-07-01

    The methods used for evaluating the void and temperature reactivity coefficients for the MAPLE-X10 Reactor are described and factors used in estimating their accuracy are discussed. The report presents representative transient analysis results using the CATHENA thermalhydraulics code. The role of the reactivity coefficients and their precision is discussed. The results are reviewed in terms of their safety implications

  11. Various reactivity effects value for assuring fast reactor core inherent safety

    International Nuclear Information System (INIS)

    Belov, S.B.; Vasilyev, B.A.

    1991-01-01

    The paper presents the results of temperature and power reactivity feedback components calculations for fast reactors with different core volume when using oxide, carbide, nitride and metal fuel. Reactor parameters change in loss of flow without scram and transient over power without scram accidents was evaluated. The importance of various reactivity feedback components in restricting the consequences of these accidents has been analyzed. (author)

  12. Pressure transients in pipeline systems

    DEFF Research Database (Denmark)

    Voigt, Kristian

    1998-01-01

    This text is to give an overview of the necessary background to do investigation of pressure transients via simulations. It will describe briefly the Method of Characteristics which is the defacto standard for simulating pressure transients. Much of the text has been adopted from the book Pressur...

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

    Energy Technology Data Exchange (ETDEWEB)

    Vazquez, M.; Martin-Fuertes, F.

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  15. Study of the initiation of subcooled boiling during power transients

    International Nuclear Information System (INIS)

    VanVleet, R.J.

    1985-01-01

    An experimental investigation of boiling initiation during power transients has been conducted for horizontal-cylinder heating elements in degassed distilled water. Platinum elements, 0.127 and 0.250 mm in diameter, were internally heated electrically at a controlled superficial heat flux (power applied divided by surface area) increasing linearly with time at rates of 0.035 and 0.35 MW/m 2 s and corresponding test durations of 20 and 2 seconds. Tests were carried out at saturation temperatures from 100 to 195 0 C with bulk fluid subcooling from 0 to 30 K. During the course of a power transient, element temperature and superficial heat flux were measured electrically and the boiling initiation time was determined optically. It was found that the conditions for boiling initiation depended strongly on the pressure-temperature history of the heating element and surround fluid prior to the transient. Boiling initiation times were found to agree qualitatively with predictions of a model based on the contact-angle hysteresis concept. Brief prepressurization prior to a transient was found to increase dramatically the temperature and heat flux required for boiling initiation because of deactivation of boiling initiation sites. However, sites were re-activated during the transient and, in subsequent tests without prepressurization, no elevation in boiling initiation conditions was observed and results were in quantitative agreement with predictions of the model

  16. Transient regional osteoporosis.

    Science.gov (United States)

    Cano-Marquina, Antonio; Tarín, Juan J; García-Pérez, Miguel-Ángel; Cano, Antonio

    2014-04-01

    Transient regional osteoporosis (TRO) is a disease that predisposes to fragility fracture in weight bearing joints of mid-life women and men. Pregnant women may also suffer the process, usually at the hip. The prevalence of TRO is lower than the systemic form, associated with postmenopause and advanced age, but may be falsely diminished by under-diagnosis. The disease may be uni- or bilateral, and may migrate to distinct joints. One main feature of TRO is spontaneous recovery. Pain and progressive limitation in the functionality of the affected joint(s) are key symptoms. In the case of the form associated with pregnancy, difficulties in diagnosis derive from the relatively young age at presentation and from the clinical overlapping with the frequent aches during gestation. Densitometric osteoporosis in the affected region is not always present, but bone marrow edema, with or without joint effusion, is detected by magnetic resonance. There are not treatment guidelines, but the association of antiresorptives to symptomatic treatment seems to be beneficial. Surgery or other orthopedic interventions can be required for specific indications, like hip fracture, intra-medullary decompression, or other. Copyright © 2014 Elsevier Ireland Ltd. All rights reserved.

  17. Applied hydraulic transients

    CERN Document Server

    Chaudhry, M Hanif

    2014-01-01

    This book covers hydraulic transients in a comprehensive and systematic manner from introduction to advanced level and presents various methods of analysis for computer solution. The field of application of the book is very broad and diverse and covers areas such as hydroelectric projects, pumped storage schemes, water-supply systems, cooling-water systems, oil pipelines and industrial piping systems. Strong emphasis is given to practical applications, including several case studies, problems of applied nature, and design criteria. This will help design engineers and introduce students to real-life projects. This book also: ·         Presents modern methods of analysis suitable for computer analysis, such as the method of characteristics, explicit and implicit finite-difference methods and matrix methods ·         Includes case studies of actual projects ·         Provides extensive and complete treatment of governed hydraulic turbines ·         Presents design charts, desi...

  18. Output of the type 4051 and 4061 period meters during startup transients

    International Nuclear Information System (INIS)

    Cummins, J.D.

    1963-05-01

    The report describes a digital computer programme for the Ferranti Mercury computer. With this programme startup transients for the recently developed period meters Types 4051 and 4061 may be computed. The reactivity disturbances considered are steps and terminated ramps of reactivity. Due allowance is taken of the variable time constant which is a feature of these period meters. The reactor may be critical or subcritical initially as desired and the initial input time constant of the period meter may be zero or finite. Some representative transients obtained with the programme are presented and discussed. (author)

  19. Transient Stability Enhancement in Power System Using Static VAR Compensator (SVC

    Directory of Open Access Journals (Sweden)

    Youssef MOULOUDI

    2012-12-01

    Full Text Available In this paper, an indirect adaptive fuzzy excitation and static VAR (unit of reactive power, volt-ampere reactive compensator (SVC controller is proposed to enhance transient stability for the power system, which based on input-output linearization technique. A three-bus system, which contains a generator and static VAR compensator (SVC, is considered in this paper, the SVC is located at the midpoint of the transmission lines. Simulation results show that the proposed controller compared with a controller based on tradition linearization technique can enhance the transient stability of the power system under a large sudden fault, which may occur nearly at the generator bus terminal.

  20. Reactive solute transport in acidic streams

    Science.gov (United States)

    Broshears, R.E.

    1996-01-01

    Spatial and temporal profiles of Ph and concentrations of toxic metals in streams affected by acid mine drainage are the result of the interplay of physical and biogeochemical processes. This paper describes a reactive solute transport model that provides a physically and thermodynamically quantitative interpretation of these profiles. The model combines a transport module that includes advection-dispersion and transient storage with a geochemical speciation module based on MINTEQA2. Input to the model includes stream hydrologic properties derived from tracer-dilution experiments, headwater and lateral inflow concentrations analyzed in field samples, and a thermodynamic database. Simulations reproduced the general features of steady-state patterns of observed pH and concentrations of aluminum and sulfate in St. Kevin Gulch, an acid mine drainage stream near Leadville, Colorado. These patterns were altered temporarily by injection of sodium carbonate into the stream. A transient simulation reproduced the observed effects of the base injection.

  1. Radial core expansion reactivity feedback in advanced LMRs: uncertainties and their effects on inherent safety

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Moran, T.J.

    1988-01-01

    An analytical model for calculating radial core expansion, based on the thermal and elastic bowing of a single subassembly at the core periphery, is used to quantify the effect of uncertainties on this reactivity feedback mechanism. This model has been verified and validated with experimental and numerical results. The impact of these uncertainties on the safety margins in unprotected transients is investigated with SASSYS/SAS4A, which includes this model for calculating the reactivity feedback from radial core expansion. The magnitudes of these uncertainties are not sufficient to preclude the use of radial core expansion reactivity feedback in transient analysis

  2. Reactive Programming in Java

    CERN Document Server

    CERN. Geneva

    2017-01-01

    Reactive Programming in gaining a lot of excitement. Many libraries, tools, and frameworks are beginning to make use of reactive libraries. Besides, applications dealing with big data or high frequency data can benefit from this programming paradigm. Come to this presentation to learn about what reactive programming is, what kind of problems it solves, how it solves them. We will take an example oriented approach to learning the programming model and the abstraction.

  3. BN600 reactivity definition

    International Nuclear Information System (INIS)

    Zheltyshev, V.; Ivanov, A.

    2000-01-01

    Since 1980, the fast BN600 reactor with sodium coolant has been operated at Beloyarsk Nuclear Power Plant. The periodic monitoring of the reactivity modifications should be implemented in compliance with the standards and regulations applied in nuclear power engineering. The reactivity measurements are carried out in order to confirm the basic neutronic features of a BN600 reactor. The reactivity measurements are aimed to justify that nuclear safety is provided in course of the in-reactor installation of the experimental core components. Two reactivity meters are to be used on BN600 operation: 1. Digital on-line reactivity calculated under stationary reactor operation on power (approximation of the point-wise kinetics is applied). 2. Second reactivity meter used to define the reactor control rod operating components efficiency under reactor startup and take account of the changing efficiency of the sensor, however, this is more time-consumptive than the on-line reactivity meter. The application of two reactivity meters allows for the monitoring of the reactor reactivity under every operating mode. (authors)

  4. Explosive and radio-selected Transients: Transient Astronomy with ...

    Indian Academy of Sciences (India)

    40

    sitive measurements will lead to very accurate mass loss estimation in these supernovae. .... transients are powerful probes of intervening media owing to dispersion ...... A., & Chandra, P. 2011, Nature Communications,. 2, 175. Chakraborti, S.

  5. Transient-Switch-Signal Suppressor

    Science.gov (United States)

    Bozeman, Richard J., Jr.

    1995-01-01

    Circuit delays transmission of switch-opening or switch-closing signal until after preset suppression time. Used to prevent transmission of undesired momentary switch signal. Basic mode of operation simple. Beginning of switch signal initiates timing sequence. If switch signal persists after preset suppression time, circuit transmits switch signal to external circuitry. If switch signal no longer present after suppression time, switch signal deemed transient, and circuit does not pass signal on to external circuitry, as though no transient switch signal. Suppression time preset at value large enough to allow for damping of underlying pressure wave or other mechanical transient.

  6. Electromagnetic transients in power cables

    CERN Document Server

    da Silva, Filipe Faria

    2013-01-01

    From the more basic concepts to the most advanced ones where long and laborious simulation models are required, Electromagnetic Transients in Power Cables provides a thorough insight into the study of electromagnetic transients and underground power cables. Explanations and demonstrations of different electromagnetic transient phenomena are provided, from simple lumped-parameter circuits to complex cable-based high voltage networks, as well as instructions on how to model the cables.Supported throughout by illustrations, circuit diagrams and simulation results, each chapter contains exercises,

  7. Mechanisms of ignition by transient energy deposition: Regimes of combustion wave propagation

    OpenAIRE

    Kiverin, A. D.; Kassoy, D. R.; Ivanov, M. F.; Liberman, M. A.

    2013-01-01

    Regimes of chemical reaction wave propagating in reactive gaseous mixtures, whose chemistry is governed by chain-branching kinetics, are studied depending on the characteristics of a transient thermal energy deposition localized in a finite volume of reactive gas. Different regimes of the reaction wave propagation are initiated depending on the amount of deposited thermal energy, power of the source, and the size of the hot spot. The main parameters which define regimes of the combustion wave...

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  9. Reactivity feedback models for SSC-K

    Energy Technology Data Exchange (ETDEWEB)

    Han, Do Hee; Kwon, Young Min; Kim, Kyung Du; Chang, Won Pyo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    Safety of KALIMER is assured by the inherent safety of the core and passive safety of the safety-related systems. For the safety analysis of a new reactor design such as KALIMER, analysis models, which are consistent with the design, have to be developed for a plant-wide transient and safety analysis code. Efforts for the development of reactivity feedback models for SSC-K, which is now being developed for the safety analysis of KALIMER, is described in this report. Models for Doppler, sodium density/void, fuel axial expansion, core radial expansion, and CRDL expansion have been developed. Test runs have been performed for the unprotected accident for the verification of the models. Use of KALIMER reactivity coefficients and future development of models for GEM and PSDRS would make it possible to analyze the response of KALIMER under TOP as well as LOF and LOHS accident conditions using SSC-K. (author). 5 refs., 64 figs., 2 tabs.

  10. Reactivity accident analysis in MTR cores

    International Nuclear Information System (INIS)

    Waldman, R.M.; Vertullo, A.C.

    1987-01-01

    The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

  11. Future Transient Testing of Advanced Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack

    2009-09-01

    The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by

  12. Future Transient Testing of Advanced Fuels

    International Nuclear Information System (INIS)

    Carmack, Jon

    2009-01-01

    The transient in-reactor fuels testing workshop was held on May 4-5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat energie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric - Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  15. Electrospinning of reactive mesogens

    NARCIS (Netherlands)

    Yao, J.; Picot, O.T.; Hughes-Brittain, N.F.; Bastiaansen, C.W.M.; Peijs, T.

    2016-01-01

    The reinforcement potential of reactive liquid crystals or reactive mesogens (RMs) in electrospun fibers was investigated through the blending of two types of RMs (RM257 and RM82) with two types of thermoplastics; polyamide 6 (PA6) and poly(methyl methacrylate) (PMMA). Polymer/RM blends were

  16. Transient or permanent fisheye views

    DEFF Research Database (Denmark)

    Jakobsen, Mikkel Rønne; Hornbæk, Kasper

    2012-01-01

    Transient use of information visualization may support specific tasks without permanently changing the user interface. Transient visualizations provide immediate and transient use of information visualization close to and in the context of the user’s focus of attention. Little is known, however......, about the benefits and limitations of transient visualizations. We describe an experiment that compares the usability of a fisheye view that participants could call up temporarily, a permanent fisheye view, and a linear view: all interfaces gave access to source code in the editor of a widespread...... programming environment. Fourteen participants performed varied tasks involving navigation and understanding of source code. Participants used the three interfaces for between four and six hours in all. Time and accuracy measures were inconclusive, but subjective data showed a preference for the permanent...

  17. Transient thyrotoxicosis during nivolumab treatment

    NARCIS (Netherlands)

    van Kooten, M. J.; van den Berg, G.; Glaudemans, A. W. J. M.; Hiltermann, T. J. N.; Groen, H. J. M.; Rutgers, A.; Links, T. P.

    Two patients presented with transient thyrotoxicosis within 2-4 weeks after starting treatment with nivolumab. This thyrotoxicosis turned into hypothyroidism within 6-8 weeks. Temporary treatment with a beta blocker may be sufficient.

  18. Transient Tsunamis in Lakes

    Science.gov (United States)

    Couston, L.; Mei, C.; Alam, M.

    2013-12-01

    A large number of lakes are surrounded by steep and unstable mountains with slopes prone to failure. As a result, landslides are likely to occur and impact water sitting in closed reservoirs. These rare geological phenomena pose serious threats to dam reservoirs and nearshore facilities because they can generate unexpectedly large tsunami waves. In fact, the tallest wave experienced by contemporary humans occurred because of a landslide in the narrow bay of Lituya in 1958, and five years later, a deadly landslide tsunami overtopped Lake Vajont's dam, flooding and damaging villages along the lakefront and in the Piave valley. If unstable slopes and potential slides are detected ahead of time, inundation maps can be drawn to help people know the risks, and mitigate the destructive power of the ensuing waves. These maps give the maximum wave runup height along the lake's vertical and sloping boundaries, and can be obtained by numerical simulations. Keeping track of the moving shorelines along beaches is challenging in classical Eulerian formulations because the horizontal extent of the fluid domain can change over time. As a result, assuming a solid slide and nonbreaking waves, here we develop a nonlinear shallow-water model equation in the Lagrangian framework to address the problem of transient landslide-tsunamis. In this manner, the shorelines' three-dimensional motion is part of the solution. The model equation is hyperbolic and can be solved numerically by finite differences. Here, a 4th order Runge-Kutta method and a compact finite-difference scheme are implemented to integrate in time and spatially discretize the forced shallow-water equation in Lagrangian coordinates. The formulation is applied to different lake and slide geometries to better understand the effects of the lake's finite lengths and slide's forcing mechanism on the generated wavefield. Specifically, for a slide moving down a plane beach, we show that edge-waves trapped by the shoreline and free

  19. Transient two-phase flow

    International Nuclear Information System (INIS)

    Hsu, Y.Y.

    1974-01-01

    The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

  20. Determination of the design excess reactivity for the TREAT Upgrade reactor

    International Nuclear Information System (INIS)

    Bhattacharyya, S.K.; Hanan, N.A.

    1983-01-01

    The excess reactivity designed to be built into a reactor core is a primary determinant of the fissile loadings of the fuel rods in the core. For the TREAT Upgrade (TU) reactor the considerations that enter into the determination of the excess reactivity are different from those of conventional power reactors. The reactor is designed to operate in an adiabatic transient mode for reactor safety in-pile test programs. The primary constituent of the excess reactivity is the calculated reactivity required to perform the most demanding transient experiments. Because of the unavailability of supporting critical experiments for the core design, the uncertainty terms that add on to this basic constituent are rather large. The burnup effects in TU are negligible and no refueling is planned. In this paper the determination of the design excess reactivity of the TREAT Upgrade reactor is discussed

  1. Performance of neutron kinetics models for ADS transient analyses

    International Nuclear Information System (INIS)

    Rineiski, A.; Maschek, W.; Rimpault, G.

    2002-01-01

    can also apply this approach for estimating errors of point-kinetics simulations or for ameliorating the employed point-kinetics models. Though the performance of the point-kinetics model can be insufficient in the subcritical case, the quasi-static approach is still valid if the shape steps are chosen properly. It is worthwhile to mention that in combination with properly computed correction factor tables, one can use the reactivity and power distributions obtained for 'critical' reactor models; this approach can simplify ADS-related application of conventional accident analyses codes (developed in the past for transient analyses of critical reactors). However, for analyzing severe transients in ADSs, which involve gross core material configuration changes, one can hardly avoid using of space-time kinetics methods, this holds similarly for critical reactor systems. (authors)

  2. Transient magnetoviscosity of dilute ferrofluids

    International Nuclear Information System (INIS)

    Soto-Aquino, Denisse; Rinaldi, Carlos

    2011-01-01

    The magnetic field induced change in the viscosity of a ferrofluid, commonly known as the magnetoviscous effect and parameterized through the magnetoviscosity, is one of the most interesting and practically relevant aspects of ferrofluid phenomena. Although the steady state behavior of ferrofluids under conditions of applied constant magnetic fields has received considerable attention, comparatively little attention has been given to the transient response of the magnetoviscosity to changes in the applied magnetic field or rate of shear deformation. Such transient response can provide further insight into the dynamics of ferrofluids and find practical application in the design of devices that take advantage of the magnetoviscous effect and inevitably must deal with changes in the applied magnetic field and deformation. In this contribution Brownian dynamics simulations and a simple model based on the ferrohydrodynamics equations are applied to explore the dependence of the transient magnetoviscosity for two cases: (I) a ferrofluid in a constant shear flow wherein the magnetic field is suddenly turned on, and (II) a ferrofluid in a constant magnetic field wherein the shear flow is suddenly started. Both simulations and analysis show that the transient approach to a steady state magnetoviscosity can be either monotonic or oscillatory depending on the relative magnitudes of the applied magnetic field and shear rate. - Research Highlights: →Rotational Brownian dynamics simulations were used to study the transient behavior of the magnetoviscosity of ferrofluids. →Damped and oscillatory approach to steady state magnetoviscosity was observed for step changes in shear rate and magnetic field. →A model based on the ferrohydrodynamics equations qualitatively captured the damped and oscillatory features of the transient response →The transient behavior is due to the interplay of hydrodynamic, magnetic, and Brownian torques on the suspended particles.

  3. Advanced methods for BWR transient and stability analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schmidt, A; Wehle, F; Opel, S; Velten, R [AREVA, AREVA NP, Erlangen (Germany)

    2008-07-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  4. Advanced methods for BWR transient and stability analysis

    International Nuclear Information System (INIS)

    Schmidt, A.; Wehle, F.; Opel, S.; Velten, R.

    2008-01-01

    The design of advanced Boiling Water Reactor (BWR) fuel assemblies and cores is governed by the basic requirement of safe, reliable and flexible reactor operation with optimal fuel utilization. AREVA NP's comprehensive steady state and transient BWR methodology allows the designer to respond quickly and effectively to customer needs. AREVA NP uses S-RELAP5/RAMONA as the appropriate methodology for the representation of the entire plant. The 3D neutron kinetics and thermal-hydraulics code has been developed for the prediction of system, fuel and core behavior and provides additional margins for normal operation and transients. Of major importance is the extensive validation of the methodology. The validation is based on measurements at AREVA NP's test facilities, and comparison of the predictions with a great wealth of measured data gathered from BWR plants during many years of operation. Three of the main fields of interest are stability analysis, operational transients and reactivity initiated accidents (RIAs). The introduced 3D methodology for operational transients shows significant margin regarding the operational limit of critical power ratio, which has been approved by the German licensing authority. Regarding BWR stability a large number of measurements at different plants under various conditions have been performed and successfully post-calculated with RAMONA. This is the basis of reliable pre-calculations of the locations of regional and core-wide stability boundaries. (authors)

  5. Transient plasma cobalamin elevation in patients with pneumonia - two case reports

    DEFF Research Database (Denmark)

    Rahbek, Martin Torp; Scheller, Rudolf; Nybo, Mads

    2018-01-01

    We report two cases of transient significantly elevated plasma cobalamin (B12) in geriatric patients acutely admitted with fever, increased C-reactive protein and X-ray verified pneumonia. Extensive diagnostic workup did not reveal kidney or liver disease, neither any signs of cancer. Furthermore...

  6. Comparison of SAS3A and MELT-III predictions for a transient overpower hypothetical accident

    International Nuclear Information System (INIS)

    Wilburn, N.P.

    1976-01-01

    A comparison is made of the predictions of the two major codes SAS3A and MELT-III for the hypothetical unprotected transient overpower accident in the FFTF. The predictions of temperatures, fuel restructuring, fuel melting, reactivity feedbacks, and core power are compared

  7. Pressure transients across HEPA filters

    International Nuclear Information System (INIS)

    Gregory, W.; Reynolds, G.; Ricketts, C.; Smith, P.R.

    1977-01-01

    Nuclear fuel cycle facilities require ventilation for health and safety reasons. High efficiency particulate air (HEPA) filters are located within ventilation systems to trap radioactive dust released in reprocessing and fabrication operations. Pressure transients within the air cleaning systems may be such that the effectiveness of the filtration system is questioned under certain accident conditions. These pressure transients can result from both natural and man-caused phenomena: atmospheric pressure drop caused by a tornado or explosions and nuclear excursions initiate pressure pulses that could create undesirable conditions across HEPA filters. Tornado depressurization is a relatively slow transient as compared to pressure pulses that result from combustible hydrogen-air mixtures. Experimental investigation of these pressure transients across air cleaning equipment has been undertaken by Los Alamos Scientific Laboratory and New Mexico State University. An experimental apparatus has been constructed to impose pressure pulses across HEPA filters. The experimental equipment is described as well as preliminary results using variable pressurization rates. Two modes of filtration of an aerosol injected upstream of the filter is examined. A laser instrumentation for measuring the aerosol release, during the transient, is described

  8. Recent development of transient electronics

    Directory of Open Access Journals (Sweden)

    Huanyu Cheng

    2016-01-01

    Full Text Available Transient electronics are an emerging class of electronics with the unique characteristic to completely dissolve within a programmed period of time. Since no harmful byproducts are released, these electronics can be used in the human body as a diagnostic tool, for instance, or they can be used as environmentally friendly alternatives to existing electronics which disintegrate when exposed to water. Thus, the most crucial aspect of transient electronics is their ability to disintegrate in a practical manner and a review of the literature on this topic is essential for understanding the current capabilities of transient electronics and areas of future research. In the past, only partial dissolution of transient electronics was possible, however, total dissolution has been achieved with a recent discovery that silicon nanomembrane undergoes hydrolysis. The use of single- and multi-layered structures has also been explored as a way to extend the lifetime of the electronics. Analytical models have been developed to study the dissolution of various functional materials as well as the devices constructed from this set of functional materials and these models prove to be useful in the design of the transient electronics.

  9. Wide Field Radio Transient Surveys

    Science.gov (United States)

    Bower, Geoffrey

    2011-04-01

    The time domain of the radio wavelength sky has been only sparsely explored. Nevertheless, serendipitous discovery and results from limited surveys indicate that there is much to be found on timescales from nanoseconds to years and at wavelengths from meters to millimeters. These observations have revealed unexpected phenomena such as rotating radio transients and coherent pulses from brown dwarfs. Additionally, archival studies have revealed an unknown class of radio transients without radio, optical, or high-energy hosts. The new generation of centimeter-wave radio telescopes such as the Allen Telescope Array (ATA) will exploit wide fields of view and flexible digital signal processing to systematically explore radio transient parameter space, as well as lay the scientific and technical foundation for the Square Kilometer Array. Known unknowns that will be the target of future transient surveys include orphan gamma-ray burst afterglows, radio supernovae, tidally-disrupted stars, flare stars, and magnetars. While probing the variable sky, these surveys will also provide unprecedented information on the static radio sky. I will present results from three large ATA surveys (the Fly's Eye survey, the ATA Twenty CM Survey (ATATS), and the Pi GHz Survey (PiGSS)) and several small ATA transient searches. Finally, I will discuss the landscape and opportunities for future instruments at centimeter wavelengths.

  10. TRACY transient experiment databook. 2) ramp withdrawal experiment

    International Nuclear Information System (INIS)

    Nakajima, Ken; Yamane, Yuichi; Ogawa, Kazuhiko; Aizawa, Eiju; Yanagisawa, Hiroshi; Miyoshi, Yoshinori

    2002-03-01

    This is a databook of TRACY ''ramp withdrawal'' experiments. TRACY is a reactor to perform supercritical experiments using low-enriched uranyl nitrate aqueous solution. The excess reactivity of TRACY is 3$ at maximum, and it is inserted by feeding the solution to a core tank or by withdrawing a control rod, which is called as the transient rod, from the core. In the ramp withdrawal experiment, the supercritical experiment is initiated by withdrawing the transient rod from the core in a constant speed using a motor drive system. The data in the present databook consist of datasheets and graphs. Experimental conditions and typical values of measured parameters are tabulated in the datasheet. In the graph, power and temperature profiles are plotted. Those data are useful for the investigation of criticality accidents with fissile solutions, and for validation of criticality accident analysis codes. (author)

  11. A faster reactor transient analysis methodology for PCs

    International Nuclear Information System (INIS)

    Ott, K.O.

    1991-10-01

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

  12. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    Carew, J.F.; Neogy, P.

    1983-01-01

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

  13. Digital reactivity meter

    International Nuclear Information System (INIS)

    Jiang Zongbing

    1996-02-01

    The importance and the usual methods of reactivity measurement in a nuclear reactor are presented. Emphasis is put upon the calculation principle, software and hardware components, main specifications, application, as well as the features of the digital reactivity meter. The test results of operation in various reactors shown that the meter possess the following features: high accuracy, short response time, low output noise, high resolution, wide measuring range, simple and flexible to operate, high stability and reliability. In addition, the reactivity meter can save the measuring data automatically and have a perfect capability of self-verifying. It not only meet the requirement of the reactivity measurement in nuclear power plant, but also can be applied to various types of reactors. (1 tab.)

  14. Stress Reactivity in Insomnia.

    Science.gov (United States)

    Gehrman, Philip R; Hall, Martica; Barilla, Holly; Buysse, Daniel; Perlis, Michael; Gooneratne, Nalaka; Ross, Richard J

    2016-01-01

    This study examined whether individuals with primary insomnia (PI) are more reactive to stress than good sleepers (GS). PI and GS (n = 20 per group), matched on gender and age, completed three nights of polysomnography. On the stress night, participants received a mild electric shock and were told they could receive additional shocks during the night. Saliva samples were obtained for analysis of cortisol and alpha amylase along with self-report and visual analog scales (VAS). There was very little evidence of increased stress on the stress night, compared to the baseline night. There was also no evidence of greater stress reactivity in the PI group for any sleep or for salivary measures. In the GS group, stress reactivity measured by VAS scales was positively associated with an increase in sleep latency in the experimental night on exploratory analyses. Individuals with PI did not show greater stress reactivity compared to GS.

  15. Structure, Reactivity and Dynamics

    Indian Academy of Sciences (India)

    Understanding structure, reactivity and dynamics is the core issue in chemical ... functional theory (DFT) calculations, molecular dynamics (MD) simulations, light- ... between water and protein oxygen atoms, the superionic conductors which ...

  16. Taskable Reactive Agent Communities

    National Research Council Canada - National Science Library

    Myers, Karen

    2002-01-01

    The focus of Taskable Reactive Agent Communities (TRAC) project was to develop mixed-initiative technology to enable humans to supervise and manage teams of agents as they perform tasks in dynamic environments...

  17. Reactive sputter deposition

    CERN Document Server

    Mahieu, Stijn

    2008-01-01

    In this valuable work, all aspects of the reactive magnetron sputtering process, from the discharge up to the resulting thin film growth, are described in detail, allowing the reader to understand the complete process. Hence, this book gives necessary information for those who want to start with reactive magnetron sputtering, understand and investigate the technique, control their sputtering process and tune their existing process, obtaining the desired thin films.

  18. Fuel pin response to an overpower transient in an LMFBR

    International Nuclear Information System (INIS)

    Grosberg, A.J.; Head, J.L.

    1979-01-01

    This paper describes a method by which the ability of a whole-core code accurately to predict the time and location of the first fuel pin failures may be tested. The method involves the use of a relatively simple whole-core code to 'drive' a sophisticated fuel pin code, which is far too complex to be used within a whole-core code but which is potentially capable of modelling reliably the response of an individual fuel pin. The method cannot follow accurately the subsequent course of the transient because the simple whole-core code does not model the reactivity effects of events which may follow pin failure. The codes used were the simple whole-core code FUTURE and the fuel pin behaviour code FRUMP. The paper describes an application of the method to analyse a hypothetical LMFBR accident in which the control rods were assumed to be driven from the core at maximum speed, with all trip circuits failed. Taking 0.5% clad strain as a clad failure criterion, failure was predicted to occur at the top of the active core at about 10s into the transient. A repeat analysis, using an alternative clad yield criterion which is thought to be more realistic, indicated failure at the same position but 24s into the transient. This is after the onset of sodium boiling. Pin failure at the top of the core are likely to cause negative reactivity changes. In this hypothetical accident, pin failures are likely, therefore, to have a moderating effect on the course of the transient. (orig.)

  19. HEDL experimental transient overpower program

    International Nuclear Information System (INIS)

    Hikido, T.; Culley, G.E.

    1976-01-01

    HEDL is conducting a series of experiments to evaluate the performance of Fast Flux Test Facility (FFTF) prototypic fuel pins up to the point of cladding breach. A primary objective of the program is to demonstrate the adequacy of fuel pin and Plant Protective System (PPS) designs for terminated transients. Transient tests of prototypic FFTF fuel pins previously irradiated in the Experimental Breeder Reactor-II (EBR-II) have demonstrated the adequacy of the PPS and fuel pin designs and indicate that a very substantial margin exists between PPS-terminated transients and that required to produce fuel pin cladding failure. Additional experiments are planned to extend the data base to high burnup, high fluence fuel pin specimens

  20. Transient voltage oscillations in coils

    International Nuclear Information System (INIS)

    Chowdhuri, P.

    1985-01-01

    Magnet coils may be excited into internal voltage oscillations by transient voltages. Such oscillations may electrically stress the magnet's dielectric components to many times its normal stress. This may precipitate a dielectric failure, and the attendant prolonged loss of service and costly repair work. Therefore, it is important to know the natural frequencies of oscillations of a magnet during the design stage, and to determine whether the expected switching transient voltages can excite the magnet into high-voltage internal oscillations. The series capacitance of a winding significantly affects its natural frequencies. However, the series capacitance is difficult to calculate, because it may comprise complex capacitance network, consisting of intra- and inter-coil turn-to-turn capacitances of the coil sections. A method of calculating the series capacitance of a winding is proposed. This method is rigorous but simple to execute. The time-varying transient voltages along the winding are also calculated

  1. Transient analysis of DTT rakes

    International Nuclear Information System (INIS)

    Kamath, P.S.; Lahey, R.T. Jr.

    1981-01-01

    This paper presents an analytical model for the determination of the cross-sectionally averaged transient mass flux of a two-phase fluid flowing in a conduit instrumented by a Drag-Disk Turbine Transducer (DTT) Rake and a multibeam gamma densitometer. Parametric studies indicate that for a typical blowdown transient, dynamic effects such as rotor inertia can be important for the turbine-meter. In contrast, for the drag-disk, a frequency response analysis showed that the quasisteady solution is valid below a forcing frequency of about 10 Hz, which is faster than the time scale normally encountered during blowdowns. The model showed reasonably good agreement with full scale transient rake data, where the flow regimes were mostly homogeneous or stratified, thus indicating that the model is suitable for the analysis of a DTT rake. (orig.)

  2. Transient analysis of multicavity klystrons

    International Nuclear Information System (INIS)

    Lavine, T.L.; Miller, R.H.; Morton, P.L.; Ruth, R.D.

    1988-09-01

    We describe a model for analytic analysis of transients in multicavity klystron output power and phase. Cavities are modeled as resonant circuits, while bunching of the beam is modeled using linear space-charge wave theory. Our analysis has been implemented in a computer program which we use in designing multicavity klystrons with stable output power and phase. We present as examples transient analysis of a relativistic klystron using a magnetic pulse compression modulator, and of a conventional klystron designed to use phase shifting techniques for RF pulse compression. 4 refs., 4 figs

  3. Transient formation of forbidden lines

    International Nuclear Information System (INIS)

    Rosmej, F.B.; Rosmej, O.N.

    1996-01-01

    An explanation of anomalously long time scales in the transient formation of forbidden lines is proposed. The concept is based on a collisionally induced density dependence of the relaxation times of metastable level populations in transient plasma. Generalization leads to an incorporation of diffusion phenomena. We demonstrate this new concept for the simplest atomic system: the He-like isoelectronic sequence. A new interpretation of the observed long duration and anomalously high intensity of spin-forbidden emission in hot plasmas is given. (author)

  4. Transient formation of forbidden lines

    Energy Technology Data Exchange (ETDEWEB)

    Rosmej, F.B. [Bochum Univ., Ruhr (Germany). Inst. fuer Experimentalphysik V; Rosmej, O.N. [VNIIFTRI, Moscow Region (Russian Federation). MISDC

    1996-05-14

    An explanation of anomalously long time scales in the transient formation of forbidden lines is proposed. The concept is based on a collisionally induced density dependence of the relaxation times of metastable level populations in transient plasma. Generalization leads to an incorporation of diffusion phenomena. We demonstrate this new concept for the simplest atomic system: the He-like isoelectronic sequence. A new interpretation of the observed long duration and anomalously high intensity of spin-forbidden emission in hot plasmas is given. (author).

  5. Reactive power compensator

    Science.gov (United States)

    El-Sharkawi, Mohamed A.; Venkata, Subrahmanyam S.; Chen, Mingliang; Andexler, George; Huang, Tony

    1992-01-01

    A system and method for determining and providing reactive power compensation for an inductive load. A reactive power compensator (50,50') monitors the voltage and current flowing through each of three distribution lines (52a, 52b, 52c), which are supplying three-phase power to one or more inductive loads. Using signals indicative of the current on each of these lines when the voltage waveform on the line crosses zero, the reactive power compensator determines a reactive power compensator capacitance that must be connected to the lines to maintain a desired VAR level, power factor, or line voltage. Alternatively, an operator can manually select a specific capacitance for connection to each line, or the capacitance can be selected based on a time schedule. The reactive power compensator produces control signals, which are coupled through optical fibers (102/106) to a switch driver (110, 110') to select specific compensation capacitors (112) for connections to each line. The switch driver develops triggering signals that are supplied to a plurality of series-connected solid state switches (350), which control charge current in one direction in respect to ground for each compensation capacitor. During each cycle, current flows from ground to charge the capacitors as the voltage on the line begins to go negative from its positive peak value. The triggering signals are applied to gate the solid state switches into a conducting state when the potential on the lines and on the capacitors reaches a negative peak value, thereby minimizing both the potential difference and across the charge current through the switches when they begin to conduct. Any harmonic distortion on the potential and current carried by the lines is filtered out from the current and potential signals used by the reactive power compensator so that it does not affect the determination of the required reactive compensation.

  6. Reactive power compensator

    Energy Technology Data Exchange (ETDEWEB)

    El-Sharkawi, Mohamed A. (Renton, WA); Venkata, Subrahmanyam S. (Woodinville, WA); Chen, Mingliang (Kirkland, WA); Andexler, George (Everett, WA); Huang, Tony (Seattle, WA)

    1992-01-01

    A system and method for determining and providing reactive power compensation for an inductive load. A reactive power compensator (50,50') monitors the voltage and current flowing through each of three distribution lines (52a, 52b, 52c), which are supplying three-phase power to one or more inductive loads. Using signals indicative of the current on each of these lines when the voltage waveform on the line crosses zero, the reactive power compensator determines a reactive power compensator capacitance that must be connected to the lines to maintain a desired VAR level, power factor, or line voltage. Alternatively, an operator can manually select a specific capacitance for connection to each line, or the capacitance can be selected based on a time schedule. The reactive power compensator produces control signals, which are coupled through optical fibers (102/106) to a switch driver (110, 110') to select specific compensation capacitors (112) for connections to each line. The switch driver develops triggering signals that are supplied to a plurality of series-connected solid state switches (350), which control charge current in one direction in respect to ground for each compensation capacitor. During each cycle, current flows from ground to charge the capacitors as the voltage on the line begins to go negative from its positive peak value. The triggering signals are applied to gate the solid state switches into a conducting state when the potential on the lines and on the capacitors reaches a negative peak value, thereby minimizing both the potential difference and across the charge current through the switches when they begin to conduct. Any harmonic distortion on the potential and current carried by the lines is filtered out from the current and potential signals used by the reactive power compensator so that it does not affect the determination of the required reactive compensation.

  7. Transient attenuation in optical fibers

    International Nuclear Information System (INIS)

    Hopkins, A.A.; Kelly, R.E.; Looney, L.D.; Lyons, P.B.

    1984-01-01

    Low and high energy pulsed electron beams were used to generate radiation-induced transient attenuation in high-OH, Suprasil core, PCS fibers, demonstrating the energy dependence of the radiation damage and recovery mechanisms. A radiation resistant low-OH fiber was studied and its performance contrasted to that of high-OH materials. Several fibers with differing core compositions were also studied

  8. Charging transient in polyvinyl formal

    Indian Academy of Sciences (India)

    Unknown

    401–406. © Indian Academy of Sciences. 401. Charging transient in polyvinyl formal. P K KHARE*, P L JAIN† and R K PANDEY‡. Department of Postgraduate Studies & Research in Physics & Electronics, Rani Durgavati University,. Jabalpur 482 001, India. †Department of Physics, Government PG College, Damoh 470 ...

  9. Transient filament stretching rheometer II

    DEFF Research Database (Denmark)

    Kolte, Mette Irene; Rasmussen, Henrik K.; Hassager, Ole

    1997-01-01

    The Lagrangian sspecification is used to simulate the transient stretching filament rheometer. Simulations are performed for dilute PIB-solutions modeled as a four mode Oldroyd-B fluid and a semidilute PIB-solution modeled as a non-linear single integral equation. The simulations are compared...

  10. Fuel rod behaviour during transients

    International Nuclear Information System (INIS)

    Hughes, H.; Haste, T.J.; Cameron, R.F.; Sinclair, J.E.

    1982-04-01

    The fuel pin performance code SLEUTH, the transient codes FRAP-T5 and TRAFIC and the clad deformation code CANSWEL-2 are described. It is shown how the codes treat gas release, pin cooling, cladding deformation and interaction, gap conductance etc. The materials properties used are indicated. (author)

  11. Simulation Model of a Transient

    DEFF Research Database (Denmark)

    Jauch, Clemens; Sørensen, Poul; Bak-Jensen, Birgitte

    2005-01-01

    This paper describes the simulation model of a controller that enables an active-stall wind turbine to ride through transient faults. The simulated wind turbine is connected to a simple model of a power system. Certain fault scenarios are specified and the turbine shall be able to sustain operati...

  12. Stationary and Transient Response Statistics

    DEFF Research Database (Denmark)

    Madsen, Peter Hauge; Krenk, Steen

    1982-01-01

    The covariance functions for the transient response of a linear MDOF-system due to stationary time limited excitation with an arbitrary frequency content are related directly to the covariance functions of the stationary response. For rational spectral density functions closed form expressions fo...

  13. Transient anisotropic magnetic field calculation

    International Nuclear Information System (INIS)

    Jesenik, Marko; Gorican, Viktor; Trlep, Mladen; Hamler, Anton; Stumberger, Bojan

    2006-01-01

    For anisotropic magnetic material, nonlinear magnetic characteristics of the material are described with magnetization curves for different magnetization directions. The paper presents transient finite element calculation of the magnetic field in the anisotropic magnetic material based on the measured magnetization curves for different magnetization directions. For the verification of the calculation method some results of the calculation are compared with the measurement

  14. Digital reactivity meter

    International Nuclear Information System (INIS)

    Copie, M.; Valantic, B.

    1978-01-01

    Digital reactivity meters (DRM) are mostly used as measuring instruments, e.g. for calibration of control rods, and there are only a few cases of their incorporation into the control systems of the reactors. To move in this direction there is more development work needed. First of all, fast algorithms are needed for inverse kinetics equations to relieve the computer for more important tasks of reactor model solving in real time. The next problem, currently under investigation, is the incorporation of the reactor thermal-hydraulic model into the DRM so that it can be used in the power range. Such an extension of DHM allows presentation not only of the instantaneous reactivity of the system, but also the inserted reactivity can be estimated from the temperature reactivity feed-backs. One of the applications of this concept is the anomalous digital reactivity monitor (ADRN) as part of the reactor protection system. As a solution of the first problem, a fast algorithm for solving the inverse kinetics equations has been implemented in the off-line program RODCAL on CDC 1700 computer and tested for its accuracy by performing different control rod calibrations on the reactor TRIGA

  15. 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

  16. LMFBR. Off normal, transient test facilities and programs in the USA

    International Nuclear Information System (INIS)

    Herbst, R.J.

    1985-01-01

    The United States fast breeder reactor development program has included operational transient analyses and experiments to verify the predicted performance of core components. Operational transient testing has focused on off-normal operation during Plant Protection System terminated transient-overpower events. In-pile and out-of-pile tests have been used to simulate predicted thermal and mechanical strain cycles and measure component response. The spectrum of reactivity ramp rates investigated in TOP tests has recently been expanded to include rates of less than $0.1/s. These slow ramp rate studies are being done in cooperation with the Japanese. The US has also cooperated with the UK in the transient testing of Prototype Fast Reactor fuel pins

  17. [Transient amnesia in the elderly].

    Science.gov (United States)

    Sellal, François

    2006-03-01

    The two main aetiologies of transient amnesia in the elderly are idiopathic transient global amnesia (TGA) and iatrogenic or toxic amnesia. Vascular and epileptic amnesia are less common. According to the literature, transient psychogenic amnesia, which is a frequent cause of amnesia at age 30 to 50, is very rare in the elderly. TGA is the prototypical picture of transient amnesia. It occurs more often after age 50, with no identified cause, even if some authors accept emotional stress or minor head trauma as occasional precipitants. The mechanism of TGA remains a matter of discussion. It may be the consequence of a spreading depression similar to that described in migraine with aura, but other arguments support an ischemic mechanism. Iatrogenic amnesias are mainly caused by benzodiazepines (BZs) or anticholinergics. The former may occur in a non-anxious subject, who is not a usual consumer of BZ and takes a single dose. The latter are more often due to a hypersensitivity to anticholinergic drugs, in particular in patients presenting with a covert, incipient Alzheimer's disease. A vascular origin must be considered when amnesia is accompanied by other neurological symptoms, and when the regression of the amnesic disorder is slow, lasting several days. It results from lesions involving various mechanisms and locations, mainly subcortical. Partial seizures, most often mesio-temporal, more rarely frontal, may be the cause of transient amnesia in the elderly, in the absence of a past history of epilepsy. The red flag supportive of an epileptic origin is the repetition of stereotyped amnesic episodes. EEG demonstration of seizures may be difficult and the response to antiepileptic drugs effective on partial seizures is usually good.

  18. Impact of seasonal forcing on reactive ecological systems.

    Science.gov (United States)

    Vesipa, Riccardo; Ridolfi, Luca

    2017-04-21

    Our focus is on the short-term dynamics of reactive ecological systems which are stable in the long term. In these systems, perturbations can exhibit significant transient amplifications before asymptotically decaying. This peculiar behavior has attracted increasing attention. However, reactive systems have so far been investigated assuming that external environmental characteristics remain constant, although environmental conditions (e.g., temperature, moisture, water availability, etc.) can undergo substantial changes due to seasonal cycles. In order to fill this gap, we propose applying the adjoint non-modal analysis to study the impact of seasonal variations of environmental conditions on reactive systems. This tool allows the transient dynamics of a perturbation affecting non-autonomous ecological systems to be described. To show the potential of this approach, a seasonally forced prey-predator model with a Holling II type functional response is studied as an exemplifying case. We demonstrate that seasonalities can greatly affect the transient dynamics of the system. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Research of the transient management in TQNPC

    International Nuclear Information System (INIS)

    Guo Longzhang; Lin Chuanqing

    2008-01-01

    Transient management is the basic technical subject in nuclear power plant. Since the Third Qinshan nuclear power company (TQNPC) successful completes the commissioning in 2003, the transient management work start at the transient management item selection and the flow definition. Now TQNPC have a complete transient management system and the management flow. In the last two years, TNQPC have finished the historic transient data collection for two units, and confirmed that the plant's key systems and equipments are at safe state. The development of the transient management subject would build a reliable foundation for the plant safe operation, plant lifetime management and periodic safety review. (author)

  20. Power and power-to-flow reactivity transfer functions in EBR-II [Experimental Breeder Reactor II] fuel

    International Nuclear Information System (INIS)

    Grimm, K.N.; Meneghetti, D.

    1989-01-01

    Reactivity transfer functions are important in determining the reactivity history during a power transient. Overall nodal transfer functions have been calculated for different subassembly types in the Experimental Breeder Reactor II (EBR-II). Steady-state calculations for temperature changes and, hence, reactivities for power changes have been separated into power and power-to-flow-dependent terms. Axial nodal transfer functions separated into power and power-to-flow-dependent components are reported in this paper for a typical EBR-II fuel pin. This provides an improved understanding of the time dependence of these components in transient situations

  1. Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

    International Nuclear Information System (INIS)

    Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras

    2003-01-01

    This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

  2. BARS - a heterogeneous code for 3D pin-by-pin LWR steady-state and transient calculation

    International Nuclear Information System (INIS)

    Avvakumov, A.V.; Malofeev, V.M.

    2000-01-01

    A 3D pin-by-pin dynamic model for LWR detailed calculation was developed. The model is based on a coupling of the BARS neutronic code with the RELAP5/MOD3.2 thermal hydraulic code. This model is intended to calculate a fuel cycle, a xenon transient, and a wide range of reactivity initiated accidents in a WWER and a PWR. Galanin-Feinberg heterogeneous method was realized in the BARS code. Some results for a validation of the heterogeneous method are presented for reactivity coefficients, a pin-by-pin power distribution, and a fast pulse transient. (Authors)

  3. Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

    Science.gov (United States)

    Clamens, Olivier; Lecerf, Johann; Hudelot, Jean-Pascal; Duc, Bertrand; Cadiou, Thierry; Blaise, Patrick; Biard, Bruno

    2018-01-01

    CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN) and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He) situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

  4. Assessment of the 3He pressure inside the CABRI transient rods - Development of a surrogate model based on measurements and complementary CFD calculations

    Directory of Open Access Journals (Sweden)

    Clamens Olivier

    2018-01-01

    Full Text Available CABRI is an experimental pulse reactor, funded by the French Nuclear Safety and Radioprotection Institute (IRSN and operated by CEA at the Cadarache research center. It is designed to study fuel behavior under RIA conditions. In order to produce the power transients, reactivity is injected by depressurization of a neutron absorber (3He situated in transient rods inside the reactor core. The shapes of power transients depend on the total amount of reactivity injected and on the injection speed. The injected reactivity can be calculated by conversion of the 3He gas density into units of reactivity. So, it is of upmost importance to properly master gas density evolution in transient rods during a power transient. The 3He depressurization was studied by CFD calculations and completed with measurements using pressure transducers. The CFD calculations show that the density evolution is slower than the pressure drop. Surrogate models were built based on CFD calculations and validated against preliminary tests in the CABRI transient system. Studies also show that it is harder to predict the depressurization during the power transients because of neutron/3He capture reactions that induce a gas heating. This phenomenon can be studied by a multiphysics approach based on reaction rate calculation thanks to Monte Carlo code and study the resulting heating effect with the validated CFD simulation.

  5. Nuclear power plant transients: where are we

    International Nuclear Information System (INIS)

    Majumdar, D.

    1984-05-01

    This document is in part a postconference review and summary of the American Nuclear Society sponsored Anticipated and Abnormal Plant Transients in Light Water Reactors Conference held in Jackson, Wyoming, September 26-29, 1983, and in part a reflection upon the issues of plant transients and their impact on the viability of nuclear power. This document discusses state-of-the-art knowledge, deficiencies, and future directions in the plant transients area as seen through this conference. It describes briefly what was reported in this conference, emphasizes areas where it is felt there is confidence in the nuclear industry, and also discusses where the experts did not have a consensus. Areas covered in the document include major issues in operational transients, transient management, transient events experience base, the status of the analytical tools and their capabilities, probabilistic risk assessment applications in operational transients, and human factors impact on plant transients management

  6. Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

    International Nuclear Information System (INIS)

    Ortensi, Javier; Baker, Benjamin; Wang, Yaqi; Schunert, Sebastian; DeHart, Mark

    2017-01-01

    This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$ 2 $, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the

  7. Transient Simulation of the Multi-SERTTA Experiment with MAMMOTH

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); deHart, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-11

    This work details the MAMMOTH reactor physics simulations of the Static Environment Rodlet Transient Test Apparatus (SERTTA) conducted at Idaho National Laboratory in FY-2017. TREAT static-environment experiment vehicles are being developed to enable transient testing of Pressurized Water Reactor (PWR) type fuel specimens, including fuel concepts with enhanced accident tolerance (Accident Tolerant Fuels, ATF). The MAMMOTH simulations include point reactor kinetics as well as spatial dynamics for a temperature-limited transient. The strongly coupled multi-physics solutions of the neutron flux and temperature fields are second order accurate both in the spatial and temporal domains. MAMMOTH produces pellet stack powers that are within 1.5% of the Monte Carlo reference solutions. Some discrepancies between the MCNP model used in the design of the flux collars and the Serpent/MAMMOTH models lead to higher power and energy deposition values in Multi-SERTTA unit 1. The TREAT core results compare well with the safety case computed with point reactor kinetics in RELAP5-3D. The reactor period is 44 msec, which corresponds to a reactivity insertion of 2.685% delta k/k$. The peak core power in the spatial dynamics simulation is 431 MW, which the point kinetics model over-predicts by 12%. The pulse width at half the maximum power is 0.177 sec. Subtle transient effects are apparent at the beginning insertion in the experimental samples due to the control rod removal. Additional difference due to transient effects are observed in the sample powers and enthalpy. The time dependence of the power coupling factor (PCF) is calculated for the various fuel stacks of the Multi-SERTTA vehicle. Sample temperatures in excess of 3100 K, the melting point UO$_2$, are computed with the adiabatic heat transfer model. The planned shaped-transient might introduce additional effects that cannot be predicted with PRK models. Future modeling will be focused on the shaped-transient by improving the

  8. Spring 5 & reactive streams

    CERN Multimedia

    CERN. Geneva; Clozel, Brian

    2017-01-01

    Spring is a framework widely used by the world-wide Java community, and it is also extensively used at CERN. The accelerator control system is constituted of 10 million lines of Java code, spread across more than 1000 projects (jars) developed by 160 software engineers. Around half of this (all server-side Java code) is based on the Spring framework. Warning: the speakers will assume that people attending the seminar are familiar with Java and Spring’s basic concepts. Spring 5.0 and Spring Boot 2.0 updates (45 min) This talk will cover the big ticket items in the 5.0 release of Spring (including Kotlin support, @Nullable and JDK9) and provide an update on Spring Boot 2.0, which is scheduled for the end of the year. Reactive Spring (1h) Spring Framework 5.0 has been released - and it now supports reactive applications in the Spring ecosystem. During this presentation, we'll talk about the reactive foundations of Spring Framework with the Reactor project and the reactive streams specification. We'll al...

  9. Reactivity of nitriles

    International Nuclear Information System (INIS)

    Kukushkin, Yu.N.

    1987-01-01

    Reactivity of coordination nitriles in transition metal (Ru, Mo, W, Zr, Hf) complexes, namely: transformation of nitriles of the first coordination sphere into N-acyl-substituted amides, amidines, nitrile interaction; with water, alkalines, alcoholes, hydrogen, azide and cyanide ions is considered. Introduction of acetonitrile molecule to uranium (4)-carbon double bond is discussed

  10. Clojure reactive programming

    CERN Document Server

    Borges, Leonardo

    2015-01-01

    If you are a Clojure developer who is interested in using Reactive Programming to build asynchronous and concurrent applications, this book is for you. Knowledge of Clojure and Leiningen is required. Basic understanding of ClojureScript will be helpful for the web chapters, although it is not strictly necessary.

  11. A Universal Reactive Machine

    DEFF Research Database (Denmark)

    Andersen, Henrik Reif; Mørk, Simon; Sørensen, Morten U.

    1997-01-01

    Turing showed the existence of a model universal for the set of Turing machines in the sense that given an encoding of any Turing machine asinput the universal Turing machine simulates it. We introduce the concept of universality for reactive systems and construct a CCS processuniversal...

  12. Chemical Reactivity Test (CRT)

    Energy Technology Data Exchange (ETDEWEB)

    Zaka, F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-12-13

    The Chemical Reactivity Test (CRT) is used to determine the thermal stability of High Explosives (HEs) and chemical compatibility between (HEs) and alien materials. The CRT is one of the small-scale safety tests performed on HE at the High Explosives Applications Facility (HEAF).

  13. Reactive power compensating system

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Timothy J. (Redondo Beach, CA); El-Sharkawi, Mohamed A. (Renton, WA); Venkata, Subrahmanyam S. (Seattle, WA)

    1987-01-01

    The reactive power of an induction machine is compensated by providing fixed capacitors on each phase line for the minimum compensation required, sensing the current on one line at the time its voltage crosses zero to determine the actual compensation required for each phase, and selecting switched capacitors on each line to provide the balance of the compensation required.

  14. Reactive Power Compensating System.

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Timothy J.; El-Sharkawi, Mohamed A.; Venkata, Subrahmanyam S.

    1985-01-04

    The circuit was designed for the specific application of wind-driven induction generators. It has great potential for application in any situation where a varying reactive power load is present, such as with induction motors or generators, or for transmission network compensation.

  15. The iodine reactivity

    International Nuclear Information System (INIS)

    2003-01-01

    The iodine is an important element because it has long life isotopes (such as iodine 129) and a great mobility in natural media. Iodine presents a complex chemistry because of its volatility and its strong redox reactivity. The S.E.C.R. works to better understand the reactivity of this element in different natural, industrial or biological environments. It plays a part in thermochemical sites as a possible way of hydrogen formation. This seminar gives some aspects relative to the chemical reactivity of iodine, since its thermochemistry in the I/S cycles to produce hydrogen to its reactivity in the natural medium and its potential radiological impact. This document includes 4 presentations transparencies) dealing with: the 129 I cycle rejected in the low radioactive gaseous and liquid effluents of the La Hague reprocessing plant (C. Frechou); a bibliographic review of iodine retention in soils (F. Bazer-Bachi); the hydrogen production and the iodine/sulfur thermochemical cycle (role of iodine in the process); and the direct characterization by electro-spray ionization mass spectroscopy of iodine fixation by fulvic acids (P. Reiller, B. Amekraz, C. Moulin, V. Moulin)

  16. Upscaling of reactive flows

    NARCIS (Netherlands)

    Kumar, K.

    2012-01-01

    The thesis deals with the upscaling of reactive flows in complex geometry. The reactions which may include deposition or dissolution take place at a part of the boundary and depending on the size of the reaction domain, the changes in the pore structure that are due to the deposition process may or

  17. Fast Flux Test Facility (FFTF) feedback reactivity components

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1988-04-01

    The static tests conducted during Cycle 8A (1986) of the FFTF have allowed, for the first time, the experimental determination of each of the feedback reactivities caused by the following mechanisms: fuel axial expansion, control rod repositioning, core radial expansion, and subassembly bowing. A semiempirical equation was obtained to describe each of these feedback components that depended only on the relevant reactor temperature (bowing was presented in a tabular form). The Doppler and sodium density reactivities were calculated using existing mechanistic methods. Although they could also be fitted with closed-form equations depending only on temperatures, these equations are not needed in transient analyses using whole core safety computer codes, which use mechanistic methods. The static feedback reactivity model was extended to obtain a dynamic model via the concept of ''time constants.'' Besides being used for transient analyses in the FFTF, these feedback equations constitute a database for the validation and/or calibration of mechanistic feedback reactivity models. 2 refs., 6 tabs

  18. Nonlinear Diffusion and Transient Osmosis

    International Nuclear Information System (INIS)

    Igarashi, Akira; Rondoni, Lamberto; Botrugno, Antonio; Pizzi, Marco

    2011-01-01

    We investigate both analytically and numerically the concentration dynamics of a solution in two containers connected by a narrow and short channel, in which diffusion obeys a porous medium equation. We also consider the variation of the pressure in the containers due to the flow of matter in the channel. In particular, we identify a phenomenon, which depends on the transport of matter across nano-porous membranes, which we call ''transient osmosis . We find that nonlinear diffusion of the porous medium equation type allows numerous different osmotic-like phenomena, which are not present in the case of ordinary Fickian diffusion. Experimental results suggest one possible candidate for transiently osmotic processes. (electromagnetism, optics, acoustics, heat transfer, classical mechanics, and fluid dynamics)

  19. Fast thermal transients on valve

    International Nuclear Information System (INIS)

    Ferjancic, M.; Stok, B.; Halilovic, M.; Koc, P.; Mole, N.; Otrin, Z.; Kotar, A.

    2007-01-01

    One of the regulatory body methods to supervise nuclear safety of a nuclear power plant is a review of plant modifications and evaluation of their impact on plant operating experience. The Slovenian Nuclear Safety Administration (SNSA) licensed in April 2003 the use of leak-before-break (LBB) methodology in the Krsko NPP for the primary loop including surge line and connecting pipelines with minimal diameter of 6 inch. The SNSA decision based also on fracture mechanics analyses that include direct pipe failure mechanisms such as water hammer, creep damage, erosion and corrosion, fatigue and environmental conditions over the entire life of the plant. The evaluation of the operating transients pointed out, that presumed loadings, used for the LBB analysis, did not incorporate all the fast thermal transients data. For that purpose the SNSA requested Faculty of Mechanical Engineering (FS) in Ljubljana to perform additional analyses. The results of the analysis shall confirm the validity of the LBB analysis. (author)

  20. Pressure transient in liquid lines

    International Nuclear Information System (INIS)

    Sun, J.G.; Wang, X.Q.

    1995-01-01

    The pressure surge that results from a step change of flow in liquid pipelines, commonly known as water hammer, was analyzed by an eigenfunction method. A differential-integral Pressure wave equation and a linearized velocity equation were derived from the equations of mass and momentum conservation. Waveform distortion due to viscous dissipation and pipe-wall elastic expansion is characterized by a dimensionless transmission number K. The pressure surge condition, which is mathematically singular, was used in the solution procedure. The exact solutions from numerical calculation of the differential-integral equation provide a complete Pressure transient in the pipe. The problems are also calculated With the general-purpose computer code COMMIX, which solves the exact mass conservation equation and Navier-Stokes equations. These solutions were compared with published experimental results, and agreement was good. The effect of turbulence on the pressure transient is discussed in the light of COMMIX calculational results

  1. Superresolution microscopy with transient binding.

    Science.gov (United States)

    Molle, Julia; Raab, Mario; Holzmeister, Susanne; Schmitt-Monreal, Daniel; Grohmann, Dina; He, Zhike; Tinnefeld, Philip

    2016-06-01

    For single-molecule localization based superresolution, the concentration of fluorescent labels has to be thinned out. This is commonly achieved by photophysically or photochemically deactivating subsets of molecules. Alternatively, apparent switching of molecules can be achieved by transient binding of fluorescent labels. Here, a diffusing dye yields bright fluorescent spots when binding to the structure of interest. As the binding interaction is weak, the labeling is reversible and the dye ligand construct diffuses back into solution. This approach of achieving superresolution by transient binding (STB) is reviewed in this manuscript. Different realizations of STB are discussed and compared to other localization-based superresolution modalities. We propose the development of labeling strategies that will make STB a highly versatile tool for superresolution microscopy at highest resolution. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. Electromagnetic Transients in Power Cables

    DEFF Research Database (Denmark)

    Silva, Filipe Faria Da; Bak, Claus Leth

    . The chapter ends by proposing a systematic method that can be used when doing the insulation co-ordination study for a line, as well as the modelling requirements, both modelling depth and modelling detail of the equipment, for the study of the different types of transients followed by a step-by-step generic...... typically used for the screens of cables (both-ends bonding and cross-boding) and also presents methods that can be used to estimate the maximum current of a cable for different types of soils, i.e. thermal calculations. The end of the chapter introduces the shunt reactor, which is an important element...... detail of the equipment, for the study of the different types of transients followed by a step-by-step generic example....

  3. Rap1 signaling is required for suppression of Ras-generated reactive oxygen species and protection against oxidative stress in T lymphocytes

    NARCIS (Netherlands)

    Remans, Philip H. J.; Gringhuis, Sonja I.; van Laar, Jacob M.; Sanders, Marjolein E.; Papendrecht-van der Voort, Ellen A. M.; Zwartkruis, Fried J. T.; Levarht, E. W. Nivine; Rosas, Marcela; Coffer, Paul J.; Breedveld, Ferdinand C.; Bos, Johannes L.; Tak, Paul P.; Verweij, Cornelis L.; Reedquist, Kris A.

    2004-01-01

    Transient production of reactive oxygen species (ROS) plays an important role in optimizing transcriptional and proliferative responses to TCR signaling in T lymphocytes. Conversely, chronic oxidative stress leads to decreased proliferative responses and enhanced transcription of inflammatory gene

  4. Features of Onset and Clinical Course of Reactive Arthritis in Children

    Directory of Open Access Journals (Sweden)

    I.S. Lebets

    2013-09-01

    Results. Reactive arthritis of chlamydial etiology is characterized by lesion of large and medium-sized joints of the lower limbs, which is often accompanied by short-term morning stiffness and rapid onset of transient hypomyatrophy. Reiter’s disease may develop rarely. Mycoplasma-induced reactive arthritis is characterized by debut with arthritis of knee, ankle, wrist and small joints of the hand, the development of bursitis and hypomyatrophy. Feature of Ureaplasma arthritis is the formation of bursitis in the heel and tendinitis. Reactive arthritis associated with elevated titers to antistreptolysin O differs with polymorphism of articular syndrome manifestations and, to some extent, of similarity with juvenile rheumatoid arthritis. Unspecified reactive arthritis has a number of the general features with others reactive arthritis and it is characterized by rather benign clinical course, long preservation of joints function and low laboratory activity. Relapse rate of reactive arthritis increases with an increase of duration of illness.

  5. Techniques for computing reactivity changes caused by fuel axial expansion in LMR's

    International Nuclear Information System (INIS)

    Khalil, H.

    1988-01-01

    An evaluation is made of the accuracy of methods used to compute reactivity changes caused by axial fuel relocation in fast reactors. Results are presented to demonstrate the validity of assumptions commonly made such as linearity of reactivity with fuel elongation, additivity of local reactivity contributions, and the adequacy of standard perturbation techniques. Accurate prediction of the reactivity loss caused by axial swelling of metallic fuel is shown to require proper representation of the burnup dependence of the expansion reactivity. Some accuracy limitations in the methods used in transient analyses, which are based on the use of fuel worth tables, are identified, and efficient ways to improve accuracy are described. Implementation of these corrections produced expansion reactivity estimates within 5% of higher-order method for a metal-fueled FFTF core representation. 18 refs., 3 figs., 3 tabs

  6. LLL transient-electromagnetics-measurement facility

    International Nuclear Information System (INIS)

    Deadrick, F.J.; Miller, E.K.; Hudson, H.G.

    1975-01-01

    The operation and hardware of the Lawrence Livermore Laboratory's transient-electromagnetics (EM)-measurement facility are described. The transient-EM range is useful for determining the time-domain transient responses of structures to incident EM pulses. To illustrate the accuracy and utility of the EM-measurement facility, actual experimental measurements are compared to numerically computed values

  7. Learning from anticipated and abnormal plant transients

    International Nuclear Information System (INIS)

    Varnado, B.

    1983-01-01

    A report is given of the American Nuclear Society topical meeting on Anticipated and Abnormal Transients in Light Water Reactors held in Jackson, Wyoming in September 1983. Industry involvement in the evaluation of operating experience, human error contributions, transient management, thermal hydraulic modelling, the role of probabilistic risk assessment and the cost of transient incidents are discussed. (U.K.)

  8. A transient absorption study of allophycocyanin

    Indian Academy of Sciences (India)

    Transient dynamics of allophycocyanin trimers and monomers are observed by using the pump-probe, transient absorption technique. The origin of spectral components of the transient absorption spectra is discussed in terms of both kinetics and spectroscopy. We find that the energy gap between the ground and excited ...

  9. Adaptive sampling of AEM transients

    Science.gov (United States)

    Di Massa, Domenico; Florio, Giovanni; Viezzoli, Andrea

    2016-02-01

    This paper focuses on the sampling of the electromagnetic transient as acquired by airborne time-domain electromagnetic (TDEM) systems. Typically, the sampling of the electromagnetic transient is done using a fixed number of gates whose width grows logarithmically (log-gating). The log-gating has two main benefits: improving the signal to noise (S/N) ratio at late times, when the electromagnetic signal has amplitudes equal or lower than the natural background noise, and ensuring a good resolution at the early times. However, as a result of fixed time gates, the conventional log-gating does not consider any geological variations in the surveyed area, nor the possibly varying characteristics of the measured signal. We show, using synthetic models, how a different, flexible sampling scheme can increase the resolution of resistivity models. We propose a new sampling method, which adapts the gating on the base of the slope variations in the electromagnetic (EM) transient. The use of such an alternative sampling scheme aims to get more accurate inverse models by extracting the geoelectrical information from the measured data in an optimal way.

  10. Transient virulence of emerging pathogens.

    Science.gov (United States)

    Bolker, Benjamin M; Nanda, Arjun; Shah, Dharmini

    2010-05-06

    Should emerging pathogens be unusually virulent? If so, why? Existing theories of virulence evolution based on a tradeoff between high transmission rates and long infectious periods imply that epidemic growth conditions will select for higher virulence, possibly leading to a transient peak in virulence near the beginning of an epidemic. This transient selection could lead to high virulence in emerging pathogens. Using a simple model of the epidemiological and evolutionary dynamics of emerging pathogens, along with rough estimates of parameters for pathogens such as severe acute respiratory syndrome, West Nile virus and myxomatosis, we estimated the potential magnitude and timing of such transient virulence peaks. Pathogens that are moderately evolvable, highly transmissible, and highly virulent at equilibrium could briefly double their virulence during an epidemic; thus, epidemic-phase selection could contribute significantly to the virulence of emerging pathogens. In order to further assess the potential significance of this mechanism, we bring together data from the literature for the shapes of tradeoff curves for several pathogens (myxomatosis, HIV, and a parasite of Daphnia) and the level of genetic variation for virulence for one (myxomatosis). We discuss the need for better data on tradeoff curves and genetic variance in order to evaluate the plausibility of various scenarios of virulence evolution.

  11. Cortical computations via transient attractors.

    Directory of Open Access Journals (Sweden)

    Oliver L C Rourke

    Full Text Available The ability of sensory networks to transiently store information on the scale of seconds can confer many advantages in processing time-varying stimuli. How a network could store information on such intermediate time scales, between typical neurophysiological time scales and those of long-term memory, is typically attributed to persistent neural activity. An alternative mechanism which might allow for such information storage is through temporary modifications to the neural connectivity which decay on the same second-long time scale as the underlying memories. Earlier work that has explored this method has done so by emphasizing one attractor from a limited, pre-defined set. Here, we describe an alternative, a Transient Attractor network, which can learn any pattern presented to it, store several simultaneously, and robustly recall them on demand using targeted probes in a manner reminiscent of Hopfield networks. We hypothesize that such functionality could be usefully embedded within sensory cortex, and allow for a flexibly-gated short-term memory, as well as conferring the ability of the network to perform automatic de-noising, and separation of input signals into distinct perceptual objects. We demonstrate that the stored information can be refreshed to extend storage time, is not sensitive to noise in the system, and can be turned on or off by simple neuromodulation. The diverse capabilities of transient attractors, as well as their resemblance to many features observed in sensory cortex, suggest the possibility that their actions might underlie neural processing in many sensory areas.

  12. Cortical computations via transient attractors.

    Science.gov (United States)

    Rourke, Oliver L C; Butts, Daniel A

    2017-01-01

    The ability of sensory networks to transiently store information on the scale of seconds can confer many advantages in processing time-varying stimuli. How a network could store information on such intermediate time scales, between typical neurophysiological time scales and those of long-term memory, is typically attributed to persistent neural activity. An alternative mechanism which might allow for such information storage is through temporary modifications to the neural connectivity which decay on the same second-long time scale as the underlying memories. Earlier work that has explored this method has done so by emphasizing one attractor from a limited, pre-defined set. Here, we describe an alternative, a Transient Attractor network, which can learn any pattern presented to it, store several simultaneously, and robustly recall them on demand using targeted probes in a manner reminiscent of Hopfield networks. We hypothesize that such functionality could be usefully embedded within sensory cortex, and allow for a flexibly-gated short-term memory, as well as conferring the ability of the network to perform automatic de-noising, and separation of input signals into distinct perceptual objects. We demonstrate that the stored information can be refreshed to extend storage time, is not sensitive to noise in the system, and can be turned on or off by simple neuromodulation. The diverse capabilities of transient attractors, as well as their resemblance to many features observed in sensory cortex, suggest the possibility that their actions might underlie neural processing in many sensory areas.

  13. TRAB, a transient analysis program for BWR. Part 1

    International Nuclear Information System (INIS)

    Rajamaeki, Markku.

    1980-03-01

    TRAB is a transient analysis program for BWR. The present report describes its principles. The program has been developed from TRAWA-program. It models the interior of the pressure vessel and related subsystems of BWR viz. reactor core, recirculation loop including the upper part of the vessel, recirculation pumps, incoming and outgoing flow systems, and control and protection systems. Concerning core phenomena and all flow channel hydraulics the submodels are one-dimensional of main features. The geometry is very flexible. The program has been made particularly to simulate various reactivity transients, but it is applicable more generally to reactor incidents and accidents in which no flow reversal or no emptying of the circuit must occur below the water level. The program is extensively supplied by input and output capabilities. The user can act upon the simulation of a transient by defining external disturbances, scheduled timevariations for any system variable, by modeling new subsystems, which are representable with ordinary linear differential equations, and by defining relations of functional form between system variables. The run of the program can be saved and restarted. (author)

  14. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  15. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  16. Reactivity balance for a soluble boron-free small modular reactor

    Directory of Open Access Journals (Sweden)

    Lezani van der Merwe

    2018-06-01

    Full Text Available Elimination of soluble boron from reactor design eliminates boron-induced reactivity accidents and leads to a more negative moderator temperature coefficient. However, a large negative moderator temperature coefficient can lead to large reactivity feedback that could allow the reactor to return to power when it cools down from hot full power to cold zero power. In soluble boron-free small modular reactor (SMR design, only control rods are available to control such rapid core transient.The purpose of this study is to investigate whether an SMR would have enough control rod worth to compensate for large reactivity feedback. The investigation begins with classification of reactivity and completes an analysis of the reactivity balance in each reactor state for the SMR model.The control rod worth requirement obtained from the reactivity balance is a minimum control rod worth to maintain the reactor critical during the whole cycle. The minimum available rod worth must be larger than the control rod worth requirement to manipulate the reactor safely in each reactor state. It is found that the SMR does have enough control rod worth available during rapid transient to maintain the SMR at subcritical below k-effectives of 0.99 for both hot zero power and cold zero power. Keywords: Control Rod Worth, Reactivity Balance, Reactivity Feedback, Small Modular Reactor, Soluble Boron Free

  17. Immune reactivities against gums.

    Science.gov (United States)

    Vojdani, Aristo; Vojdani, Charlene

    2015-01-01

    Different kinds of gums from various sources enjoy an extremely broad range of commercial and industrial use, from food and pharmaceuticals to printing and adhesives. Although generally recognized as safe by the US Food and Drug Administration (FDA), gums have a history of association with sensitive or allergic reactions. In addition, studies have shown that gums have a structural, molecular similarity to a number of common foods. A possibility exists for cross-reactivity. Due to the widespread use of gums in almost every aspect of modern life, the overall goal of the current investigation was to determine the degree of immune reactivity to various gum antigens in the sera of individuals representing the general population. The study was a randomized, controlled trial. 288 sera purchased from a commercial source. The sera was screened for immunoglobulin G (IgG) and immunoglobulin E (IgE) antibodies against extracts of mastic gum, carrageenan, xantham gum, guar gum, gum tragacanth, locust bean gum, and β-glucan, using indirect enzyme-linked immunosorbent assay (ELISA) testing. For each gum antigen, inhibition testing was performed on the 4 sera that showed the highest IgG and IgE immune reactivity against the different gums used in the study. Inhibition testing on these same sera for sesame albumin, lentil, corn, rice, pineapple, peanut, pea protein, shrimp, or kidney bean was used to determine the cross-reactivity of these foods with the gum. Of the 288 samples, 4.2%-27% of the specimens showed a significant elevation in IgG antibodies against various gums. Only 4 of 288, or 1.4%, showed a simultaneous elevation of the IgG antibody against all 7 gum extracts. For the IgE antibody, 15.6%-29.1% of the specimens showed an elevation against the various gums. A significant percentage of the specimens, 12.8%, simultaneously produced IgE antibodies against all 7 tested extracts. Overall, the percentage of elevation in IgE antibodies against different gum extracts, with

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. Bearing for the reactivation

    International Nuclear Information System (INIS)

    Santamaria Alexandra

    2003-01-01

    Ecopetrol undertook an aggressive plan to reactivate the activities of seismic that allows fulfilling the goals proposed for this year (2003). Although the production registered a descent of 9%, the financial results throw utilities for $1.1 trillion pesos to the closing of September and contributions in bonuses for $1.2 trillions. The author also refers to the general balance, to the finances, raw production, taxes and transfers

  20. Transient hyperthyroidism of hyperemesis gravidarum.

    Science.gov (United States)

    Tan, Jackie Y L; Loh, Keh Chuan; Yeo, George S H; Chee, Yam Cheng

    2002-06-01

    To characterise the clinical, biochemical and thyroid antibody profile in women with transient hyperthyroidism of hyperemesis gravidarum. Prospective observational study. Hospital inpatient gynaecological ward. Women admitted with hyperemesis gravidarum and found to have hyperthyroidism. Fifty-three women were admitted with hyperemesis gravidarum and were found to have hyperthyroidism. Each woman was examined for clinical signs of thyroid disease and underwent investigations including urea, creatinine, electrolytes, liver function test, thyroid antibody profile and serial thyroid function test until normalisation. Gestation at which thyroid function normalised, clinical and thyroid antibody profile and pregnancy outcome (birthweight, gestation at delivery and Apgar score at 5 minutes). Full data were available for 44 women. Free T4 levels normalised by 15 weeks of gestation in the 39 women with transient hyperthyroidism while TSH remained suppressed until 19 weeks of gestation. None of these women were clinically hyperthyroid. Thyroid antibodies were not found in most of them. Median birthweight in the infants of mothers who experienced weight loss of > 5% of their pre-pregnancy weight was lower compared with those of women who did not (P = 0.093). Five women were diagnosed with Graves' disease based on clinical features and thyroid antibody profile. In transient hyperthyroidism of hyperemesis gravidarum, thyroid function normalises by the middle of the second trimester without anti-thyroid treatment. Clinically overt hyperthyroidism and thyroid antibodies are usually absent. Apart from a non-significant trend towards lower birthweights in the infants of mothers who experienced significant weight loss, pregnancy outcome was generally good. Routine assessment of thyroid function is unnecessary for women with hyperemesis gravidarum in the absence of any clinical features of hyperthyroidism.

  1. Wind Power Impact to Transient and Voltage Stability of the Power System in Eastern Denmark

    DEFF Research Database (Denmark)

    Rasmussen, Joana; Jørgensen, Preben; Palsson, Magni Thor

    2005-01-01

    Voltage stability, transient stability and reactive power compensation are extremely important issues for largescale integration of wind power in areas distant from the main transmission system in Eastern Denmark. This paper describes the application of a dynamic wind farm model in simulation...... studies for assessments of a large wind power penetration. The simulation results reveal problems with voltage stability due to the characteristic of wind turbine generation as well as the inability of the power system to meet the reactive power demand. Furthermore, the established model is applied...

  2. Entropy-based critical reaction time for mixing-controlled reactive transport

    DEFF Research Database (Denmark)

    Chiogna, Gabriele; Rolle, Massimo

    2017-01-01

    Entropy-based metrics, such as the dilution index, have been proposed to quantify dilution and reactive mixing in solute transport problems. In this work, we derive the transient advection dispersion equation for the entropy density of a reactive plume. We restrict our analysis to the case where...... the concentration distribution of the transported species is Gaussian and we observe that, even in case of an instantaneous complete bimolecular reaction, dilution caused by dispersive processes dominates the entropy balance at early times and results in the net increase of the entropy density of a reactive species...

  3. Programming Reactive Extensions and LINQ

    CERN Document Server

    Liberty, Jesse

    2011-01-01

    Pro Reactive Extensions and LINQ is a deep dive into the next important technology for .NET developers: Reactive Extensions. This in-depth tutorial goes beyond what is available anywhere else to teach how to write WPF, Silverlight, and Windows Phone applications using the Reactive Extensions (Rx) to handle events and asynchronous method calls. Reactive programming allows you to turn those aspects of your code that are currently imperative into something much more event-driven and flexible. For this reason, it's sometimes referred to as LINQ for Events. Reactive programming hinges on the concep

  4. Unified Behavior Framework for Reactive Robot Control in Real-Time Systems

    Science.gov (United States)

    2007-03-01

    maintain coherent operation in concurrent programs by employing standard communication and synchronization patterns. Some typical ones are: semaphores ...through the semaphore . Signals, whether persistent or transient, are used to communicate between threads as a means of synchronizing their progress...tasks to be decomposed into collections of low-level primitive behaviors, Figure 2.b. This approach takes on the self- contradictory term, reactive

  5. Hidden photoinduced reactivity of the blue fluorescent protein mKalama1

    DEFF Research Database (Denmark)

    Vegh, Russell B.; Bloch, Dmitry A.; Bommarius, Andreas S.

    2015-01-01

    , is largely unexplored. Here, by using transient absorption spectroscopy spanning the time scale from picoseconds to seconds, we reveal a hidden reactivity of the bright blue-light emitting protein mKalama1 previously thought to be inert. This protein shows no excited-state proton transfer during its...

  6. The course of transient hypochondriasis.

    Science.gov (United States)

    Barsky, A J; Cleary, P D; Sarnie, M K; Klerman, G L

    1993-03-01

    This study examined the longitudinal course of patients known to have had a previous episode of transient hypochondriasis. Twenty-two transiently hypochondriacal patients and 24 nonhypochondriacal patients from the same general medical clinic were reexamined after an average of 22 months with the use of self-report questionnaires, structured diagnostic interviews, and medical record review. The hypochondriacal patients continued to manifest significantly more hypochondriacal symptoms, more somatization, and more psychopathological symptoms at follow-up. They also reported significantly more amplification of bodily sensations and more functional disability and utilized more medical care. These differences persisted after control for differences in medical morbidity and marital status. Only one hypochondriacal patient, however, had a DSM-III-R diagnosis of hypochondriasis at follow-up. Multivariate analyses revealed that the only significant predictors of hypochondriacal symptoms at follow-up were hypochondriacal symptoms and the tendency to amplify bodily sensations at the baseline evaluation. Hypochondriacal symptoms appear to have some temporal stability: patients who experienced hypochondriacal episodes at the beginning of the study were significantly more hypochondriacal 2 years later than comparison patients. They were not, however, any more likely to develop DSM-III-R-defined hypochondriasis. Thus, hypochondriacal symptoms may be distinct from the axis I disorder. The data are also compatible with the hypothesis that preexisting amplification of bodily sensations is an important predictor of subsequent hypochondriacal symptoms.

  7. Transient global amnesia: current perspectives

    Directory of Open Access Journals (Sweden)

    Spiegel DR

    2017-10-01

    Full Text Available David R Spiegel, Justin Smith, Ryan R Wade, Nithya Cherukuru, Aneel Ursani, Yuliya Dobruskina, Taylor Crist, Robert F Busch, Rahim M Dhanani, Nicholas Dreyer Department of Psychiatry and Behavioral Sciences, Eastern Virginia Medical School, Norfolk, VA, USA Abstract: Transient global amnesia (TGA is a clinical syndrome characterized by the sudden onset of an extraordinarily large reduction of anterograde and a somewhat milder reduction of retrograde episodic long-term memory. Additionally, executive functions are described as diminished. Although it is suggested that various factors, such as migraine, focal ischemia, venous flow abnormalities, and epileptic phenomena, are involved in the pathophysiology and differential diagnosis of TGA, the factors triggering the emergence of these lesions are still elusive. Recent data suggest that the vulnerability of CA1 neurons to metabolic stress plays a pivotal part in the pathophysiological cascade, leading to an impairment of hippocampal function during TGA. In this review, we discuss clinical aspects, new imaging findings, and recent clinical–epidemiological data with regard to the phenotype, functional anatomy, and putative cellular mechanisms of TGA. Keywords: transient global amnesia, vascular, migraines, psychiatric

  8. Transient trimethylaminuria related to menstruation

    Science.gov (United States)

    Shimizu, Makiko; Cashman, John R; Yamazaki, Hiroshi

    2007-01-01

    Background Trimethylaminuria, or fish odor syndrome, includes a transient or mild malodor caused by an excessive amount of malodorous trimethylamine as a result of body secretions. Herein, we describe data to support the proposal that menses can be an additional factor causing transient trimethylaminuria in self-reported subjects suffering from malodor and even in healthy women harboring functionally active flavin-containing monooxygenase 3 (FMO3). Methods FMO3 metabolic capacity (conversion of trimethylamine to trimethylamine N-oxide) was defined as the urinary ratio of trimethylamine N-oxide to total trimethylamine. Results Self-reported Case (A) that was homozygous for inactive Arg500stop FMO3, showed decreased metabolic capacity of FMO3 (i.e., ~10% the unaffected metabolic capacity) during 120 days of observation. For Case (B) that was homozygous for common [Glu158Lys; Glu308Gly] FMO3 polymorphisms, metabolic capacity of FMO3 was almost ~90%, except for a few days surrounding menstruation showing 90%) metabolic capacity, however, on days around menstruation the FMO3 metabolic capacity was decreased to ~60–70%. Conclusion Together, these results indicate that abnormal FMO3 capacity is caused by menstruation particularly in the presence, in homozygous form, of mild genetic variants such as [Glu158Lys; Glu308Gly] that cause a reduced FMO3 function. PMID:17257434

  9. Prismatic Core Coupled Transient Benchmark

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

    2011-01-01

    The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

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

    International Nuclear Information System (INIS)

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

    1979-03-01

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

  11. Reactivity costs in MARIA reactor

    International Nuclear Information System (INIS)

    Marcinkowska, Zuzanna E.; Pytel, Krzysztof M.; Frydrysiak, Andrzej

    2017-01-01

    Highlights: • The methodology for calculating consumed fuel cost of excess reactivity is proposed. • Correlation between time integral of the core excess reactivity and released energy. • Reactivity price gives number of fuel elements required for given excess reactivity. - Abstract: For the reactor operation at high power level and carrying out experiments and irradiations the major cost of reactor operation is the expense of nuclear fuel. In this paper the methodology for calculating consumed fuel cost-relatedness of excess reactivity is proposed. Reactivity costs have been determined on the basis of operating data. A number of examples of calculating the reactivity costs for processes such as: strong absorbing material irradiation, molybdenium-99 production, beryllium matrix poisoning and increased moderator temperature illustrates proposed method.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  13. Reactivity insertion accident analysis

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

    1990-04-01

    The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

  14. The influence of reactive current on wind farm LVRT behavior

    Energy Technology Data Exchange (ETDEWEB)

    Li, Qing; Zhang, Mei; He, Jing; Qin, Shi-yao [China Electric Power Research Institute, Beijing (China)

    2012-07-01

    The Low voltage ride through (LVRT) capability of the whole wind farm is required in Chinese grid code published in 2011. In order to analyze the influence of reactive current on wind farm during grid fault, a 100 MW wind farm was simulated with the wind turbines which have been tested. Based on the validated wind turbine model, the wind farm was detailed modelled in DigSILENT/PowerFactory. The model of wind turbines, transformers, feeders, main transformers, static var compensator, and transmission lines was considered in the simulation. Under the weak and strong grid conditions, the wind farm was simulated with different wind turbine reactive current behavior during grid fault, respectively. The voltage distribution, active and reactive power transient behavior at the point of interconnection was analyzed. The results show that wind farm LVRT behavior is related to reactive current and LVRT capability of wind turbine, wind farm electrical structure and grid conditions. And it is very important for wind turbine to have a flexible dynamic reactive current control capability. (orig.)

  15. A COMETHE version with transient capability

    International Nuclear Information System (INIS)

    Vliet, J. van; Lebon, G.; Mathieu, P.

    1980-01-01

    A version of the COMETHE code is under development to simulate transient situations. This paper focuses on some aspects of the transient heat transfer models. Initially the coupling between transient heat transfer and other thermomechanical models is discussed. An estimation of the thermal characteristic times shows that the cladding temperatures are often in quasi-steady state. In order to reduce the computing time, calculations are therefore switched from a transient to a quasi-static numerical procedure as soon as such a quasi-equilibrium is detected. The temperature calculation is performed by use of the Lebon-Lambermont restricted variational principle, with piecewise polynoms as trial functions. The method has been checked by comparison with some exact results and yields good agreement for transient as well as for quasi-static situations. This method therefore provides a valuable tool for the simulation of the transient behaviour of nuclear reactor fuel rods. (orig.)

  16. Experience with transients in German NPPs

    International Nuclear Information System (INIS)

    Lindauer, E.

    1984-01-01

    This chapter examines reactor accidents in the Federal Republic of Germany based on the formal reporting system for licensee event reports (LERs) and a special investigation on all unplanned power variations in 3 PWRs. The significant transients experienced by BWR type reactors are analyzed. The main goal is to find weak points which caused the transient or influenced its course in an unfavorable way in order to improve the affected plant and others. The complete survey of all transients, with normally little or no safety relevance, allows statistical evaluations and the analysis of trends. It is concluded that significant transients were mainly experienced at older plants, whereas plants of an advanced design produced very few significant transients. The most frequent human errors which lead to transients are failure search in electronic systems and errors during design and commissioning

  17. Analysis of transient signals by Wavelet transform

    International Nuclear Information System (INIS)

    Penha, Rosani Libardi da; Silva, Aucyone A. da; Ting, Daniel K.S.; Oliveira Neto, Jose Messias de

    2000-01-01

    The objective of this work is to apply the Wavelet Transform in transient signals. The Wavelet technique can outline the short time events that are not easily detected using traditional techniques. In this work, the Wavelet Transform is compared with Fourier Transform, by using simulated data and rotor rig data. This data contain known transients. The wavelet could follow all the transients, what do not happen to the Fourier techniques. (author)

  18. Instrument response during overpower transients at TREAT

    International Nuclear Information System (INIS)

    Meek, C.C.; Bauer, T.H.; Hill, D.J.; Froehle, P.H.; Klickman, A.E.; Tylka, J.P.; Doerner, R.C.; Wright, A.E.

    1982-01-01

    A program to empirically analyze data residuals or noise to determine instrument response that occurs during in-pile transient tests is out-lined. As an example, thermocouple response in the Mark III loop during a severe overpower transient in TREAT is studied both in frequency space and in real-time. Time intervals studied included both constant power and burst portions of the power transient. Thermocouple time constants were computed. Benefits and limitations of the method are discussed

  19. Evaluation of reactivity and Xe behavior during daily load following operation

    International Nuclear Information System (INIS)

    Sakamoto, Yasunori; Araki, Tsuneyasu; Yamamoto, Fumiaki

    1992-01-01

    A boiling water reactor (BWR) has an excellent load following capability provided by a core flow control, which is used for changing a reactor power level and for compensating the subsequent Xe concentration change. The core characteristics during load following operations are investigated in detail, using our reactor core simulator. Comparisons of changes of the Doppler reactivity, the void reactivity and the Xe reactivity during transients are performed. Also the features of Xe transient during load following operations are shown. It has been shown that the core flow change required to compensate the Xe reactivity change produces much greater change of the void reactivity than that required for power level changes, and that the resulting local power change in the lower part of the core is greater than that in the upper part, because the Xe concentration change in the lower part is hardly compensated by the core flow control. Also the effects of power level changes, cycle patterns, and initial concentration of Xe and I on the Xe transient behavior have been investigated. (author)

  20. Study of transient rod extraction failure without RBM in a BWR

    International Nuclear Information System (INIS)

    Vallejo Q, J. A.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L.

    2015-09-01

    The study and analysis of the operational transients are important for predicting the behavior of a system to short-term events and the impact that would cause this transient. For the nuclear industry these studies are indispensable due to economic, environmental and social impacts that could cause an accident during the operation of a nuclear reactor. In this paper the preparation, simulation and analysis results of the transient rod extraction failure in which not taken into operation the RBM is presented. The study was conducted for a BWR of 2027 MWt, in an intermediate cycle of its useful life and using the computer code Simulate-3K a scenario of anomalies was created in the core reactivity which gave a coherent prediction to the type of presented event. (Author)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  2. Reactive documentation system

    Science.gov (United States)

    Boehnlein, Thomas R.; Kramb, Victoria

    2018-04-01

    Proper formal documentation of computer acquired NDE experimental data generated during research is critical to the longevity and usefulness of the data. Without documentation describing how and why the data was acquired, NDE research teams lose capability such as their ability to generate new information from previously collected data or provide adequate information so that their work can be replicated by others seeking to validate their research. Despite the critical nature of this issue, NDE data is still being generated in research labs without appropriate documentation. By generating documentation in series with data, equal priority is given to both activities during the research process. One way to achieve this is to use a reactive documentation system (RDS). RDS prompts an operator to document the data as it is generated rather than relying on the operator to decide when and what to document. This paper discusses how such a system can be implemented in a dynamic environment made up of in-house and third party NDE data acquisition systems without creating additional burden on the operator. The reactive documentation approach presented here is agnostic enough that the principles can be applied to any operator controlled, computer based, data acquisition system.

  3. OPTICAL TRANSIENT DETECTOR (OTD) LIGHTNING V1

    Data.gov (United States)

    National Aeronautics and Space Administration — The Optical Transient Detector (OTD) records optical measurements of global lightning events in the daytime and nighttime. The data includes individual point...

  4. Transient shielded liquid hydrogen containers

    International Nuclear Information System (INIS)

    Varghese, A.P.; Herring, R.H.

    1990-01-01

    The storage of hydrogen in the liquid phase has been limited in duration due to the thermal performance constraints of conventional Liquid Hydrogen containers available. Conventional Liquid Hydrogen containers lose hydrogen because of their relatively high heat leak and variations in usage pattern of hydrogen due to shutdowns. Local regulations also discourage venting of hydrogen. Long term storage of Liquid Hydrogen without product loss was usually accomplished using Liquid Nitrogen sacrificial shields. This paper reports on a new low heat leak container developed and patented that will extend the storage time of liquid hydrogen by five hundred percent. The principle of operation of the Transient Shields which makes the extraordinary performance of this container feasible is described in this paper. Also covered are the impact of this new container on present applications of hydrogen and the new opportunities afforded to Liquid hydrogen in the world hydrogen market

  5. Transient ischemic attack: diagnostic evaluation.

    Science.gov (United States)

    Messé, Steven R; Jauch, Edward C

    2008-08-01

    A transient ischemic attack portends significant risk of a stroke. Consequently, the diagnostic evaluation in the emergency department is focused on identifying high-risk causes so that preventive strategies can be implemented. The evaluation consists of a facilitated evaluation of the patient's metabolic, cardiac, and neurovascular systems. At a minimum, the following tests are recommended: fingerstick glucose level, electrolyte levels, CBC count, urinalysis, and coagulation studies; noncontrast computed tomography (CT) of the head; electrocardiography; and continuous telemetry monitoring. Vascular imaging studies, such as carotid ultrasonography, CT angiography, or magnetic resonance angiography, should be performed on an urgent basis and prioritized according to the patient's risk stratification for disease. Consideration should be given for echocardiography if no large vessel abnormality is identified.

  6. Measurement of fast transient pressures

    International Nuclear Information System (INIS)

    Procaccia, Henri

    1978-01-01

    The accuracy, reliability and sensitivity of a pressure transducers define its principal static characteristics. When the quantity measured varies with time, the measurement carries a dynamic error and a delay depending on the frequency of this variation. Hence, when fast pressure changes in a fluid have to be determined, different kinds of pressure transducers can be used depending on their inherent dynamic characteristics which must be compared with those of the transient phenomenon to be analysed. The text describes the pressure transducers generally employed in industry for analysing such phenomenon and gives two practical applications developed in the EDF: the first submits the measurements and results of pump cavitation tests carried out at the Vitry II EDF power station; the second deals with hammer blows particularly noticed in nuclear power stations and required the use of transducers of exceptionally high performance such as strain gauge transducers and piezoelectric transducers (response time within 1m sec.) [fr

  7. Predictive modeling of transient storage and nutrient uptake: Implications for stream restoration

    Science.gov (United States)

    O'Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.

    2010-01-01

    This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO3–) uptake rates inferred using the nutrient spiraling model underestimated the total NO3– mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO3– mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.

  8. Predictive Modeling of Transient Storage and Nutrient Uptake: Implications for Stream Restoration

    Energy Technology Data Exchange (ETDEWEB)

    O’Connor, Ben L.; Hondzo, Miki; Harvey, Judson W.

    2010-12-01

    This study examined two key aspects of reactive transport modeling for stream restoration purposes: the accuracy of the nutrient spiraling and transient storage models for quantifying reach-scale nutrient uptake, and the ability to quantify transport parameters using measurements and scaling techniques in order to improve upon traditional conservative tracer fitting methods. Nitrate (NO-3)(NO3-) uptake rates inferred using the nutrient spiraling model underestimated the total NO-3NO3- mass loss by 82%, which was attributed to the exclusion of dispersion and transient storage. The transient storage model was more accurate with respect to the NO-3NO3- mass loss (±20%) and also demonstrated that uptake in the main channel was more significant than in storage zones. Conservative tracer fitting was unable to produce transport parameter estimates for a riffle-pool transition of the study reach, while forward modeling of solute transport using measured/scaled transport parameters matched conservative tracer breakthrough curves for all reaches. Additionally, solute exchange between the main channel and embayment surface storage zones was quantified using first-order theory. These results demonstrate that it is vital to account for transient storage in quantifying nutrient uptake, and the continued development of measurement/scaling techniques is needed for reactive transport modeling of streams with complex hydraulic and geomorphic conditions.

  9. Burst-suppression is reactive to photic stimulation in comatose children with acquired brain injury

    DEFF Research Database (Denmark)

    Nita, Dragos A.; Moldovan, Mihai; Sharma, Roy

    2016-01-01

    reactivity. We quantified reactivity by measuring the change in the burst ratio (fraction of time in burst) following photic stimulation. Results: Photic stimulation evoked bursts in all patients, resulting in a transient increase in the burst ratio, while the mean heart rate remained unchanged......Objective: Burst-suppression is an electroencephalographic pattern observed during coma. In individuals without known brain pathologies undergoing deep general anesthesia, somatosensory stimulation transiently increases the occurrence of bursts. We investigated the reactivity of burst......-suppression in children with acquired brain injury. Methods: Intensive care unit electroencephalographic monitoring recordings containing burst-suppression were obtained from 5 comatose children with acquired brain injury of various etiologies. Intermittent photic stimulation was performed at 1 Hz for 1 min to assess...

  10. Transient photoconductivity in amorphous semiconductors

    International Nuclear Information System (INIS)

    Mpawenayo, P.

    1997-07-01

    Localized states in amorphous semiconductors are divided in disorder induced shallow trap levels and dangling bonds deep states. Dangling bonds are assumed here to be either neutral or charged and their energy distribution is a single gaussian. Here, it is shown analytically that transient photocurrent in amorphous semiconductors is fully controlled by charge carriers transitions between localized states for one part and tunneling hopping carriers on the other. Localized dangling bonds deep states act as non radiative recombination centres, while hopping tunnelling is assisted by the Coulomb interaction between defects sites. The half-width of defects distribution is the disorder parameter that determines the carrier hopping time between defects sites. The macroscopic time that explains the long decay response times observed will all types of amorphous semiconductors is duly thought to be temperature dependent. Basic equations developed by Longeaud and Kleider are solved for the general case of a semiconductor after photo-generation. It turns out that the transient photoconductivity decay has two components; one with short response times from carriers trap-release transitions between shallow levels and extended states and a hopping component made of inter-dependent exponentials whose time constants span in larger ranges depending on disorder. The photoconductivity hopping component appears as an additional term to be added to photocurrents derived from existing models. The results of the present study explain and complete the power law decay derived in the multiple trapping models developed 20 years ago only in the approximation of the short response time regime. The long response time regime is described by the hopping macroscopic time. The present model is verified for all samples of amorphous semiconductors known so far. Finally, it is proposed to improved the modulated photoconductivity calculation techniques by including the long-lasting hopping dark documents

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

    International Nuclear Information System (INIS)

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

    1999-02-01

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

  12. TRACY transient experiment databook. 3) Ramp feed experiment

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken; Yamane, Yuichi; Ogawa, Kazuhiko; Aizawa, Eiju; Yanagisawa, Hiroshi; Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-03-01

    This is a databook of TRACY ''ramp feed'' experiments. TRACY is a reactor to perform supercritical experiments using low-enriched uranyl nitrate aqueous solution. The excess reactivity of TRACY is 3$ at maximum, and it is inserted by feeding the solution to a core tank or by withdrawing a control rod, which is called as the transient rod, from the core. In the ramp feed experiment, the supercritical experiment is initiated by feeding the fuel solution to the core tank in a constant feed rate. The data in the present databook consist of datasheets and graphs. Experimental conditions and typical values of measured parameters are tabulated in the datasheet. In the graph, power and temperature profiles are plotted. Those data are useful for the investigation of criticality accidents with fissile solutions, and for validation of criticality accident analysis codes. (author)

  13. Analytical Solution to the Pneumatic Transient Rod System at ACRR

    International Nuclear Information System (INIS)

    Fehr, Brandon Michael

    2016-01-01

    The ACRR pulse is pneumatically driven by nitrogen in a system of pipes, valves and hoses up to the connection of the pneumatic system and mechanical linkages of the transient rod (TR). The main components of the TR pneumatic system are the regulator, accumulator, solenoid valve and piston-cylinder assembly. The purpose of this analysis is to analyze the flow of nitrogen through the TR pneumatic system in order to develop a motion profile of the piston during the pulse and be able to predict the pressure distributions inside both the cylinder and accumulators. The predicted pressure distributions will be validated against pressure transducer data, while the motion profile will be compared to proximity switch data. By predicting the motion of the piston, pulse timing will be determined and provided to the engineers/operators for verification. The motion profile will provide an acceleration distribution to be used in Razorback to more accurately predict reactivity insertion into the system.

  14. Analytical Solution to the Pneumatic Transient Rod System at ACRR

    Energy Technology Data Exchange (ETDEWEB)

    Fehr, Brandon Michael [Georgia Inst. of Technology, Atlanta, GA (United States)

    2016-01-08

    The ACRR pulse is pneumatically driven by nitrogen in a system of pipes, valves and hoses up to the connection of the pneumatic system and mechanical linkages of the transient rod (TR). The main components of the TR pneumatic system are the regulator, accumulator, solenoid valve and piston-cylinder assembly. The purpose of this analysis is to analyze the flow of nitrogen through the TR pneumatic system in order to develop a motion profile of the piston during the pulse and be able to predict the pressure distributions inside both the cylinder and accumulators. The predicted pressure distributions will be validated against pressure transducer data, while the motion profile will be compared to proximity switch data. By predicting the motion of the piston, pulse timing will be determined and provided to the engineers/operators for verification. The motion profile will provide an acceleration distribution to be used in Razorback to more accurately predict reactivity insertion into the system.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  16. Enhanced Severe Transient Analysis for Prevention Technical Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    This document outlines the development of a high fidelity, best estimate nuclear power plant severe transient simulation capability that will complement or enhance the integral system codes historically used for licensing and analysis of severe accidents. As with other tools in the Risk Informed Safety Margin Characterization (RISMC) Toolkit, the ultimate user of Enhanced Severe Transient Analysis and Prevention (ESTAP) capability is the plant decision-maker; the deliverable to that customer is a modern, simulation-based safety analysis capability, applicable to a much broader class of safety issues than is traditional Light Water Reactor (LWR) licensing analysis. Currently, the RISMC pathway’s major emphasis is placed on developing RELAP-7, a next-generation safety analysis code, and on showing how to use RELAP-7 to analyze margin from a modern point of view: that is, by characterizing margin in terms of the probabilistic spectra of the “loads” applied to systems, structures, and components (SSCs), and the “capacity” of those SSCs to resist those loads without failing. The first objective of the ESTAP task, and the focus of one task of this effort, is to augment RELAP-7 analyses with user-selected multi-dimensional, multi-phase models of specific plant components to simulate complex phenomena that may lead to, or exacerbate, severe transients and core damage. Such phenomena include: coolant crossflow between PWR assemblies during a severe reactivity transient, stratified single or two-phase coolant flow in primary coolant piping, inhomogeneous mixing of emergency coolant water or boric acid with hot primary coolant, and water hammer. These are well-documented phenomena associated with plant transients but that are generally not captured in system codes. They are, however, generally limited to specific components, structures, and operating conditions. The second ESTAP task is to similarly augment a severe (post-core damage) accident integral analyses code

  17. Reactive Power from Distributed Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kueck, John; Kirby, Brendan; Rizy, Tom; Li, Fangxing; Fall, Ndeye

    2006-12-15

    Distributed energy is an attractive option for solving reactive power and distribution system voltage problems because of its proximity to load. But the cost of retrofitting DE devices to absorb or produce reactive power needs to be reduced. There also needs to be a market mechanism in place for ISOs, RTOs, and transmission operators to procure reactive power from the customer side of the meter where DE usually resides. (author)

  18. Reactive Power from Distributed Energy

    International Nuclear Information System (INIS)

    Kueck, John; Kirby, Brendan; Rizy, Tom; Li, Fangxing; Fall, Ndeye

    2006-01-01

    Distributed energy is an attractive option for solving reactive power and distribution system voltage problems because of its proximity to load. But the cost of retrofitting DE devices to absorb or produce reactive power needs to be reduced. There also needs to be a market mechanism in place for ISOs, RTOs, and transmission operators to procure reactive power from the customer side of the meter where DE usually resides. (author)

  19. Reactive programming in eventsourcing systems

    OpenAIRE

    Kučinskas, Žilvinas

    2017-01-01

    Eventsourcing describes current state as series of events that occurred in a system. Events hold all information that is needed to recreate current state. This method allows to achieve high volume of transactions, and enables efficient replication. Whereas reactive programming lets implement reactive systems in declarative style, decomposing logic into smaller, easier to understand components. Thesis aims to create reactive programming program interface, incorporating both principles. Applyin...

  20. Reactive Programming in Standard ML

    OpenAIRE

    Pucella, Riccardo

    2004-01-01

    Reactive systems are systems that maintain an ongoing interaction with their environment, activated by receiving input events from the environment and producing output events in response. Modern programming languages designed to program such systems use a paradigm based on the notions of instants and activations. We describe a library for Standard ML that provides basic primitives for programming reactive systems. The library is a low-level system upon which more sophisticated reactive behavi...

  1. Dynamic remedial action scheme using online transient stability analysis

    Science.gov (United States)

    Shrestha, Arun

    Economic pressure and environmental factors have forced the modern power systems to operate closer to their stability limits. However, maintaining transient stability is a fundamental requirement for the operation of interconnected power systems. In North America, power systems are planned and operated to withstand the loss of any single or multiple elements without violating North American Electric Reliability Corporation (NERC) system performance criteria. For a contingency resulting in the loss of multiple elements (Category C), emergency transient stability controls may be necessary to stabilize the power system. Emergency control is designed to sense abnormal conditions and subsequently take pre-determined remedial actions to prevent instability. Commonly known as either Remedial Action Schemes (RAS) or as Special/System Protection Schemes (SPS), these emergency control approaches have been extensively adopted by utilities. RAS are designed to address specific problems, e.g. to increase power transfer, to provide reactive support, to address generator instability, to limit thermal overloads, etc. Possible remedial actions include generator tripping, load shedding, capacitor and reactor switching, static VAR control, etc. Among various RAS types, generation shedding is the most effective and widely used emergency control means for maintaining system stability. In this dissertation, an optimal power flow (OPF)-based generation-shedding RAS is proposed. This scheme uses online transient stability calculation and generator cost function to determine appropriate remedial actions. For transient stability calculation, SIngle Machine Equivalent (SIME) technique is used, which reduces the multimachine power system model to a One-Machine Infinite Bus (OMIB) equivalent and identifies critical machines. Unlike conventional RAS, which are designed using offline simulations, online stability calculations make the proposed RAS dynamic and adapting to any power system

  2. Clopidogrel discontinuation and platelet reactivity following coronary stenting

    LENUS (Irish Health Repository)

    2011-01-01

    Summary. Aims: Antiplatelet therapy with aspirin and clopidogrel is recommended for 1 year after drug-eluting stent (DES) implantation or myocardial infarction. However, the discontinuation of antiplatelet therapy has become an important issue as recent studies have suggested a clustering of ischemic events within 90 days of clopidogrel withdrawal. The objective of this investigation was to explore the hypothesis that there is a transient ‘rebound’ increase in platelet reactivity within 3 months of clopidogrel discontinuation. Methods and Results: In this prospective study, platelet function was assessed in patients taking aspirin and clopidogrel for at least 1 year following DES implantation. Platelet aggregation was measured using a modification of light transmission aggregometry in response to multiple concentrations of adenosine diphosphate (ADP), epinephrine, arachidonic acid, thrombin receptor activating peptide and collagen. Clopidogrel was stopped and platelet function was reassessed 1 week, 1 month and 3 months later. Thirty-two patients on dual antiplatelet therapy were recruited. Discontinuation of clopidogrel increased platelet aggregation to all agonists, except arachidonic acid. Platelet aggregation in response to ADP (2.5, 5, 10, and 20 μm) and epinephrine (5 and 20 μm) was significantly increased at 1 month compared with 3 months following clopidogrel withdrawal. Thus, a transient period of increased platelet reactivity to both ADP and epinephrine was observed 1 month after clopidogrel discontinuation. Conclusions: This study demonstrates a transient increase in platelet reactivity 1 month after clopidogrel withdrawal. This phenomenon may, in part, explain the known clustering of thrombotic events observed after clopidogrel discontinuation. This observation requires confirmation in larger populations.

  3. Massive florid reactive periostitis

    International Nuclear Information System (INIS)

    Nance, K.V.; Renner, J.B.; Brashear, H.R.; Siegal, G.P.; North Carolina Univ., Chapel Hill, NC

    1990-01-01

    Florid reactive periostitis is a rare, benign process usually occurring in the small, tubular bones of the hands and feet. Typically the lesion occurs in an adolescent or young adult and presents as a small area of pain and erythema over the affected bone. Although the histologic features may suggest malignancy, there is usually little radiographic evidence to support such a diagnosis. In the following report an unusual example of this entity is described whose large size and relentless local progression led to initial diagnostic uncertainty and eventual aggressive management. This case suggests that a wide spectrum of radiologic and morphologic changes may be seen in this entity and that a seemingly unrelated genetic disease may alter the typical clinical course. (orig.)

  4. Pembrolizumab reactivates pulmonary granulomatosis

    Directory of Open Access Journals (Sweden)

    Majdi Al-dliw

    2017-01-01

    Full Text Available Sarcoid like reaction is a well-known entity that occurs as a consequence to several malignancies or their therapies. Immunotherapy has gained a lot of interest in the past few years and has recently gained approval as first line therapy in multiple advanced stage malignancies. Pneumonitis has been described as complication of such therapy. Granulomatous inflammation has been only rarely reported subsequent to immunotherapy. We describe a case of granulomatous inflammation reactivation affecting the lungs in a patient previously exposed to Pembrolizumab and have evidence of a distant granulomatous infection. We discuss potential mechanisms of the inflammation and assert the importance of immunosuppression in controlling the dis-inhibited immune system.

  5. Reactive Oxygen Species

    DEFF Research Database (Denmark)

    Franchina, Davide G.; Dostert, Catherine; Brenner, Dirk

    2018-01-01

    T cells are a central component of defenses against pathogens and tumors. Their effector functions are sustained by specific metabolic changes that occur upon activation, and these have been the focus of renewed interest. Energy production inevitably generates unwanted products, namely reactive...... and transcription factors, influencing the outcome of the T cell response. We discuss here how ROS can directly fine-tune metabolism and effector functions of T cells....... oxygen species (ROS), which have long been known to trigger cell death. However, there is now evidence that ROS also act as intracellular signaling molecules both in steady-state and upon antigen recognition. The levels and localization of ROS contribute to the redox modeling of effector proteins...

  6. Control of ZrH reactor reactivity perturbations during orbital maneuvers

    International Nuclear Information System (INIS)

    Audette, R.F.

    1970-01-01

    Scheduled and inadvertent vehicle maneuvers in manned and unmanned space missions may result in reactivity perturbations to the ZrH reactor due to fuel and control drum motion from acceleration forces. Potential power and outlet coolant temperature excursions could result in interruptions of PCS power generation, or excessive coolant temperatures if uncontrolled. This analysis compares potential uncontrolled reactor transients with allowable transients for uninterrupted electrical power generation from a Brayton system, and presents a control scheme to limit transient reactor outlet temperatures to 1250 0 F for a system designed to operate at a nominal 1200 0 F reactor outlet. Potential uncontrolled transients could result in a reactor outlet temperature swing of +-77 0 F about a nominal 1200 0 F and a reactor power swing of +92 Kwt and -67 Kwt about a nominal 130 Kwt for the Brayton System. (U.S.)

  7. Weigle Reactivation in Acinetobacter Calcoaceticus

    DEFF Research Database (Denmark)

    Berenstein, Dvora

    1982-01-01

    phage and host survivals of about 5 times 10-6 and 1 times 10-1, respectively. Intracellular development of W-reactivated P78 was followed by one-step growth experiments. Conditions which allowed maximal W-reactivation also extended the period of phage production and yielded a somewhat reduced burst......Weigle (W)-reactivation was demonstrated in Acinetobacter calcoaceticus for the UV-irra-diated lysogenic phage P78. The reactivation factor (survival of irradiated phage on irradiated bacteria/ survival on unirradiated bacteria) reached a maximum value of 20. This was obtained at UV-doses giving...

  8. Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

    International Nuclear Information System (INIS)

    Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

    2013-01-01

    The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

  9. Generalization of Wilemski-Fixman-Weiss decoupling approximation to the case involving multiple sinks of different sizes, shapes, and reactivities.

    Science.gov (United States)

    Uhm, Jesik; Lee, Jinuk; Eun, Changsun; Lee, Sangyoub

    2006-08-07

    We generalize the Wilemski-Fixman-Weiss decoupling approximation to calculate the transient rate of absorption of point particles into multiple sinks of different sizes, shapes, and reactivities. As an application we consider the case involving two spherical sinks. We obtain a Laplace-transform expression for the transient rate that is in excellent agreement with computer simulations. The long-time steady-state rate has a relatively simple expression, which clearly shows the dependence on the diffusion constant of the particles and on the sizes and reactivities of sinks, and its numerical result is in good agreement with the known exact result that is given in terms of recursion relations.

  10. Modeling and analysis of thermal-hydraulic response of uranium-aluminum reactor fuel plates under transient heatup conditions

    Energy Technology Data Exchange (ETDEWEB)

    Navarro-Valenti, S.; Kim, S.H.; Georgevich, V. [Oak Ridge National Lab., TN (United States)] [and others

    1995-09-01

    The purpose of this paper is to describe the analysis performed to predict the thermal behavior of fuel miniplates under rapid transient heatup conditions. The possibility of explosive boiling was considered, and it was concluded that the heating rates are not large enough for explosive boiling to occur. However, transient boiling effects were pronounced. Because of the complexity of transient pool boiling and the unavailability of experimental data for the situations studied, an approximation was made that predicted the data very well within the uncertainties present. If pool boiling from the miniplates had been assumed to be steady during the heating pulse, the experimental data would have been greatly overestimated. This fact demonstrates the importance of considering the transient nature of heat transfer in the analysis of reactivity excursion accidents. An additional contribution of the present work is that it provided data on highly subcooled steady nulceate boiling from the cooling portion of the thermocouple traces.

  11. Plasma transport studies using transient techniques

    International Nuclear Information System (INIS)

    Simonen, T.C.; Brower, D.L.; Efthimion, P.

    1991-01-01

    Selected topics from the Transient Transport sessions of the Transport Task Force Workshop, held February 19-23, 1990, in Hilton Head, South Carolina are summarized. Presentations on sawtooth propagation, ECH modulation, particle modulation, and H-mode transitions are included. The research results presented indicated a growing theoretical understanding and experimental sophistication in the application of transient techniques to transport studies. (Author)

  12. The LOFAR Transients Key Science Project

    NARCIS (Netherlands)

    Stappers, B.; Fender, R.; Wijers, R.

    2009-01-01

    The Transients Key Science Project (TKP) is one of six Key Science Projects of the next generation radio telescope LOFAR. Its aim is the study of transient and variable low-frequency radio sources with an extremely broad science case ranging from relativistic jet sources to pulsars, exoplanets,

  13. Transient receptor potential channels in essential hypertension

    DEFF Research Database (Denmark)

    Liu, Daoyan; Scholze, Alexandra; Zhu, Zhiming

    2006-01-01

    The role of nonselective cation channels of the transient receptor potential channel (TRPC) family in essential hypertension has not yet been investigated.......The role of nonselective cation channels of the transient receptor potential channel (TRPC) family in essential hypertension has not yet been investigated....

  14. Physical modelling of a rapid boron dilution transient

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, N.G.; Hemstroem, B.; Karlsson, R. [Vattenfall Utveckling AB, Aelvkarleby (Sweden); Jacobson, S. [Vattenfall AB, Ringhals, Vaeroebacka (Sweden)

    1995-09-01

    The analysis of boron dilution accidents in pressurised water reactors has traditionally assumed that mixing is instantaneous and complete everywhere, eliminating in this way the possibility of concentration inhomogeneities. Situations can nevertheless arise where a volume of coolant with a low boron concentration may eventually enter the core and generate a severe reactivity transient. The work presented in this paper deals with a category of Rapid Boron Dilution Events characterised by a rapid start of a Reactor Coolant Pump (RCP) with a plug of relatively unborated water present in the RCS pipe. Model tests have been made at Vattenfall Utveckling AB in a simplified 1:5 scale model of a Westinghouse PWR. Conductivity measurements are used to determine dimensionless boron concentration. The main purpose of this experimental work is to define an experimental benchmark against which a mathematical model can be tested. The final goal is to be able to numerically predict Boron Dilution Transients. This work has been performed as a part of a Co-operative Agreement with Electricite` de France (EDF).

  15. 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)

  16. Nuclear data propagation with burnup. Impact on SFR reactivity coefficients

    International Nuclear Information System (INIS)

    Buiron, Laurent; Plisson-Rieunier, Daniele

    2017-01-01

    For the next generation fast reactor design, the Generation IV International Forum (GIF) defined global objectives in terms of safety improvement, sustainability, waste minimization and non-proliferation. Among the possibilities studied at CEA, Sodium cooled Fast Reactor (SFR) are studied as potential industrial tools for next decade's deployment. Many efforts have been made in the last years to obtain advanced industrial core designs that comply with these goals. Concerning safety issues, particular efforts have been made in order to obtain core designs that can be resilient to accidental transients. The 'safety' level of such new designs is often characterized by their 'natural' behavior under unprotected transients such as loss of flow or hypothetical transient over power. Transient analysis needs several accurate neutronic input data such as reactivity coefficient and kinetic parameters. Beside estimation of the level of 'absolute' values, associated uncertainties have also to be evaluated for the whole set of relevant data. These estimations have to be performed for different core state such as end of cycle core for feedback coefficient. This means that uncertainties have to be obtained not only a fixed time but also have to be propagated all through irradiation. To do so, we need to couple Boltzman and Bateman equations at sensitivities level. The coupling process could be done with the help of the perturbation theory which gives adapted framework suited for deterministic calculation codes. This coupling is currently in progress in ERANOS code system. The actual implementation gives access to estimation of sensitivities for both reactivity coefficients and mass balance. After a brief theoretical description of Boltzman/Bateman coupling capabilities in ERANOS, the study presented in this paper focuses on sensitivity and uncertainties estimation for the main feedback coefficients involved in fast reactor transients: the

  17. Nuclear Fuel Behaviour during Reactivity Initiated Accidents. Workshop Proceedings

    International Nuclear Information System (INIS)

    2010-01-01

    A reactivity initiated accident (RIA) is a nuclear reactor accident that involves an unwanted increase in fission rate and reactor power. The power increase may damage the reactor core. The main objective of the workshop was to review the current status of the experimental and analytical studies of the fuel behavior during the RIA transients in PWR and BWR reactors and the acceptance criteria for RIA in use and under consideration. The workshop was organized in an opening session and 5 technical sessions: 1) Recent experimental results and experimental techniques used; 2) Modelling and Data Interpretation; 3) Code Assessment; 4) RIA Core Analysis and 5) Revision and application of safety criteria

  18. Synchronizing noisy nonidentical oscillators by transient uncoupling

    Energy Technology Data Exchange (ETDEWEB)

    Tandon, Aditya, E-mail: adityat@iitk.ac.in; Mannattil, Manu, E-mail: mmanu@iitk.ac.in [Department of Physics, Indian Institute of Technology Kanpur, Kanpur, Uttar Pradesh 208016 (India); Schröder, Malte, E-mail: malte@nld.ds.mpg.de [Network Dynamics, Max Planck Institute for Dynamics and Self-Organization (MPIDS), 37077 Göttingen (Germany); Timme, Marc, E-mail: timme@nld.ds.mpg.de [Network Dynamics, Max Planck Institute for Dynamics and Self-Organization (MPIDS), 37077 Göttingen (Germany); Department of Physics, Technical University of Darmstadt, 64289 Darmstadt (Germany); Chakraborty, Sagar, E-mail: sagarc@iitk.ac.in [Department of Physics, Indian Institute of Technology Kanpur, Kanpur, Uttar Pradesh 208016 (India); Mechanics and Applied Mathematics Group, Indian Institute of Technology Kanpur, Kanpur, Uttar Pradesh 208016 (India)

    2016-09-15

    Synchronization is the process of achieving identical dynamics among coupled identical units. If the units are different from each other, their dynamics cannot become identical; yet, after transients, there may emerge a functional relationship between them—a phenomenon termed “generalized synchronization.” Here, we show that the concept of transient uncoupling, recently introduced for synchronizing identical units, also supports generalized synchronization among nonidentical chaotic units. Generalized synchronization can be achieved by transient uncoupling even when it is impossible by regular coupling. We furthermore demonstrate that transient uncoupling stabilizes synchronization in the presence of common noise. Transient uncoupling works best if the units stay uncoupled whenever the driven orbit visits regions that are locally diverging in its phase space. Thus, to select a favorable uncoupling region, we propose an intuitive method that measures the local divergence at the phase points of the driven unit's trajectory by linearizing the flow and subsequently suppresses the divergence by uncoupling.

  19. Transient survivability of LMR oxide fuel pins

    International Nuclear Information System (INIS)

    Weber, E.T.; Pitner, A.L.; Bard, F.E.; Culley, G.E.; Hunter, C.W.

    1986-01-01

    Fuel pin integrity during transient events must be assessed for both the core design and safety analysis phases of a reactor project. A significant increase in the experience related to limits of integrity for oxide fuel pins in transient overpower events has been realized from testing of fuel pins irradiated in FFTF and PFR. Fourteen FFTF irradiated fuel pins were tested in TREAT, representing a range of burnups, overpower ramp rates and maximum overpower conditions. Results of these tests along with similar testing in the PFR/TREAT program, provide a demonstration of significant safety margins for oxide fuel pins. Useful information applied in analytical extrapolation of fuel pin test data have been developed from laboratory transient tests on irradiated fuel cladding (FCTT) and on unirradiated fuel pellet deformation. These refinements in oxide fuel transient performance are being applied in assessment of transient capabilities of long lifetime fuel designs using ferritic cladding

  20. Characterizing transient noise in the LIGO detectors

    Science.gov (United States)

    Nuttall, L. K.

    2018-05-01

    Data from the LIGO detectors typically contain many non-Gaussian noise transients which arise due to instrumental and environmental conditions. These non-Gaussian transients can be an issue for the modelled and unmodelled transient gravitational-wave searches, as they can mask or mimic a true signal. Data quality can change quite rapidly, making it imperative to track and find new sources of transient noise so that data are minimally contaminated. Several examples of transient noise and the tools used to track them are presented. These instances serve to highlight the diverse range of noise sources present at the LIGO detectors during their second observing run. This article is part of a discussion meeting issue `The promises of gravitational-wave astronomy'.

  1. Nuclear reactors transients identification and classification system

    International Nuclear Information System (INIS)

    Bianchi, Paulo Henrique

    2008-01-01

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

  2. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  3. Reactivation of BK polyomavirus in patients with multiple sclerosis receiving natalizumab therapy.

    LENUS (Irish Health Repository)

    Lonergan, Roisin M

    2012-02-01

    Natalizumab therapy in multiple sclerosis has been associated with JC polyomavirus-induced progressive multifocal leucoencephalopathy. We hypothesized that natalizumab may also lead to reactivation of BK, a related human polyomavirus capable of causing morbidity in immunosuppressed groups. Patients with relapsing remitting multiple sclerosis treated with natalizumab were prospectively monitored for reactivation of BK virus in blood and urine samples, and for evidence of associated renal dysfunction. In this cohort, JC and BK DNA in blood and urine; cytomegalovirus (CMV) DNA in blood and urine; CD4 and CD8 T-lymphocyte counts and ratios in peripheral blood; and renal function were monitored at regular intervals. BK subtyping and noncoding control region sequencing was performed on samples demonstrating reactivation. Prior to commencement of natalizumab therapy, 3 of 36 patients with multiple sclerosis (8.3%) had BK viruria and BK reactivation occurred in 12 of 54 patients (22.2%). BK viruria was transient in 7, continuous in 2 patients, and persistent viruria was associated with transient viremia. Concomitant JC and CMV viral loads were undetectable. CD4:CD8 ratios fluctuated, but absolute CD4 counts did not fall below normal limits. In four of seven patients with BK virus reactivation, transient reductions in CD4 counts were observed at onset of BK viruria: these resolved in three of four patients on resuppression of BK replication. No renal dysfunction was observed in the cohort. BK virus reactivation can occur during natalizumab therapy; however, the significance in the absence of renal dysfunction is unclear. We propose regular monitoring for BK reactivation or at least for evidence of renal dysfunction in patients receiving natalizumab.

  4. Clinical applications of transient elastography

    Directory of Open Access Journals (Sweden)

    Kyu Sik Jung

    2012-06-01

    Full Text Available Chronic liver disease represents a major public health problem, accounting for significant morbidity and mortality worldwide. As prognosis and management depend mainly on the amount and progression of liver fibrosis, accurate quantification of liver fibrosis is essential for therapeutic decision-making and follow-up of chronic liver diseases. Even though liver biopsy is the gold standard for evaluation of liver fibrosis, non-invasive methods that could substitute for invasive procedures have been investigated during past decades. Transient elastography (TE, FibroScan® is a novel non-invasive method for assessment of liver fibrosis with chronic liver disease. TE can be performed in the outpatient clinic with immediate results and excellent reproducibility. Its diagnostic accuracy for assessment of liver fibrosis has been demonstrated in patients with chronic viral hepatitis; as a result, unnecessary liver biopsy could be avoided in some patients. Moreover, due to its excellent patient acceptance, TE could be used for monitoring disease progression or predicting development of liver-related complications. This review aims at discussing the usefulness of TE in clinical practice.

  5. Defect detection using transient thermography

    International Nuclear Information System (INIS)

    Mohd Zaki Umar; Ibrahim Ahmad; Ab Razak Hamzah; Wan Saffiey Wan Abdullah

    2008-08-01

    An experimental research had been carried out to study the potential of transient thermography in detecting sub-surface defect of non-metal material. In this research, eight pieces of bakelite material were used as samples. Each samples had a sub-surface defect in the circular shape with different diameters and depths. Experiment was conducted using one-sided Pulsed Thermal technique. Heating of samples were done using 30 kWatt adjustable quartz lamp while infra red (IR) images of samples were recorded using THV 550 IR camera. These IR images were then analysed with ThermofitTMPro software to obtain the Maximum Absolute Differential Temperature Signal value, ΔΤ m ax and the time of its appearance, τ m ax (ΔΤ). Result showed that all defects were able to be detected even for the smallest and deepest defect (diameter = 5 mm and depth = 4 mm). However the highest value of Differential Temperature Signal (ΔΤ m ax), were obtained at defect with the largest diameter, 20 mm and at the shallowest depth, 1 mm. As a conclusion, the sensitivity of the pulsed thermography technique to detect sub-surface defects of bakelite material is proportionately related with the size of defect diameter if the defects are at the same depth. On the contrary, the sensitivity of the pulsed thermography technique inversely related with the depth of defect if the defects have similar diameter size. (Author)

  6. Transient shocks beyond the heliopause

    International Nuclear Information System (INIS)

    Fermo, R L; Pogorelov, N V; Burlaga, L F

    2015-01-01

    The heliopause is a rich, dynamic surface affected by the time-dependent solar wind. Stream interactions due to coronal mass ejections (CMEs), corotating interaction regions (CIRs), and other transient phenomena are known to merge producing global merged interaction regions (GMIRs). Numerical simulations of the solar wind interaction with the local interstellar medium (LISM) show that GMIRs, as well other time-dependent structures in the solar wind, may produce compression/rarefaction waves and shocks in the LISM behind the heliopause. These shocks may initiate wave activity observed by the Voyager spacecraft. The magnetometer onboard Voyager 1 indeed observed a few structures that may be interpreted as shocks. We present numerical simulations of such shocks in the year of 2000, when both Voyager spacecraft were in the supersonic solar wind region, and in 2012, when Voyager 1 observed traveling shocks. In the former case, Voyager observations themselves provide time- dependent boundary conditions in the solar wind. In the latter case, we use OMNI data at 1 AU to analyze the plasma and magnetic field behavior after Voyager 1 crossed the heliospheric boundary. Numerical results are compared with spacecraft observations. (paper)

  7. Transient phenomena in multiphase flow

    International Nuclear Information System (INIS)

    Afgan, N.H.

    1988-01-01

    This book is devoted to formulation of the two-phase system. Emphasis is given to classical instantaneous equations of mass momentum and energy for local conditions and respective averaging procedures and their relevance to the structure of transfer laws. In formulating an equation for a two-velocity continuum, two-phase dispersed flow, two-velocity and local inertial effects associated with contraction and expansion of the mixture have been considered. Particular attention is paid to the effects of interface topology and area concentration as well as the latter's dependence on interfacial transfer laws. Also covered are low bubble concentrations in basic nonuniform unsteady flow where interactions between bubbles are negligible but where the effects of bubbles must still be considered. Special emphasis has been given to the pairwise interaction of the bubble and respective hydrodynamic equations describing the motion of a pair of spherical bubbles through a liquid This book introduces turbulence phenomena in two-phase flow and related problems of phase distribution in two-phase flow. This includes an extensive survey of turbulence and phase distribution models in transient two-phase flow. It is shown that if the turbulent structure of the continuous phase of bubbly two-phase is either measured or can be predicted, then the observed lateral phase distribution can be determined by using an multidimensional two-fluid model in which all lateral forces are properly modeled

  8. Regarding KUR Reactivity Measurement System

    International Nuclear Information System (INIS)

    Nakamori, Akira; Hasegawa, Kei; Tsuchiyama, Tatsuo; Yamamoto, Toshihiro; Okumura, Ryo; Sano, Tadafumi

    2012-01-01

    This article reported: (1) the outline of the reactivity measurement system of Kyoto University Research Reactor (KUR), (2) the calibration data of control rod, (3) the problems and the countermeasures for range switching of linear output meter. For the laptop PC for the reactivity measurement system, there are four input signals: (1) linear output meter, (2) logarithmic output meter, (3) core temperature gauge, and (4) control rod position. The hardware of reactivity measurement system is controlled with Labview installed on the laptop. Output, reactivity, reactor period, and the change in reactivity due to temperature effect or Xenon effect are internally calculated and displayed in real-time with Labview based on the four signals above. Calculation results are recorded in the form of a spreadsheet. At KUR, the reactor core arrangement was changed, so the control rod was re-calibrated. At this time, calculated and experimental values of reactivity based on the reactivity measurement system were compared, and it was confirmed that the reactivity calculation by Labview was accurate. The range switching of linear output meter in the nuclear instrumentation should automatically change within the laptop, however sometimes this did not function properly in the early stage. It was speculated that undefined percent values during the transition of percent value were included in the calculation and caused calculation errors. The range switching started working properly after fixing this issue. (S.K.)

  9. Reactive agents and perceptual ambiguity

    NARCIS (Netherlands)

    Dartel, M. van; Sprinkhuizen-Kuyper, I.G.; Postma, E.O.; Herik, H.J. van den

    2005-01-01

    Reactive agents are generally believed to be incapable of coping with perceptual ambiguity (i.e., identical sensory states that require different responses). However, a recent finding suggests that reactive agents can cope with perceptual ambiguity in a simple model (Nolfi, 2002). This paper

  10. A Tariff for Reactive Power

    Energy Technology Data Exchange (ETDEWEB)

    Kueck, John D [ORNL; Kirby, Brendan J [ORNL; Li, Fangxing [ORNL; Tufon, Christopher [Pacific Gas and Electric Company; Isemonger, Alan [California Independent System Operator

    2008-07-01

    Two kinds of power are required to operate an electric power system: real power, measured in watts, and reactive power, measured in volt-amperes reactive or VARs. Reactive power supply is one of a class of power system reliability services collectively known as ancillary services, and is essential for the reliable operation of the bulk power system. Reactive power flows when current leads or lags behind voltage. Typically, the current in a distribution system lags behind voltage because of inductive loads such as motors. Reactive power flow wastes energy and capacity and causes voltage droop. To correct lagging power flow, leading reactive power (current leading voltage) is supplied to bring the current into phase with voltage. When the current is in phase with voltage, there is a reduction in system losses, an increase in system capacity, and a rise in voltage. Reactive power can be supplied from either static or dynamic VAR sources. Static sources are typically transmission and distribution equipment, such as capacitors at substations, and their cost has historically been included in the revenue requirement of the transmission operator (TO), and recovered through cost-of-service rates. By contrast, dynamic sources are typically generators capable of producing variable levels of reactive power by automatically controlling the generator to regulate voltage. Transmission system devices such as synchronous condensers can also provide dynamic reactive power. A class of solid state devices (called flexible AC transmission system devices or FACTs) can provide dynamic reactive power. One specific device has the unfortunate name of static VAR compensator (SVC), where 'static' refers to the solid state nature of the device (it does not include rotating equipment) and not to the production of static reactive power. Dynamic sources at the distribution level, while more costly would be very useful in helping to regulate local voltage. Local voltage regulation would

  11. PROCEEDINGS: MULTIPOLLUTANT SORBENT REACTIVITY ...

    Science.gov (United States)

    The report is a compilation of technical papers and visual aids presented by representatives of industry, academia, and government agencies at a workshop on multipollutant sorbent reactivity that was held at EPA's Environmental Research Center in Research Triangle Park, NC, on July 19-20, 1994. There were 16 technical presentations in three sessions, and a panel discussion between six research experts. The workshop was a forum for the exchange of ideas and information on the use of sorbents to control air emissions of acid gases (sulfur dioxide, nitrogen oxides, and hydrogen chloride); mercury and dioxins; and toxic metals, primarily from fossil fuel combustion. A secondary purpose for conducting the workshop was to help guide EPA's research planning activities. A general theme of the workshop was that a strategy of controlling many pollutants with a single system rather than systems to control individual pollutants should be a research goal. Some research needs cited were: hazardous air pollutant removal by flue gas desulfurization systems, dioxin formation and control, mercury control, waste minimization, impact of ash recycling on metals partitioning, impact of urea and sorbents on other pollutants, high temperature filtration, impact of coal cleaning on metals partitioning, and modeling dispersion of sorbents in flue gas. information

  12. Reactivation with productivity

    International Nuclear Information System (INIS)

    Garcia, Carlos Hernando

    2002-01-01

    A market to five years that it will move near $63.000 millions, starting from the production of 254.000 reserves that Ecopetrol requires for its maintenance and operation, it was projected with base in the offer study and it demands that they carried out the universities Javeriana and Industrial of Santander for the Colombian Company of Petroleum around the metal mechanic sector. In accordance with the figures of the report, Ecopetrol, like one of the state entities selected by the national government to design pilot programs, guided to reactivate the Colombian industry; it is projecting a good perspective for the Colombian economy and the invigoration of the national productive sector. In practical terms, the report points out that Ecopetrol, in its different operative centers, will require in next five years the quantity of had restored before mentioned in the lines of mechanical stamps, centrifugal bombs, inter chambers of heat, compressors and valves of security; pieces that are elaborated by international makers in 99%. To produce them nationally would represent to the company an economy of 52% of the total value of the purchases in next five years and a reduction of time of delivery of 17 weeks to one week

  13. Fission gas behavior during fast thermal transients

    International Nuclear Information System (INIS)

    Esteves, R.G.

    1976-01-01

    The behavior of non-equilibrium fission in fuel elements undergoing fast thermal transients is analyzed. To facilitate the analysis, a new variable, the equilibrium variable (EV) is defined. This variable, together with bubble radius, completely specifies a bubble with respect to its size and equilibrium condition. The analysis is coded using a two-variable (radius and EV) multigroup numerical approximation that accepts as input the time-temperature history, the time-fission rate history, and the time-thermal gradient history of the fuel element. Studies were performed to test the code for convergence with respect to the time interval and the number of groups chosen. For a series of transient simulation studies, the measurements obtained at HEDL (microscopic examination of intragranular porosity in oxide fuel transient-tested in TREAT) are used. Two different transient histories were selected; the first, a high-temperature transient (HTT) with a peak at 2477 0 K and the second, a low-temperature transient (LTT) with a peak-temperature at 2000 0 K. The LTT was simulated for three different conditions: Bubbles were allowed to move via (a) only biased migration, (b) via random migration, and (c) via both mechanisms. The HTT was also run for both mechanisms. The agreement with HEDL microscopic observations was fair for bubbles smaller than 964 A in diameter, and poor for larger bubbles. Bubbles that grew during the heat-up part of the transient were frozen at a larger size during the cool down

  14. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

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

  15. Assessment of transient stability of cable based transmission grids with reactive power compensation

    DEFF Research Database (Denmark)

    Foo, Yii; Dall, Laurids; Silva, Filipe Miguel Faria da

    2017-01-01

    Underground transmission cables are gaining popularity due to its applications near cities and aesthetic purpose. For example in Denmark, the transmission power grid is changing significantly as many conventional overhead lines (OHL) are replaced by cables and more is expected over the coming years...... through a series of sensitivity analysis with respect to the cables compensation degree. A separate case of disconnecting the SRs of the faulted line is also carried out. The tendencies are initially observed and explained for smaller systems, Single-Machine Infinite Bus (SMIB) and 9-bus system...

  16. TRIGA control rod position and reactivity transient Monitoring by Neural Networks

    International Nuclear Information System (INIS)

    Rosa, R.; Palomba, M.; Sepielli, M.

    2008-01-01

    Plant sensors drift or malfunction and operator actions in nuclear reactor control can be supported by sensor on-line monitoring, and data validation through soft-computing process. On-line recalibration can often avoid manual calibration or drifting component replacement. DSP requires prompt response to the modified conditions. Artificial Neural Network (ANN) and Fuzzy logic ensure: prompt response, link with field measurement and physical system behaviour, data incoming interpretation, and detection of discrepancy for mis-calibration or sensor faults. ANN (Artificial Neural Network) is a system based on the operation of biological neural networks. Although computing is day by day advancing, there are certain tasks that a program made for a common microprocessor is unable to perform. A software implementation of an ANN can be made with Pros and Cons. Pros: A neural network can perform tasks that a linear program can not; When an element of the neural network fails, it can continue without any problem by their parallel nature; A neural network learns and does not need to be reprogrammed; It can be implemented in any application; It can be implemented without any problem. Cons: The architecture of a neural network is different from the architecture of microprocessors therefore needs to be emulated; it requires high processing time for large neural networks; and the neural network needs training to operate. Three possibilities of training exist: Supervised learning: the network is trained providing input and matching output patterns; Unsupervised learning: input patterns are not a priori classified and the system must develop its own representation of the input stimuli; Reinforcement Learning: intermediate form of the above two types of learning, the learning machine does some action on the environment and gets a feedback response from the environment. Two TRIGAN ANN applications are considered: control rod position and fuel temperature. The outcome obtained in this two applications have been satisfactory as the error in steady state resulted less than the expected one and the training method quite effective. Testing will continue with increasing of data scanning rate to improve the answer during status transitions and investigate how to decrease oscillations during the steady-states

  17. Reassessment of the basis for NRC fuel damage criteria for reactivity transients

    International Nuclear Information System (INIS)

    McCardell, R.K.

    1994-01-01

    The present basis for NRC Fuel Damage Criteria was obtained from experiments performed in the Special Power Excursion Reactor Test (SPERT) IV Reactor Capsule Driver Core (CDC) at the Idaho National Engineering Laboratory (INEL) between 1967 and 1970. Most of the CDC test fuel rods were previously unirradiated and the failure threshold for these unirradiated fuel rods was measured to be about 200 calories per gram of UO 2 radially averaged fuel enthalpy at the axial peak

  18. Characterizing SI Engine Transient Fuel Consumption in ALPHA

    Science.gov (United States)

    Examine typical transient engine operation encountered over the EPA's vehicle and engine testing drive cycles to characterize that transient fuel usage, and then describe the changes made to ALPHA to better model transient engine operation.

  19. New developments in French transient monitoring: SYSFAC

    International Nuclear Information System (INIS)

    L'huby, Y.; Genette, P.; Faidy, C.; Kappler, F.; Balley, J.; Bimont, G.

    1991-01-01

    After more than ten years of experience with Transient Monitoring and Logging Procedure (TMLP) and six years of successfully experience with Fatiguemeters, EDF has decided to study a new concept of Fatigue Monitoring System: SYSFAC. This new automatic system which is developed to be operating in all the French PWR units is composed of three modules: mechanical transient logging, functional transient logging and fatiguemeters. This application must be connected to the on-site data acquisition system without complementary instrumentation on the plant. (author)

  20. Transient phenomena in electrical power systems

    CERN Document Server

    Venikov, V A; Higinbotham, W

    1964-01-01

    Electronics and Instrumentation, Volume 24: Transient Phenomena in Electrical Power Systems presents the methods for calculating the stability and the transient behavior of systems with forced excitation control. This book provides information pertinent to the analysis of transient phenomena in electro-mechanical systems.Organized into five chapters, this volume begins with an overview of the principal requirements in an excitation system. This text then explains the electromagnetic and electro-mechanical phenomena, taking into account the mutual action between the components of the system. Ot

  1. Transient analysis capabilities at ABB-CE

    International Nuclear Information System (INIS)

    Kling, C.L.

    1992-01-01

    The transient capabilities at ABB-Combustion Engineering (ABB-CE) Nuclear Power are a function of the computer hardware and related network used, the computer software that has evolved over the years, and the commercial technical exchange agreements with other related organizations and customers. ABB-CEA is changing from a mainframe/personal computer network to a distributed workstation/personal computer local area network. The paper discusses computer hardware, mainframe computing, personal computers, mainframe/personal computer networks, workstations, transient analysis computer software, design/operation transient analysis codes, safety (licensed) analysis codes, cooperation with ABB-Atom, and customer support

  2. Transient burnout in flow reduction condition

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Kuroyanagi, Toshiyuki

    1981-01-01

    A transient flow reduction burnout experiment was conducted with water in a uniformly heated, vertically oriented tube. Test pressures ranged from 0.5 to 3.9 MPa. An analytical method was developed to obtain transient burnout conditions at the exit. A simple correlation to predict the deviation of the transient burnout mass velocity at the tube exit from the steady state mass velocity obtained as a function of steam-water density ratio and flow reduction rate. The correlation was also compared with the other data. (author)

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  4. Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

    International Nuclear Information System (INIS)

    Mueller, C.J.

    1986-01-01

    The concept of inherent shutdown is emphasized in the approach to the design of innovative, small pool-type liquid-metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower events in evolving metal and oxide innovative designs

  5. Impacts of reactivity feedback uncertainties on inherent shutdown in innovative designs

    International Nuclear Information System (INIS)

    Mueller, C.J.

    1986-01-01

    The concept of ''inherent shutdown'' is emphasized in the approach to the design of innovative, small pool-type liquid metal reactors (LMRs). This paper reports an evaluation of reactivity feedback uncertainties used in the analyses of anticipated transients without scram (ATWS) for innovative LMRs, and the associated impacts on safety margins and inherent shutdown success probabilities on unprotected loss-of-flow (LOF) events. It then assesses the ultimate importance of these uncertainties on LOF and transient overpower (TOP) events in evolving metal and oxide innovative designs

  6. Present art of reactivity determination

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko; Nakano, Masafumi; Matsuura, Shojiro

    1977-01-01

    Experimental techniques for reactivity determination of a reactor have been one of the long standing subjects in reactor physics. Recently, such a requirement was proposed by the reactor designers and operators that the values of reactivity should be measured more accurately. This is because importance is emphasized for the role of reactivity to the performance of reactor safety, economics and operability. Motivated by the requirement, some remarkable progresses are being made for the improvement of the experimental techniques. Then, the present review summarizes the research activities on this subject, identifies several reactor physics problems to be overcome, and makes mention of the future targets. (auth.)

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

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cliff Bybee

    2003-09-01

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

  8. simulation of electromagnetic transients in power systems

    African Journals Online (AJOL)

    Dr Obe

    1996-09-01

    Sep 1, 1996 ... Transients in power systems are initiated by abrupt changes to otherwise steady operating conditions. These changes would .... The method is applicable both to single transmission in real time. The method is applicable both ...

  9. The economic impact of reactor transients

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  10. Fuel cladding mechanical properties for transient analysis

    International Nuclear Information System (INIS)

    Johnson, G.D.; Hunter, C.W.; Hanson, J.E.

    1976-01-01

    Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s (5.6 0 K/s) and 200 0 F/s (111 0 K/s) to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

  11. Detection of Transient Signals in Doppler Spectra

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Signal processing is used to detect transient signals in the presence of noise. Two embodiments are disclosed. In both embodiments, the time series from a remote...

  12. Transient management using the safety function approach

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Barrow, J.H.; Bischoff, G.C.; Callaghan, V.M.; Pearce, R.T.

    1984-01-01

    The safety function approach is described. Its use in the development of a transient management procedures system includes optimal recovery procedures tailored to specific, anticipated symptom sets and a functional recovery procedure which is more general. Simulator evaluations are described

  13. Development of a transient criticality evaluation method

    International Nuclear Information System (INIS)

    Pain, C.C.; Eaton, M.D.; Miles, B.; Ziver, A.K.; Gomes, J.L.M.A.; Umpleby, A.P.; Piggott, M.D.; Goddard, A.J.H.; Oliveira, C.R.E. de

    2005-01-01

    In developing a transient criticality evaluation method we model, in full spatial/temporal detail, the neutron fluxes and consequent power and the evolving material properties - their flows, energies, phase changes etc. These methods are embodied in the generic method FETCH code which is based as far as possible on basic principles and is capable of use in exploring safety-related situations somewhat beyond the range of experiment. FETCH is a general geometry code capable of addressing a range of criticality issues in fissile materials. The code embodies both transient radiation transport and transient fluid dynamics. Work on powders, granular materials, porous media and solutions is reviewed. The capability for modelling transient criticality for chemical plant, waste matrices and advanced reactors is also outlined. (author)

  14. RFSP simulations of Darlington FINCH refuelling transient

    International Nuclear Information System (INIS)

    Carruthers, E.V.; Chow, H.C.

    1997-01-01

    Immediately after refuelling of a channel, the fresh bundles are free of fission products. Xenon-135, the most notable of the saturating fission products, builds up to an equilibrium level in about 30 h. The channel power of the refuelled channel would therefore initially peak and then drop to a steady-state level. The RFSP code can track saturating-fission-product transients and power transients. The Fully INstrumented CHannels (FINCHs) in Darlington NGS provides channel power data on the refuelling power transients. In this paper, such data has been used to identify the physical evidence of the fission-product transient effect on channel power, and to validate RFSP fission-product-driver calculation results. (author)

  15. Climatic feedbacks between stationary and transient eddies

    International Nuclear Information System (INIS)

    Branscome, L.E.

    1994-01-01

    Stationary eddies make a significant contribution to poleward heat transport during Northern Hemisphere winter, equaling the transport by transient eddies. On the other hand, stationary eddy transport during the summer is negligible. The effect of topography on time-mean stationary waves and low-frequency variability has been widely studied. In contrast, little attention has been given to the climatic feedbacks associated with stationary eddies. Furthermore, the relationship between stationary and transient eddies in the context of global and regional climate is not well understood. The response of the climate system to anthropogenic forcing is likely to have some dependence on stationary wave transport and its interaction with transient eddies. Some early GCM simulations and observational analyses indicate a strong feedback between the meridional heat fluxes of stationary and transient eddies

  16. Acoustic Transient Source Localization From an Aerostat

    National Research Council Canada - National Science Library

    Scanlon, Michael; Reiff, Christian; Noble, John

    2006-01-01

    The Army Research Laboratory (ARL) has conducted experiments using acoustic sensor arrays suspended below tethered aerostats to detect and localize transient signals from mortars, artillery and small arms fire...

  17. Attosecond transient absorption spectroscopy of molecular hydrogen

    International Nuclear Information System (INIS)

    Martín, Fernando; González-Castrillo, Alberto; Palacios, Alicia; Argenti, Luca; Cheng, Yan; Chini, Michael; Wang, Xiaowei; Chang, Zenghu

    2015-01-01

    We extend attosecond transient absorption spectroscopy (ATAS) to the study of hydrogen molecules, demonstrating the potential of the technique to resolve – simultaneously and with state resolution – both the electronic and nuclear dynamics. (paper)

  18. Processing of polymers using reactive solvents

    NARCIS (Netherlands)

    Lemstra, P.J.; Kurja, J.; Meijer, H.E.H.; Meijer, H.E.H.

    1997-01-01

    A review with many refs. on processing of polymers using reactive solvents including classification of synthetic polymers, guidelines for the selection of reactive solvents, basic aspects of processing, examples of intractable and tractable polymer/reactive solvent system

  19. Sweden: Autonomous Reactivity Control (ARC) Systems

    International Nuclear Information System (INIS)

    Qvist, Staffan A.

    2015-01-01

    The next generation of nuclear energy systems must be licensed, constructed, and operated in a manner that will provide a competitively priced supply of energy, keeping in consideration an optimum use of natural resources, while addressing nuclear safety, waste, and proliferation resistance, and the public perception concerns of the countries in which those systems are deployed. These issues are tightly interconnected, and the implementation of passive and inherent safety features is a high priority in all modern reactor designs since it helps to tackle many of the issues at once. To this end, the Autonomous Reactivity Control (ARC) system was developed to ensure excellent inherent safety performance of Generation-IV reactors while having a minimal impact on core performance and economic viability. This paper covers the principles for ARC system design and analysis, the problem of ensuring ARC system response stability and gives examples of the impact of installing ARC systems in well-known fast reactor core systems. It is shown that even with a relatively modest ARC installation, having a near-negligible impact on core performance during standard operation, cores such as the European Sodium Fast Reactor (ESFR) can be made to survive any postulated unprotected transient without coolant boiling or fuel melting

  20. Cable system transients theory, modeling and simulation

    CERN Document Server

    Ametani, Akihiro; Nagaoka, Naoto

    2015-01-01

    A systematic and comprehensive introduction to electromagnetic transient in cable systems, written by the internationally renowned pioneer in this field Presents a systematic and comprehensive introduction to electromagnetic transient in cable systems Written by the internationally renowned pioneer in the field Thorough coverage of the state of the art on the topic, presented in a well-organized, logical style, from fundamentals and practical applications A companion website is available

  1. Partial discharge transients: The field theoretical approach

    DEFF Research Database (Denmark)

    McAllister, Iain Wilson; Crichton, George C

    1998-01-01

    Up until the mid-1980s the theory of partial discharge transients was essentially static. This situation had arisen because of the fixation with the concept of void capacitance and the use of circuit theory to address what is in essence a field problem. Pedersen rejected this approach and instead...... began to apply field theory to the problem of partial discharge transients. In the present paper, the contributions of Pedersen using the field theoretical approach will be reviewed and discussed....

  2. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

    Castano, J.; Schivo, M.

    1987-01-01

    A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

  3. PC-Reactor-core transient simulation code

    International Nuclear Information System (INIS)

    Nakata, H.

    1989-10-01

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

  4. Modeling of Transients in an Enrichment Circuit

    International Nuclear Information System (INIS)

    Fernandino, Maria; Delmastro, Dario; Brasnarof, Daniel

    2003-01-01

    In the present work a mathematical model is presented in order to describe the dynamic behavior inside a closed enrichment loop, the latter representing a single stage of an uranium gaseous diffusion enrichment cascade.The analytical model is turned into a numerical model, and implemented through a computational code.Transients of two species separation were numerically analyzed, including setting times of each magnitude, behavior of each one of them during different transients, and redistribution of concentrations along the closed loop

  5. Transient Exciplex Formation Electron Transfer Mechanism

    OpenAIRE

    Michael G. Kuzmin; Irina V. Soboleva; Elena V. Dolotova

    2011-01-01

    Transient exciplex formation mechanism of excited-state electron transfer reactions is analyzed in terms of experimental data on thermodynamics and kinetics of exciplex formation and decay. Experimental profiles of free energy, enthalpy, and entropy for transient exciplex formation and decay are considered for several electron transfer reactions in various solvents. Strong electronic coupling in contact pairs of reactants causes substantial decrease of activation energy relative to that for c...

  6. Transient ischemic attack presenting in an elderly patient with transient ophthalmic manifestations

    Directory of Open Access Journals (Sweden)

    Sparshi Jain

    2016-01-01

    Full Text Available Transient ischemic attack (TIA is a transient neurological deficit of cerebrovascular origin without infarction which may last only for a short period and can have varying presentations. We report a case of 58-year-old male with presenting features of sudden onset transient vertical diplopia and transient rotatory nystagmus which self-resolved within 12 h. Patient had no history of any systemic illness. On investigating, hematological investigations and neuroimaging could not explain these sudden and transient findings. A TIA could possibly explain these sudden and transient ocular findings in our patient. This case report aims to highlight the importance of TIA for ophthalmologists. We must not ignore these findings as these could be warning signs of an impending stroke which may or may not be detected on neuroimaging. Thus, early recognition, primary prevention strategies, and timely intervention are needed.

  7. Establishment of HSV1 latency in immunodeficient mice facilitates efficient in vivo reactivation.

    Directory of Open Access Journals (Sweden)

    Chandran Ramakrishna

    2015-03-01

    Full Text Available The establishment of latent infections in sensory neurons is a remarkably effective immune evasion strategy that accounts for the widespread dissemination of life long Herpes Simplex Virus type 1 (HSV1 infections in humans. Periodic reactivation of latent virus results in asymptomatic shedding and transmission of HSV1 or recurrent disease that is usually mild but can be severe. An in-depth understanding of the mechanisms regulating the maintenance of latency and reactivation are essential for developing new approaches to block reactivation. However, the lack of a reliable mouse model that supports efficient in vivo reactivation (IVR resulting in production of infectious HSV1 and/or disease has hampered progress. Since HSV1 reactivation is enhanced in immunosuppressed hosts, we exploited the antiviral and immunomodulatory activities of IVIG (intravenous immunoglobulins to promote survival of latently infected immunodeficient Rag mice. Latently infected Rag mice derived by high dose (HD, but not low dose (LD, HSV1 inoculation exhibited spontaneous reactivation. Following hyperthermia stress (HS, the majority of HD inoculated mice developed HSV1 encephalitis (HSE rapidly and synchronously, whereas for LD inoculated mice reactivated HSV1 persisted only transiently in trigeminal ganglia (Tg. T cells, but not B cells, were required to suppress spontaneous reactivation in HD inoculated latently infected mice. Transfer of HSV1 memory but not OVA specific or naïve T cells prior to HS blocked IVR, revealing the utility of this powerful Rag latency model for studying immune mechanisms involved in control of reactivation. Crossing Rag mice to various knockout strains and infecting them with wild type or mutant HSV1 strains is expected to provide novel insights into the role of specific cellular and viral genes in reactivation, thereby facilitating identification of new targets with the potential to block reactivation.

  8. Flow transients experiments with refrigerant-12

    International Nuclear Information System (INIS)

    Celata, G.P.; D'Annibale, F.; Farello, G.E.; Setaro, T.

    1986-01-01

    Flow transients have been investigated in a wide range of thermal-hydraulics situations with Refrigerannt-12. Six pressures (including the reference to PWR and BWR characteristic liquid to vapour densities ratios), several periods of the flowrate transients coastdown during the simulated flow decays, and different specific mass flowrate have been studied emploiyng a circular duct test section (Dsub(i)=7,5 mm). Two heated lengths of the test section have been considered (L = 2300 and 1180 mm). Experimental data have shown the complete inadequacy of steady-state critical heat flux correlations in predicting the onset of boiling crisis during fast flow transients (half-flow decay time, tsub(h)lt5.0-6.0 s). The flow transient does not show dependence, in terms of DNB conditions ,upon the length of the test section: the ratio between transient and steady-state critical mass flowrate is not dependent on the tested geometry. The time interval from the start of the flowrate transient to the onset of DNB (time to crisis), has been experimentally determined for all the runs. Data analysis for a better theoretical prediction of the phenomenon has been accomplished, and a design correlation for DNB conditons and time to crisis prediction has been proposed

  9. A transient kinetic study of nickel-catalyzed methanation: Final report

    International Nuclear Information System (INIS)

    Hoost, T.E.; Goodwin, J.G. Jr.

    1988-11-01

    The results of this study are in two major parts. In Part I the use of steady-state isotopic transients of multiple elements (C, H, and O) under actual methanation reaction conditions has permitted an assessment of the reactivity of water on a Ni powder catalyst. It was concluded based on the addition of isotopic water that oxygen, once reacted to form water, is able to readsorb even where the surface coverage of CO remains high. At the low relative partial pressures of water used, however, there was no effect of added water on the formation of methane. The surface residence time of water determined from isotopic transients contains the residence time on the surface during the primary formation reaction as well as the time spent during readsorption(s). Part II addressed how a catalyst modifier (in this case Cl) affects methanation in CO hydrogenation using steady-state isotopic transient kinetic analysis (SSITKA) of methanation. The results obtained using silica-supported Ru suggest the structural rearrangements induced by the presence of chlorine, rather than selective site blocking or electronic interactions, may be the primary mechanism of chlorine modification of the catalytic properties of supported metals for CO hydrogenation. SSITKA indicated that the decrease in methanation activity with increasing initial Cl concentration was a function of a decrease in the number of reactive surface intermediates (or sites) and not of a change in site activity. 36 refs., 10 figs., 5 tabs

  10. Preliminary Assessment of Transient of Over Power Accident for DSFR-600

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Bae, Moohoon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    TRACE code was selected as one of candidates for audit code, so sodium properties and heat transfer model in the code was verified first. On the basis of MARS-LMR code input, DSFR-600 TRACE model was developed and applied to PHTS tube rupture case, one of design base events (DBE) of DSFR-600. In this study, Transients of Over Power (TOP) event is assessed using TRACE code as one another case of DBEs of DSFR-600 for preparation of audit calculation of PGSFR.One of the design base events, transients of over power of Demonstration Sodium cooled Fast Reactor was simulated using TRACE code. Predicted fuel temperature showed that the peak fuel temperature occurs when the reactor scrammed and predicted temperature was similar to the MARS-LMRs assessment by KAERI. In this study, it is found that the second peak of fuel temperature is influenced by the inventory of steam generator and the natural circulation characteristic of the reactor vessel pool. Pre-calculation of the unprotected transients of over power with conservative reactivity assumption showed that this assumption is conservative in design base even assessment. However the method of measurement and applying the core radial, fuel and control rod axial expansion reactivity feedback is crucial in BDBE assessment of SFR.

  11. Reactive Collision Avoidance Algorithm

    Science.gov (United States)

    Scharf, Daniel; Acikmese, Behcet; Ploen, Scott; Hadaegh, Fred

    2010-01-01

    The reactive collision avoidance (RCA) algorithm allows a spacecraft to find a fuel-optimal trajectory for avoiding an arbitrary number of colliding spacecraft in real time while accounting for acceleration limits. In addition to spacecraft, the technology can be used for vehicles that can accelerate in any direction, such as helicopters and submersibles. In contrast to existing, passive algorithms that simultaneously design trajectories for a cluster of vehicles working to achieve a common goal, RCA is implemented onboard spacecraft only when an imminent collision is detected, and then plans a collision avoidance maneuver for only that host vehicle, thus preventing a collision in an off-nominal situation for which passive algorithms cannot. An example scenario for such a situation might be when a spacecraft in the cluster is approaching another one, but enters safe mode and begins to drift. Functionally, the RCA detects colliding spacecraft, plans an evasion trajectory by solving the Evasion Trajectory Problem (ETP), and then recovers after the collision is avoided. A direct optimization approach was used to develop the algorithm so it can run in real time. In this innovation, a parameterized class of avoidance trajectories is specified, and then the optimal trajectory is found by searching over the parameters. The class of trajectories is selected as bang-off-bang as motivated by optimal control theory. That is, an avoiding spacecraft first applies full acceleration in a constant direction, then coasts, and finally applies full acceleration to stop. The parameter optimization problem can be solved offline and stored as a look-up table of values. Using a look-up table allows the algorithm to run in real time. Given a colliding spacecraft, the properties of the collision geometry serve as indices of the look-up table that gives the optimal trajectory. For multiple colliding spacecraft, the set of trajectories that avoid all spacecraft is rapidly searched on

  12. Mannuronic Acids : Reactivity and Selectivity

    NARCIS (Netherlands)

    Codee, Jeroen D. C.; Walvoort, Marthe T. C.; de Jong, Ana-Rae; Lodder, Gerrit; Overkleeft, Herman S.; van der Marel, Gijsbert A.

    2011-01-01

    This review describes our recent studies toward the reactivity and selectivity of mannopyranosyl uronic acid donors, which have been found to be very powerful donors for the construction of beta-mannosidic linkages.

  13. Fuel Temperature Coefficient of Reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Loewe, W.E.

    2001-07-31

    A method for measuring the fuel temperature coefficient of reactivity in a heterogeneous nuclear reactor is presented. The method, which is used during normal operation, requires that calibrated control rods be oscillated in a special way at a high reactor power level. The value of the fuel temperature coefficient of reactivity is found from the measured flux responses to these oscillations. Application of the method in a Savannah River reactor charged with natural uranium is discussed.

  14. The impact of fuel temperature reactivity coefficient on loss of reactivity control accident

    International Nuclear Information System (INIS)

    Park, J. H.; Ryu, E. H.; Song, Y. M.; Jung, J. Y.

    2012-01-01

    Nuclear reactors experience small power fluctuations or anticipated operational transients during even normal power operation. During normal operation, the reactivity is mainly controlled by liquid zone controllers, adjuster rods, mechanical control absorbers, and moderator poison. Even when the reactor power is increased abruptly and largely from an accident and when reactor control systems cannot be actuated quickly due to a fast transient, the reactor should be controlled and stabilized by its inherent safety parameter, such as a negative PCR (Power Coefficient of Reactivity) feedback. A PWR (Pressurized Water Reactor), it is well designed for the reactor to have a negative PCR so that the reactor can be safely shut down or stabilized whenever an abrupt reactivity insertion into the reactor core occurs or the reactor power is abruptly increased. However, it is known that a CANDU reactor has a small amount of PCR, as either negative or positive, because of the different design basis and safety concepts from a PWR. CNSC's regulatory and safety regime has stated that; The PCR of CANDU reactors does not pose a significant risk. Consistent with Canadian nuclear safety requirements, nuclear power plants must have an appropriate combination of inherent and engineered safety features incorporated into the design of the reactor safety and control systems. A reactor design that has a PCR is quite acceptable provided that the reactor is stable against power fluctuations, and that the probability and consequences of any potential accidents that would be aggravated by a positive reactivity feedback are maintained within CNSCprescribed limits. Recently, it was issued licensing the refurbished Wolsong unit 1 in Korea to be operated continuously after its design lifetime in which the calculated PCR was shown to have a small positive value by applying the recent physics code systems, which are composed of WIMS IST, DRAGON IST, and RFSP IST. These code systems were transferred

  15. Some safety related characteristics of Phenix, a 250 MWe fast reactor -1989 and 1990 negative reactivity trip investigations

    International Nuclear Information System (INIS)

    Chaumont, J.M.; Goux, D.; Martin, L.

    1993-01-01

    The main characteristics of the Phenix core control are summarized. The current state of the investigations related to the 1989 and 1990 negative reactivity transients are presented with emphasis on the results of the very low power tests recently performed. (authors). 5 figs., 2 refs

  16. A randomized controlled trial to test the effect of multispecies probiotics on cognitive reactivity to sad mood

    NARCIS (Netherlands)

    Steenbergen, L.; Sellaro, R.; van Hemert, S.; Bosch, J.A.; Colzato, L.S.

    2015-01-01

    Background: Recent insights into the role of the human microbiota in cognitive and affective functioning have led to the hypothesis that probiotic supplementation may act as an adjuvant strategy to ameliorate or prevent depression. Objective: Heightened cognitive reactivity to normal, transient

  17. Comprehensive Reactive Power Support of DFIG Adapted to Different Depth of Voltage Sags

    Directory of Open Access Journals (Sweden)

    Yangwu Shen

    2017-06-01

    Full Text Available The low voltage ride-through (LVRT capability of the doubly-fed induction generator (DFIG significantly impacts upon the integration of wind power into the power grid. This paper develops a novel comprehensive control strategy to enhance the LVRT and reactive power support capacities of the DFIG by installing the energy storage system (ESS. The ESS is connected to the DC-link capacitor of the DFIG and used to regulate the DC-link voltage during normal or fault operations. The unbalanced power between the captured wind power and the power injected to the grid during the transient process is absorbed or compensated by the ESS. The rotor-side converter (RSC is used to control the maximum power production and the grid-side converter (GSC is used to control the reactive power before participating in the voltage support. When the supply voltage continues to drop, the rotor speed is increased by controlling the RSC to realize the LVRT capability and help the GSC further enhance the reactive power support capability. The capacity of the GSC is dedicated to injecting the reactive power to the grid. An auxiliary transient pitch angle controller is proposed to protect the generator’s over speed. Both RSC and GSC act as reactive power sources to further enhance the voltage support capability with serious voltage sags. Simulations based on a single-machine infinite-bus power system verify the effectiveness of the developed comprehensive control strategy.

  18. [Hyper-reactive malarial splenomegaly].

    Science.gov (United States)

    Maazoun, F; Deschamps, O; Barros-Kogel, E; Ngwem, E; Fauchet, N; Buffet, P; Froissart, A

    2015-11-01

    Hyper-reactive malarial splenomegaly is a rare and severe form of chronic malaria. This condition is a common cause of splenomegaly in endemic areas. The pathophysiology of hyper-reactive malarial splenomegaly involves an intense immune reaction (predominantly B cell-driven) to repeated/chronic infections with Plasmodium sp. The diagnosis may be difficult, due to a poorly specific clinical presentation (splenomegaly, fatigue, cytopenias), a long delay between residence in a malaria-endemic area and onset of symptoms, and a frequent absence of parasites on conventional thin and thick blood smears. A strongly contributive laboratory parameter is the presence of high levels of total immunoglobulin M. When the diagnostic of hyper-reactive malarial splenomegaly is considered, search for anti-Plasmodium antibodies and Plasmodium nucleic acids (genus and species) by PCR is useful. Diagnosis of hyper-reactive malarial splenomegaly relies on the simultaneous presence of epidemiological, clinical, biological and follow-up findings. Regression of both splenomegaly and hypersplenism following antimalarial therapy allows the differential diagnosis with splenic lymphoma, a common complication of hyper-reactive malarial splenomegaly. Although rare in Western countries, hyper-reactive malarial splenomegaly deserves increased medical awareness to reduce the incidence of incorrect diagnosis, to prevent progression to splenic lymphoma and to avoid splenectomy. Copyright © 2015 Société nationale française de médecine interne (SNFMI). Published by Elsevier SAS. All rights reserved.

  19. Parametric study of a reactivity accident in a pressurized water reactor: control rod cluster ejection

    International Nuclear Information System (INIS)

    Chesnel, A.

    1985-01-01

    This research thesis concerns a class 4 accident in a PWR: the ejection of a control rod cluster from the reactor core. It aims at defining, for such an accident, the envelope values which relate the reactivity to the hot spot factor within the frame of a mode A control. The report describes the physical phenomena and their modelling during the considered transient. It presents a simple mathematical solution of the accident which shows that the main neutron parameters are the released reactivity, the delayed neutron fraction, the Doppler coefficient, and the hot spot factor. It reports a temperature sensitivity study, and discusses three-dimensional calculations of irradiation distributions

  20. Transient diabetes insipidus in pregnancy

    Science.gov (United States)

    Gunawardana, Kavinga; Grossman, Ashley

    2015-01-01

    Summary Gestational diabetes insipidus (DI) is a rare complication of pregnancy, usually developing in the third trimester and remitting spontaneously 4–6 weeks post-partum. It is mainly caused by excessive vasopressinase activity, an enzyme expressed by placental trophoblasts which metabolises arginine vasopressin (AVP). Its diagnosis is challenging, and the treatment requires desmopressin. A 38-year-old Chinese woman was referred in the 37th week of her first single-gestation due to polyuria, nocturia and polydipsia. She was known to have gestational diabetes mellitus diagnosed in the second trimester, well-controlled with diet. Her medical history was unremarkable. Physical examination demonstrated decreased skin turgor; her blood pressure was 102/63 mmHg, heart rate 78 beats/min and weight 53 kg (BMI 22.6 kg/m2). Laboratory data revealed low urine osmolality 89 mOsmol/kg (350–1000), serum osmolality 293 mOsmol/kg (278–295), serum sodium 144 mmol/l (135–145), potassium 4.1 mmol/l (3.5–5.0), urea 2.2 mmol/l (2.5–6.7), glucose 3.5 mmol/l and HbA1c 5.3%. Bilirubin, alanine transaminase, alkaline phosphatase and full blood count were normal. The patient was started on desmopressin with improvement in her symptoms, and normalisation of serum and urine osmolality (280 and 310 mOsmol/kg respectively). A fetus was delivered at the 39th week without major problems. After delivery, desmopressin was stopped and she had no further evidence of polyuria, polydipsia or nocturia. Her sodium, serum/urine osmolality at 12-weeks post-partum were normal. A pituitary magnetic resonance imaging (MRI) revealed the neurohypophyseal T1-bright spot situated ectopically, with a normal adenohypophysis and infundibulum. She remains clinically well, currently breastfeeding, and off all medication. This case illustrates some challenges in the diagnosis and management of transient gestational DI. Learning points Gestational DI is a rare complication of

  1. Transient diabetes insipidus in pregnancy.

    Science.gov (United States)

    Marques, Pedro; Gunawardana, Kavinga; Grossman, Ashley

    2015-01-01

    Gestational diabetes insipidus (DI) is a rare complication of pregnancy, usually developing in the third trimester and remitting spontaneously 4-6 weeks post-partum. It is mainly caused by excessive vasopressinase activity, an enzyme expressed by placental trophoblasts which metabolises arginine vasopressin (AVP). Its diagnosis is challenging, and the treatment requires desmopressin. A 38-year-old Chinese woman was referred in the 37th week of her first single-gestation due to polyuria, nocturia and polydipsia. She was known to have gestational diabetes mellitus diagnosed in the second trimester, well-controlled with diet. Her medical history was unremarkable. Physical examination demonstrated decreased skin turgor; her blood pressure was 102/63 mmHg, heart rate 78 beats/min and weight 53 kg (BMI 22.6 kg/m(2)). Laboratory data revealed low urine osmolality 89 mOsmol/kg (350-1000), serum osmolality 293 mOsmol/kg (278-295), serum sodium 144 mmol/l (135-145), potassium 4.1 mmol/l (3.5-5.0), urea 2.2 mmol/l (2.5-6.7), glucose 3.5 mmol/l and HbA1c 5.3%. Bilirubin, alanine transaminase, alkaline phosphatase and full blood count were normal. The patient was started on desmopressin with improvement in her symptoms, and normalisation of serum and urine osmolality (280 and 310 mOsmol/kg respectively). A fetus was delivered at the 39th week without major problems. After delivery, desmopressin was stopped and she had no further evidence of polyuria, polydipsia or nocturia. Her sodium, serum/urine osmolality at 12-weeks post-partum were normal. A pituitary magnetic resonance imaging (MRI) revealed the neurohypophyseal T1-bright spot situated ectopically, with a normal adenohypophysis and infundibulum. She remains clinically well, currently breastfeeding, and off all medication. This case illustrates some challenges in the diagnosis and management of transient gestational DI. Gestational DI is a rare complication of pregnancy occurring in two to four out of

  2. Insights into the reactivation of cobalamin-dependent methionine synthase

    Energy Technology Data Exchange (ETDEWEB)

    Koutmos, Markos; Datta, Supratim; Pattridge, Katherine A.; Smith, Janet L.; Matthews, Rowena G.; (Michigan)

    2009-12-10

    Cobalamin-dependent methionine synthase (MetH) is a modular protein that catalyzes the transfer of a methyl group from methyltetrahydrofolate to homocysteine to produce methionine and tetrahydrofolate. The cobalamin cofactor, which serves as both acceptor and donor of the methyl group, is oxidized once every {approx}2,000 catalytic cycles and must be reactivated by the uptake of an electron from reduced flavodoxin and a methyl group from S-adenosyl-L-methionine (AdoMet). Previous structures of a C-terminal fragment of MetH (MetH{sup CT}) revealed a reactivation conformation that juxtaposes the cobalamin- and AdoMet-binding domains. Here we describe 2 structures of a disulfide stabilized MetH{sup CT} ({sub s-s}MetH{sup CT}) that offer further insight into the reactivation of MetH. The structure of {sub s-s}MetH{sup CT} with cob(II)alamin and S-adenosyl-L-homocysteine represents the enzyme in the reactivation step preceding electron transfer from flavodoxin. The structure supports earlier suggestions that the enzyme acts to lower the reduction potential of the Co(II)/Co(I) couple by elongating the bond between the cobalt and its upper axial water ligand, effectively making the cobalt 4-coordinate, and illuminates the role of Tyr-1139 in the stabilization of this 4-coordinate state. The structure of {sub s-s}MetH{sub CT} with aquocobalamin may represent a transient state at the end of reactivation as the newly remethylated 5-coordinate methylcobalamin returns to the 6-coordinate state, triggering the rearrangement to a catalytic conformation.

  3. Four RNA families with functional transient structures.

    Science.gov (United States)

    Zhu, Jing Yun A; Meyer, Irmtraud M

    2015-01-01

    Protein-coding and non-coding RNA transcripts perform a wide variety of cellular functions in diverse organisms. Several of their functional roles are expressed and modulated via RNA structure. A given transcript, however, can have more than a single functional RNA structure throughout its life, a fact which has been previously overlooked. Transient RNA structures, for example, are only present during specific time intervals and cellular conditions. We here introduce four RNA families with transient RNA structures that play distinct and diverse functional roles. Moreover, we show that these transient RNA structures are structurally well-defined and evolutionarily conserved. Since Rfam annotates one structure for each family, there is either no annotation for these transient structures or no such family. Thus, our alignments either significantly update and extend the existing Rfam families or introduce a new RNA family to Rfam. For each of the four RNA families, we compile a multiple-sequence alignment based on experimentally verified transient and dominant (dominant in terms of either the thermodynamic stability and/or attention received so far) RNA secondary structures using a combination of automated search via covariance model and manual curation. The first alignment is the Trp operon leader which regulates the operon transcription in response to tryptophan abundance through alternative structures. The second alignment is the HDV ribozyme which we extend to the 5' flanking sequence. This flanking sequence is involved in the regulation of the transcript's self-cleavage activity. The third alignment is the 5' UTR of the maturation protein from Levivirus which contains a transient structure that temporarily postpones the formation of the final inhibitory structure to allow translation of maturation protein. The fourth and last alignment is the SAM riboswitch which regulates the downstream gene expression by assuming alternative structures upon binding of SAM. All

  4. Transient cognitive dynamics, metastability, and decision making.

    Directory of Open Access Journals (Sweden)

    Mikhail I Rabinovich

    2008-05-01

    Full Text Available The idea that cognitive activity can be understood using nonlinear dynamics has been intensively discussed at length for the last 15 years. One of the popular points of view is that metastable states play a key role in the execution of cognitive functions. Experimental and modeling studies suggest that most of these functions are the result of transient activity of large-scale brain networks in the presence of noise. Such transients may consist of a sequential switching between different metastable cognitive states. The main problem faced when using dynamical theory to describe transient cognitive processes is the fundamental contradiction between reproducibility and flexibility of transient behavior. In this paper, we propose a theoretical description of transient cognitive dynamics based on the interaction of functionally dependent metastable cognitive states. The mathematical image of such transient activity is a stable heteroclinic channel, i.e., a set of trajectories in the vicinity of a heteroclinic skeleton that consists of saddles and unstable separatrices that connect their surroundings. We suggest a basic mathematical model, a strongly dissipative dynamical system, and formulate the conditions for the robustness and reproducibility of cognitive transients that satisfy the competing requirements for stability and flexibility. Based on this approach, we describe here an effective solution for the problem of sequential decision making, represented as a fixed time game: a player takes sequential actions in a changing noisy environment so as to maximize a cumulative reward. As we predict and verify in computer simulations, noise plays an important role in optimizing the gain.

  5. Reactive sites influence in PMMA oligomers reactivity: a DFT study

    Science.gov (United States)

    Paz, C. V.; Vásquez, S. R.; Flores, N.; García, L.; Rico, J. L.

    2018-01-01

    In this work, we present a theoretical study of methyl methacrylate (MMA) living anionic polymerization. The study was addressed to understanding two important experimental observations made for Michael Szwarc in 1956. The unexpected effect of reactive sites concentration in the propagation rate, and the self-killer behavior of MMA (deactivating of living anionic polymerization). The theoretical calculations were performed by density functional theory (DFT) to obtain the frontier molecular orbitals values. These values were used to calculate and analyze the chemical interaction descriptors in DFT-Koopmans’ theorem. As a result, it was observed that the longest chain-length species (related with low concentration of reactive sites) exhibit the highest reactivity (behavior associated with the increase of the propagation rate). The improvement in this reactivity was attributed to the crosslinking produced in the polymethyl methacrylate chains. Meanwhile, the self-killer behavior was associated with the intermolecular forces present in the reactive sites. This behavior was associated to an obstruction in solvation, since the active sites remained active through all propagation species. The theoretical results were in good agreement with the Szwarc experiments.

  6. Development and verification of an efficient spatial neutron kinetics method for reactivity-initiated event analyses

    International Nuclear Information System (INIS)

    Ikeda, Hideaki; Takeda, Toshikazu

    2001-01-01

    A space/time nodal diffusion code based on the nodal expansion method (NEM), EPISODE, was developed in order to evaluate transient neutron behavior in light water reactor cores. The present code employs the improved quasistatic (IQS) method for spatial neutron kinetics, and neutron flux distribution is numerically obtained by solving the neutron diffusion equation with the nonlinear iteration scheme to achieve fast computation. A predictor-corrector (PC) method developed in the present study enabled to apply a coarse time mesh to the transient spatial neutron calculation than that applicable in the conventional IQS model, which improved computational efficiency further. Its computational advantage was demonstrated by applying to the numerical benchmark problems that simulate reactivity-initiated events, showing reduction of computational times up to a factor of three than the conventional IQS. The thermohydraulics model was also incorporated in EPISODE, and the capability of realistic reactivity event analyses was verified using the SPERT-III/E-Core experimental data. (author)

  7. Measurements of fuel temperature coefficient of reactivity on a commercial AGR

    International Nuclear Information System (INIS)

    Telford, A.; Bridge, M.J.

    1978-01-01

    Tests have been carried out on the commercial AGR at Hikley Point to determine the fuel temperature coefficient of reactivity, an important safety related parameter. Reactor neutron flux was measured during transients induced by movement of a bank of control rods from one steady position to another. An inverse kinetics analysis was applied to the measured flux to determine the change which occured in core reactivity as the fuel temperature changed. The variation of mean fuel temperature was deduced from the flux transient by means of a nine-plane thermal hydraulics representation of the AGR fuel channel. Results so far obtained confirm the predicted variation of fuel temperature coefficient with butn-up. (author)

  8. The PARET code and the analysis of the SPERT I transients

    Energy Technology Data Exchange (ETDEWEB)

    Woodruff, William L [Argonne National Laboratory, Argonne (United States)

    1983-09-01

    The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients.

  9. The PARET code and the analysis of the SPERT I transients

    International Nuclear Information System (INIS)

    Woodruff, William L.

    1983-01-01

    The PARET code has been adapted for the testing of methods and models and for subsequent use in the analysis of transient behavior in research reactors. Comparisons with the experimental results from the SPERT-I transients are provided. The code has also been applied to the analysis of the IAEA 10 MW benchmark cores for protected and unprotected transients. The PARET code was originally developed for the analysis of the SPERT-III experiments for temperatures and pressures typical of power reactors. This code has now been modified to include a selection of flow instability, departure from nucleate boiling (DNB), single and two-phase heat transfer correlations, and a properties library considered more applicable to the low pressures, temperatures, and flow rates encountered in research reactors. The PARET code provides a coupled thermal, hydraulic, and point kinetics capability with continuous reactivity feedback, and an optional voiding model which estimates the voiding produced by subcooled boiling. The present version of the PARET code provides a convenient means of assessing the various models and correlations proposed for use in the analysis of research reactor behavior. For comparison with experiments the SPERT-I cores B-24/32, B-12/64, and D-12/25 were chosen. The B-24/32 core is similar in design to many plate type research reactors in current operation, and the D-12/25 core is of interest because the test included both nondestructive and destructive transients

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

    International Nuclear Information System (INIS)

    Schikorr, W.M.

    2001-01-01

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

  11. Switching transients in a superconducting coil

    International Nuclear Information System (INIS)

    Owen, E.W.; Shimer, D.W.

    1983-01-01

    A study is made of the transients caused by the fast dump of large superconducting coils. Theoretical analysis, computer simulation, and actual measurements are used. Theoretical analysis can only be applied to the simplest of models. In the computer simulations two models are used, one in which the coil is divided into ten segments and another in which a single coil is employed. The circuit breaker that interrupts the current to the power supply, causing a fast dump, is represented by a time and current dependent conductance. Actual measurements are limited to measurements made incidental to performance tests on the MFTF Yin-yang coils. It is found that the breaker opening time is the critical factor in determining the size and shape of the transient. Instantaneous opening of the breaker causes a lightly damped transient with large amplitude voltages to ground. Increasing the opening time causes the transient to become a monopulse of decreasing amplitude. The voltages at the external terminals are determined by the parameters of the external circuit. For fast opening times the frequency depends on the dump resistor inductance, the circuit capacitance, and the amplitude on the coil current. For slower openings the dump resistor inductance and the current determine the amplitude of the voltage to ground at the terminals. Voltages to ground are less in the interior of the coil, where transients related to the parameters of the coil itself are observed

  12. SPIRITS: Uncovering Unusual Infrared Transients with Spitzer

    International Nuclear Information System (INIS)

    Kasliwal, Mansi M.; Jencson, Jacob E.; Tinyanont, Samaporn; Cao, Yi; Cook, David; Bally, John; Masci, Frank; Armus, Lee; Cody, Ann Marie; Bond, Howard E.; Contreras, Carlos; Dykhoff, Devin A.; Amodeo, Samuel; Carlon, Robert L.; Cass, Alexander C.; Corgan, David T.; Faella, Joseph; Boyer, Martha; Cantiello, Matteo; Fox, Ori D.

    2017-01-01

    We present an ongoing, five-year systematic search for extragalactic infrared transients, dubbed SPIRITS—SPitzer InfraRed Intensive Transients Survey. In the first year, using Spitzer /IRAC, we searched 190 nearby galaxies with cadence baselines of one month and six months. We discovered over 1958 variables and 43 transients. Here, we describe the survey design and highlight 14 unusual infrared transients with no optical counterparts to deep limits, which we refer to as SPRITEs (eSPecially Red Intermediate-luminosity Transient Events). SPRITEs are in the infrared luminosity gap between novae and supernovae, with [4.5] absolute magnitudes between −11 and −14 (Vega-mag) and [3.6]–[4.5] colors between 0.3 mag and 1.6 mag. The photometric evolution of SPRITEs is diverse, ranging from <0.1 mag yr −1 to >7 mag yr −1 . SPRITEs occur in star-forming galaxies. We present an in-depth study of one of them, SPIRITS 14ajc in Messier 83, which shows shock-excited molecular hydrogen emission. This shock may have been triggered by the dynamic decay of a non-hierarchical system of massive stars that led to either the formation of a binary or a protostellar merger.

  13. SPIRITS: Uncovering Unusual Infrared Transients with Spitzer

    Energy Technology Data Exchange (ETDEWEB)

    Kasliwal, Mansi M.; Jencson, Jacob E.; Tinyanont, Samaporn; Cao, Yi; Cook, David [Division of Physics, Mathematics and Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Bally, John [Center for Astrophysics and Space Astronomy, University of Colorado, 389 UCB, Boulder, CO 80309 (United States); Masci, Frank; Armus, Lee [Infrared Processing and Analysis Center, California Institute of Technology, Pasadena, CA 91125 (United States); Cody, Ann Marie [NASA Ames Research Center, Moffett Field, CA 94035 (United States); Bond, Howard E. [Dept. of Astronomy and Astrophysics, Pennsylvania State University, University Park, PA 16802 (United States); Contreras, Carlos [Las Campanas Observatory, Carnegie Observatories, Casilla 601, La Serena (Chile); Dykhoff, Devin A.; Amodeo, Samuel; Carlon, Robert L.; Cass, Alexander C.; Corgan, David T.; Faella, Joseph [Minnesota Institute for Astrophysics, School of Physics and Astronomy, 116 Church Street, S. E., University of Minnesota, Minneapolis, MN 55455 (United States); Boyer, Martha [NASA Goddard Space Flight Center, MC 665, 8800 Greenbelt Road, Greenbelt, MD 20771 (United States); Cantiello, Matteo [Center for Computational Astrophysics, Flatiron Institute, 162 Fifth Avenue, New York, NY 10010 (United States); Fox, Ori D. [Space Telescope Science Institute, 3700 San Martin Drive, Baltimore, MD 21218 (United States); and others

    2017-04-20

    We present an ongoing, five-year systematic search for extragalactic infrared transients, dubbed SPIRITS—SPitzer InfraRed Intensive Transients Survey. In the first year, using Spitzer /IRAC, we searched 190 nearby galaxies with cadence baselines of one month and six months. We discovered over 1958 variables and 43 transients. Here, we describe the survey design and highlight 14 unusual infrared transients with no optical counterparts to deep limits, which we refer to as SPRITEs (eSPecially Red Intermediate-luminosity Transient Events). SPRITEs are in the infrared luminosity gap between novae and supernovae, with [4.5] absolute magnitudes between −11 and −14 (Vega-mag) and [3.6]–[4.5] colors between 0.3 mag and 1.6 mag. The photometric evolution of SPRITEs is diverse, ranging from <0.1 mag yr{sup −1} to >7 mag yr{sup −1}. SPRITEs occur in star-forming galaxies. We present an in-depth study of one of them, SPIRITS 14ajc in Messier 83, which shows shock-excited molecular hydrogen emission. This shock may have been triggered by the dynamic decay of a non-hierarchical system of massive stars that led to either the formation of a binary or a protostellar merger.

  14. Transient particle emission measurement with optical techniques

    Science.gov (United States)

    Bermúdez, Vicente; Luján, José M.; Serrano, José R.; Pla, Benjamín

    2008-06-01

    Particulate matter is responsible for some respiratory and cardiovascular diseases. In addition, it is one of the most important pollutants of high-speed direct injection (HSDI) passenger car engines. Current legislation requires particulate dilution tunnels for particulate matter measuring. However for development work, dilution tunnels are expensive and sometimes not useful since they are not able to quantify real-time particulate emissions during transient operation. In this study, the use of a continuous measurement opacimeter and a fast response HFID is proven to be a good alternative to obtain instantaneous particle mass emissions during transient operation (due to particulate matter consisting mainly of soot and SOF). Some methods and correlations available from literature, but developed for steady conditions, are evaluated during transient operation by comparing with mini-tunnel measurements during the entire MVEG-A transient cycle. A new correlation was also derived from this evaluation. Results for soot and SOF (obtained from the new correlation proposed) are compared with soot and SOF captured with particulate filters, which have been separated by means of an SOF extraction method. Finally, as an example of ECU design strategies using these sort of correlations, the EGR valve opening is optimized during transient operation. The optimization is performed while simultaneously taking into account instantaneous fuel consumption, particulate emissions (calculated with the proposed correlation) and other regulated engine pollutants.

  15. Loss-of-flow transient characterization in carbide-fueled LMFBRs

    International Nuclear Information System (INIS)

    Rothrock, R.B.; Morgan, M.M.; Baars, R.E.; Elson, J.S.; Wray, M.L.

    1985-01-01

    One of the benefits derived from the use of carbide fuel in advanced Liquid Metal Fast Breeder Reactors (LMFBRs) is a decreased vulnerability to certain accidents. This can be achieved through the combination of advanced fuel performance with the enhanced reactivity feedback effects and passive shutdown cooling systems characteristic of the current 'inherently safe' plant concepts. The calculated core response to an unprotected loss of flow (ULOF) accident has frequently been used as a benchmark test of these designs, and the advantages of a high-conductivity fuel in relation to this type of transient have been noted in previous analyses. To evaluate this benefit in carbide-fueled LMFBRs incorporating representative current plant design features, limited calculations have been made of a ULOF transient in a small ('modular') carbide-fueled LMFBR

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

    International Nuclear Information System (INIS)

    Heckhoff, H.D.

    1981-10-01

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

  17. Assessment of the turbine trip transient in Cofrentes NPP with TRAC-BF1

    International Nuclear Information System (INIS)

    Castrillo, F.; Gomez, A.; Gallego, I.

    1993-06-01

    This report presents the results of the assessment of TRAC-BF1 (G1-J1) code with the model of C. N. Cofrentes for simulation of the transient originated by the manual trip of the main turbine. C. N. Cofrentes is a General Electric designed BWR/6 plant, with a nominal core thermal power of 2894 Mwt, in commercial operation since 1985, owned and operated by Hidroelectrica Espanola, S. A. The plant incorporates all the characteristics of BWR/6 reactors, with two turbine driven FW pumps. As a result of this assessment a model of C. N. Cofrentes has been developed for TRAC-BF1 that fairly reproduces operational transient behavior of the plant. A special purpose code was generated to obtain reactivity coefficients, as required by TRAC-BF1, from the 3D simulator

  18. Feedback control systems for non-linear simulation of operational transients in LMFBRs

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Agrawal, A.K.; Srinivasan, E.S.

    1979-01-01

    Feedback control systems for non-linear simulation of operational transients in LMFBRs are developed. The models include (1) the reactor power control and rod drive mechanism, (2) sodium flow control and pump drive system, (3) steam generator flow control and valve actuator dynamics, and (4) the supervisory control. These models have been incorporated into the SSC code using a flexible approach, in order to accommodate some design dependent variations. The impact of system nonlinearity on the control dynamics is shown to be significant for severe perturbations. Representative result for a 10 cent and 25 cent step insertion of reactivity and a 10% ramp change in load in 40 seconds demonstrate the suitability of this model for study of operational transients without scram in LMFBRs

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

    International Nuclear Information System (INIS)

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

    1975-10-01

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

  20. TRACE/PARCS modelling of rips trip transients for Lungmen ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Chang, C. Y. [Inst. of Nuclear Engineering and Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Lin, H. T.; Wang, J. R. [Inst. of Nuclear Energy Research, No. 1000, Wenhua Rd., Longtan Township, Taoyuan County 32546, Taiwan (China); Shih, C. [Inst. of Nuclear Engineering and Science, Dept. of Engineering and System Science, National Tsing-Hua Univ., No.101, Kuang-Fu Road, Hsinchu 30013, Taiwan (China)

    2012-07-01

    The objectives of this study are to examine the performances of the steady-state results calculated by the Lungmen TRACE/PARCS model compared to SIMULATE-3 code, as well as to use the analytical results of the final safety analysis report (FSAR) to benchmark the Lungmen TRACE/PARCS model. In this study, three power generation methods in TRACE were utilized to analyze the three reactor internal pumps (RIPs) trip transient for the purpose of validating the TRACE/PARCS model. In general, the comparisons show that the transient responses of key system parameters agree well with the FSAR results, including core power, core inlet flow, reactivity, etc. Further studies will be performed in the future using Lungmen TRACE/PARCS model. After the commercial operation of Lungmen nuclear power plant, TRACE/PARCS model will be verified. (authors)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  2. Reactive Strength Index: A Poor Indicator of Reactive Strength?

    Science.gov (United States)

    Healy, Robin; Kenny, Ian; Harrison, Drew

    2017-11-28

    The primary aim was to assess the relationships between reactive strength measures and associated kinematic and kinetic performance variables achieved during drop jumps. A secondary aim was to highlight issues with the use of reactive strength measures as performance indicators. Twenty eight national and international level sprinters, consisting of fourteen men and women, participated in this cross-sectional analysis. Athletes performed drop jumps from a 0.3 m box onto a force platform with dependent variables contact time (CT), landing time (TLand), push-off time (TPush), flight time (FT), jump height (JH), reactive strength index (RSI, calculated as JH / CT), reactive strength ratio (RSR, calculated as FT / CT) and vertical leg spring stiffness (Kvert) recorded. Pearson's correlation test found very high to near perfect relationships between RSI and RSR (r = 0.91 to 0.97), with mixed relationships found between RSI, RSR and the key performance variables, (Men: r = -0.86 to -0.71 between RSI/RSR and CT, r = 0.80 to 0.92 between RSI/RSR and JH; Women: r = -0.85 to -0.56 between RSR and CT, r = 0.71 between RSI and JH). This study demonstrates that the method of assessing reactive strength (RSI versus RSR) may be influenced by the performance strategies adopted i.e. whether an athlete achieves their best reactive strength scores via low CTs, high JHs or a combination. Coaches are advised to limit the variability in performance strategies by implementing upper and / or lower CT thresholds to accurately compare performances between individuals.

  3. Simulation of reactive nanolaminates using reduced models: II. Normal propagation

    Energy Technology Data Exchange (ETDEWEB)

    Salloum, Maher; Knio, Omar M. [Department of Mechanical Engineering, The Johns Hopkins University, Baltimore, MD 21218-2686 (United States)

    2010-03-15

    Transient normal flame propagation in reactive Ni/Al multilayers is analyzed computationally. Two approaches are implemented, based on generalization of earlier methodology developed for axial propagation, and on extension of the model reduction formalism introduced in Part I. In both cases, the formulation accommodates non-uniform layering as well as the presence of inert layers. The equations of motion for the reactive system are integrated using a specially-tailored integration scheme, that combines extended-stability, Runge-Kutta-Chebychev (RKC) integration of diffusion terms with exact treatment of the chemical source term. The detailed and reduced models are first applied to the analysis of self-propagating fronts in uniformly-layered materials. Results indicate that both the front velocities and the ignition threshold are comparable for normal and axial propagation. Attention is then focused on analyzing the effect of a gap composed of inert material on reaction propagation. In particular, the impacts of gap width and thermal conductivity are briefly addressed. Finally, an example is considered illustrating reaction propagation in reactive composites combining regions corresponding to two bilayer widths. This setup is used to analyze the effect of the layering frequency on the velocity of the corresponding reaction fronts. In all cases considered, good agreement is observed between the predictions of the detailed model and the reduced model, which provides further support for adoption of the latter. (author)

  4. Biodecolorization and biodegradation of Reactive Blue by ...

    African Journals Online (AJOL)

    SERVER

    2007-06-18

    Jun 18, 2007 ... Aspergillus sp. effectively decolorized Reactive Blue and other structurally different synthetic dyes. Agitation was found to be an important ... Few chemically different dyes such as Reactive Black (75%), Reactive Yellow (70%),. Reactive Red (33%) and ..... Degradation of azo dyes by the lignin degrading ...

  5. Substation Reactive Power Regulation Strategy

    Science.gov (United States)

    Zhang, Junfeng; Zhang, Chunwang; Ma, Daqing

    2018-01-01

    With the increasing requirements on the power supply quality and reliability of distribution network, voltage and reactive power regulation of substations has become one of the indispensable ways to ensure voltage quality and reactive power balance and to improve the economy and reliability of distribution network. Therefore, it is a general concern of the current power workers and operators that what kind of flexible and effective control method should be used to adjust the on-load tap-changer (OLTC) transformer and shunt compensation capacitor in a substation to achieve reactive power balance in situ, improve voltage pass rate, increase power factor and reduce active power loss. In this paper, based on the traditional nine-zone diagram and combining with the characteristics of substation, a fuzzy variable-center nine-zone diagram control method is proposed and used to make a comprehensive regulation of substation voltage and reactive power. Through the calculation and simulation of the example, this method is proved to have satisfactorily reconciled the contradiction between reactive power and voltage in real-time control and achieved the basic goal of real-time control of the substation, providing a reference value to the practical application of the substation real-time control method.

  6. Separative performance transients in a gas centrifuge

    International Nuclear Information System (INIS)

    Olander, D.R.

    1979-01-01

    A general method has been developed to calculate the behavior of the exit compositions from a gas centrifuge under unsteady conditions. The method utilizes the basic enrichment gradient equations derived by Cohen, which, in this case, contain time derivatives of the partial 235 U inventories. These partial differential equations are converted to ordinary differential equations by a linear approximation to the axial concentration distribution for use in the inventory terms only. With this simplification, analytical solution is possible for the feed concentration transient. The transient driven by a change in the feed flow rate, however, requires numerical solution. For analysis of ideal cascades in the unsteady state, the transient flow and separation characteristics of the centrifuge must be combined with total uranium and 235 U material balances on each stage

  7. Transient effects in friction fractal asperity creep

    CERN Document Server

    Goedecke, Andreas

    2013-01-01

    Transient friction effects determine the behavior of a wide class of mechatronic systems. Classic examples are squealing brakes, stiction in robotic arms, or stick-slip in linear drives. To properly design and understand mechatronic systems of this type, good quantitative models of transient friction effects are of primary interest. The theory developed in this book approaches this problem bottom-up, by deriving the behavior of macroscopic friction surfaces from the microscopic surface physics. The model is based on two assumptions: First, rough surfaces are inherently fractal, exhibiting roughness on a wide range of scales. Second, transient friction effects are caused by creep enlargement of the real area of contact between two bodies. This work demonstrates the results of extensive Finite Element analyses of the creep behavior of surface asperities, and proposes a generalized multi-scale area iteration for calculating the time-dependent real contact between two bodies. The toolset is then demonstrated both...

  8. Transient response in granular bounded heap flows

    Science.gov (United States)

    Xiao, Hongyi; Ottino, Julio M.; Lueptow, Richard M.; Umbanhowar, Paul B.

    2017-11-01

    Heap formation, a canonical granular flow, is common in industry and is also found in nature. Here, we study the transition between steady flow states in quasi-2D bounded heaps by suddenly changing the feed rate from one fixed value to another. During the transition, in both experiments and discrete element method simulations, an additional wedge of flowing particles propagates over the rising free surface. The downstream edge of the wedge - the wedge front - moves downstream with velocity inversely proportional to the square root of time. An additional longer duration transient process continues after the wedge front reaches the downstream wall. The transient flux profile during the entire transition is well modeled by a diffusion-like equation derived from local mass balance and a local linear relation between the flux and the surface slope. Scalings for the transient kinematics during the flow transitions are developed based on the flux profiles. Funded by NSF Grant CBET-1511450.

  9. NALAP: an LMFBR system transient code

    International Nuclear Information System (INIS)

    Martin, B.A.; Agrawal, A.K.; Albright, D.C.; Epel, L.G.; Maise, G.

    1975-07-01

    NALAP is a LMFBR system transient code. This code, adapted from the light water reactor transient code RELAP 3B, simulates thermal-hydraulic response of sodium cooled fast breeder reactors when subjected to postulated accidents such as a massive pipe break as well as a variety of other upset conditions that do not disrupt the system geometry. Various components of the plant are represented by control volumes. These control volumes are connected by junctions some of which may be leak or fill junctions. The fluid flow equations are modeled as compressible, single-stream flow with momentum flux in one dimension. The transient response is computed by integrating the thermal-hydraulic conservation equations from user-initialized operating conditions by an implicit numerical scheme. Point kinetics approximation is used to represent the time dependent heat generation in the reactor core

  10. Modeling of environmentally induced transients within satellites

    Science.gov (United States)

    Stevens, N. John; Barbay, Gordon J.; Jones, Michael R.; Viswanathan, R.

    1987-01-01

    A technique is described that allows an estimation of possible spacecraft charging hazards. This technique, called SCREENS (spacecraft response to environments of space), utilizes the NASA charging analyzer program (NASCAP) to estimate the electrical stress locations and the charge stored in the dielectric coatings due to spacecraft encounter with a geomagnetic substorm environment. This information can then be used to determine the response of the spacecraft electrical system to a surface discharge by means of lumped element models. The coupling into the electronics is assumed to be due to magnetic linkage from the transient currents flowing as a result of the discharge transient. The behavior of a spinning spacecraft encountering a severe substorm is predicted using this technique. It is found that systems are potentially vulnerable to upset if transient signals enter through the ground lines.

  11. Transient risk factors of acute occupational injuries

    DEFF Research Database (Denmark)

    Østerlund, Anna H; Lander, Flemming; Nielsen, Kent

    2017-01-01

    Objectives The objectives of this study were to (i) identify transient risk factors of occupational injuries and (ii) determine if the risk varies with age, injury severity, job task, and industry risk level. Method A case-crossover design was used to examine the effect of seven specific transient...... risk factors (time pressure, disagreement with someone, feeling sick, being distracted by someone, non-routine task, altered surroundings, and broken machinery and materials) for occupational injuries. In the study, 1693 patients with occupational injuries were recruited from a total of 4002...... in relation to sex, age, job task, industry risk level, or injury severity. Conclusion Use of a case-crossover design identified several worker-related transient risk factors (time pressure, feeling sick, being distracted by someone) that led to significantly increased risks for occupational injuries...

  12. Positron beam studies of transients in semiconductors

    International Nuclear Information System (INIS)

    Beling, C.D.; Ling, C.C.; Cheung, C.K.; Naik, P.S.; Zhang, J.D.; Fung, S.

    2006-01-01

    Vacancy-sensing positron deep level transient spectroscopy (PDLTS) is a positron beam-based technique that seeks to provide information on the electronic ionization levels of vacancy defects probed by the positron through the monitoring of thermal transients. The experimental discoveries leading to the concept of vacancy-sensing PDLTS are first reviewed. The major problem associated with this technique is discussed, namely the strong electric fields establish in the near surface region of the sample during the thermal transient which tend to sweep positrons into the contact with negligible defect trapping. New simulations are presented which suggest that under certain conditions a sufficient fraction of positrons may be trapped into ionizing defects rendering PDLTS technique workable. Some suggestions are made for techniques that might avoid the problematic electric field problem, such as optical-PDLTS where deep levels are populated using light and the use of high forward bias currents for trap filling

  13. Transient Exciplex Formation Electron Transfer Mechanism

    Directory of Open Access Journals (Sweden)

    Michael G. Kuzmin

    2011-01-01

    Full Text Available Transient exciplex formation mechanism of excited-state electron transfer reactions is analyzed in terms of experimental data on thermodynamics and kinetics of exciplex formation and decay. Experimental profiles of free energy, enthalpy, and entropy for transient exciplex formation and decay are considered for several electron transfer reactions in various solvents. Strong electronic coupling in contact pairs of reactants causes substantial decrease of activation energy relative to that for conventional long-range ET mechanism, especially for endergonic reactions, and provides the possibility for medium reorganization concatenated to gradual charge shift in contrast to conventional preliminary medium and reactants reorganization. Experimental criteria for transient exciplex formation (concatenated mechanism of excited-state electron transfer are considered. Available experimental data show that this mechanism dominates for endergonic ET reactions and provides a natural explanation for a lot of known paradoxes of ET reactions.

  14. Turbofan compressor dynamics during afterburner transients

    Science.gov (United States)

    Kurkov, A. P.

    1976-01-01

    The effects of afterburner light-off and shut-down transients on the compressor stability are investigated. The reported experimental results are based on detailed high response pressure and temperature measurements on the TF30-P-3 turbofan engine. The tests were performed in an altitude test chamber simulating high altitude engine operation. It is shown that during both types of transients, flow breaks down in the forward part of the fan bypass duct. At a sufficiently low engine inlet pressure this resulted in a compressor stall. Complete flow breakdown within the compressor was preceded by a rotating stall. At some locations in the compressor, rotating stall cells initially extended only through part of the blade span. For the shutdown transient the time between first and last detected occurrence of rotating stall is related to the flow Reynolds number. An attempt was made to deduce the number and speed of propagation of rotating stall cells.

  15. Characterization of electrical appliances in transient state

    Science.gov (United States)

    Wójcik, Augustyn; Winiecki, Wiesław

    2017-08-01

    The article contains the study about electrical appliance characterization on the basis of power grid signals. To represent devices, parameters of current and voltage signals recorded during transient states are used. In this paper only transients occurring as a result of switching on devices are considered. The way of data acquisition performed in specialized measurement setup developed for electricity load monitoring is described. The paper presents the method of transients detection and the method of appliance parameters calculation. Using the set of acquired measurement data and appropriate software the set of parameters for several household appliances operating in different operating conditions was processed. Usefulness of appliances characterization in Non-Intrusive Appliance Load Monitoring System (NIALMS) with the use of proposed method is discussed focusing on obtained results.

  16. The OECD/NEA/NSC PBMR coupled neutronics/thermal hydraulics transient benchmark: The PBMR-400 core design

    International Nuclear Information System (INIS)

    Reitsma, F.; Ivanov, K.; Downar, T.; De Haas, H.; Gougar, H. D.

    2006-01-01

    The Pebble Bed Modular Reactor (PBMR) is a High-Temperature Gas-cooled Reactor (HTGR) concept to be built in South Africa. As part of the verification and validation program the definition and execution of code-to-code benchmark exercises are important. The Nuclear Energy Agency (NEA) of the Organisation for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor (PBMR) coupled neutronics/thermal hydraulics transient benchmark problem in its program. The OECD benchmark defines steady-state and transients cases, including reactivity insertion transients. It makes use of a common set of cross sections (to eliminate uncertainties between different codes) and includes specific simplifications to the design to limit the need for participants to introduce approximations in their models. In this paper the detailed specification is explained, including the test cases to be calculated and the results required from participants. (authors)

  17. Platelet activation, function, and reactivity in atherosclerotic carotid artery stenosis: a systematic review of the literature.

    LENUS (Irish Health Repository)

    Kinsella, J A

    2012-09-27

    An important proportion of transient ischemic attack or ischemic stroke is attributable to moderate or severe (50-99%) atherosclerotic carotid stenosis or occlusion. Platelet biomarkers have the potential to improve our understanding of the pathogenesis of vascular events in this patient population. A detailed systematic review was performed to collate all available data on ex vivo platelet activation and platelet function\\/reactivity in patients with carotid stenosis. Two hundred thirteen potentially relevant articles were initially identified; 26 manuscripts met criteria for inclusion in this systematic review. There was no consistent evidence of clinically informative data from urinary or soluble blood markers of platelet activation in patients with symptomatic moderate or severe carotid stenosis who might be considered suitable for carotid intervention. Data from flow cytometry studies revealed evidence of excessive platelet activation in patients in the early, sub-acute, or late phases after transient ischemic attack or stroke in association with moderate or severe carotid stenosis and in asymptomatic moderate or severe carotid stenosis compared with controls. Furthermore, pilot data suggest that platelet activation may be increased in recently symptomatic than in asymptomatic severe carotid stenosis. Excessive platelet activation and platelet hyperreactivity may play a role in the pathogenesis of first or subsequent transient ischemic attack or stroke in patients with moderate or severe carotid stenosis. Larger longitudinal studies assessing platelet activation status with flow cytometry and platelet function\\/reactivity in symptomatic vs. asymptomatic carotid stenosis are warranted to improve our understanding of the mechanisms responsible for transient ischemic attack or stroke.

  18. Transient Seepage for Levee Engineering Analyses

    Science.gov (United States)

    Tracy, F. T.

    2017-12-01

    Historically, steady-state seepage analyses have been a key tool for designing levees by practicing engineers. However, with the advances in computer modeling, transient seepage analysis has become a potentially viable tool. A complication is that the levees usually have partially saturated flow, and this is significantly more complicated in transient flow. This poster illustrates four elements of our research in partially saturated flow relating to the use of transient seepage for levee design: (1) a comparison of results from SEEP2D, SEEP/W, and SLIDE for a generic levee cross section common to the southeastern United States; (2) the results of a sensitivity study of varying saturated hydraulic conductivity, the volumetric water content function (as represented by van Genuchten), and volumetric compressibility; (3) a comparison of when soils do and do not exhibit hysteresis, and (4) a description of proper and improper use of transient seepage in levee design. The variables considered for the sensitivity and hysteresis studies are pore pressure beneath the confining layer at the toe, the flow rate through the levee system, and a levee saturation coefficient varying between 0 and 1. Getting results for SEEP2D, SEEP/W, and SLIDE to match proved more difficult than expected. After some effort, the results matched reasonably well. Differences in results were caused by various factors, including bugs, different finite element meshes, different numerical formulations of the system of nonlinear equations to be solved, and differences in convergence criteria. Varying volumetric compressibility affected the above test variables the most. The levee saturation coefficient was most affected by the use of hysteresis. The improper use of pore pressures from a transient finite element seepage solution imported into a slope stability computation was found to be the most grievous mistake in using transient seepage in the design of levees.

  19. Transient behaviour of small HTR for cogeneration

    International Nuclear Information System (INIS)

    Verkerk, E.C.; Van Heek, A.I.

    2000-01-01

    The Dutch market for combined generation of heat and power identifies a unit size of 40 MW thermal for the conceptual design of a nuclear cogeneration plant. The ACACIA system provides 14 MWe electricity combined with 17 t/h of high temperature steam (220 deg C, 10 bar) with a pebble-bed high temperature reactor directly coupled with a helium compressor and a helium turbine. The design of this small CHP unit that is used for industrial applications is mainly based on a pre-feasibility study in 1996, performed by a joint working group of five Dutch organisations, in which technical feasibility was shown. Thermal hydraulic and reactor physics analyses show favourable control characteristics during normal operation and a benign response to loss of helium coolant and loss of flow conditions. Throughout the response on these highly infrequent conditions, ample margin exists between the highest fuel temperatures and the temperature above which fuel degradation will occur. To come to quantitative statements about the ACACIA transient behaviour, a calculational coupling between the high temperature reactor core analysis code package PANTHER/DIREKT and the thermal hydraulic code RELAP5 for the energy conversion system has been made. This coupling offers a more realistic simulation of the entire system, since it removes the necessity of forcing boundary conditions on the simulation models at the data transfer points. In this paper, the models used for the dynamic components of the energy conversion system are described, and the results of the calculation for two operational transients in order to demonstrate the effects of the interaction between reactor core and its energy conversion system are shown. Several transient cases that are representative as operational transients for an HTR will be discussed, including one representing a load rejection case that shows the functioning of the control system, in particular the bypass valve. Another transient is a load following

  20. The reactivity of natural phenols

    Energy Technology Data Exchange (ETDEWEB)

    Denisov, Evgenii T; Denisova, Taisa G [Institute of Problems of Chemical Physics, Russian Academy of Sciences, Chernogolovka, Moscow Region (Russian Federation)

    2009-11-30

    This review surveys physicochemical data of natural phenols published in recent years. The structures of some compounds of this class are given. A complete set of the dissociation energies of the O-H bonds for 71 natural phenols is presented. Kinetic characteristics of the reactions of peroxyl, alkyl and thiyl radicals with natural phenols, exchange reactions of phenoxyl radicals with phenols and reactions of phenoxyl radicals with lipids, hydroperoxides, cysteine and ascorbic acid are compiled and described systematically. The reactivity of phenols in radical reactions and the factors that determine the reactivity (the enthalpy of reaction, triplet repulsion, the electronegativities of atoms at the reaction centre, the presence of pi-electrons adjacent to the reaction centre, the radii of atoms at the reaction centre, steric hindrance, the force constants of the reacting bonds) are discussed. An important role of hydrogen bonding between surrounding molecules and the OH groups of natural phenols in decreasing their reactivities is noted.