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Radial nodalization effects on BWR (boiling water reactor) stability calculations

Description: Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs.
Date: January 1, 1990
Creator: March-Leuba, J.
Partner: UNT Libraries Government Documents Department

Understanding the boiling water reactor limit cycle

Description: This paper presents an interpretation of the physical mechanisms involved in the development of limit cycle oscillations in boiling water reactors (BWRs). Based on this interpretation, approximate correlations for some oscillation parameters are developed and shown to be largely independent of the particular reactor operating condition. The stability of the limit cycle is also studied in this paper. It is shown that the BWR limit cycle may become unstable and bifurcate. The bifurcation process leads to aperiodic (chaotic) behavior of the reactor power and causes the peak oscillation powers to be larger than those from a nonbifurcated limit cycle. 7 refs., 6 figs., 1 tab.
Date: January 1, 1989
Creator: March-Leuba, J.
Partner: UNT Libraries Government Documents Department

Nonlinear dynamics and chaos in boiling water reactors

Description: There are currently 72 commercial boiling water reactors (BWRs) in operation or under construction in the western world, 37 of them in the United States. Consequently, a great effort has been devoted to the study of BWR systems under a wide range of plant operating conditions. This paper represents a contribution to this ongoing effort; its objective is to study the basic dynamic processes in BWR systems, with special emphasis on the physical interpretation of BWR dynamics. The main thrust in this work is the development of phenomenological BWR models suited for analytical studies performed in conjunction with numerical calculations. This approach leads to a deeper understanding of BWR dynamics and facilitates the interpretation of numerical results given by currently available sophisticated BWR codes. 6 refs., 14 figs., 2 tabs.
Date: January 1, 1988
Creator: March-Leuba, J.
Partner: UNT Libraries Government Documents Department

Importance of momentum dynamics in BWR neutronic stability: experimental evidence

Description: Momentum dynamics affect the boiling water reactor (BWR) neutronic stability by coupling steam void perturbations and core-inlet coolant flow. Computer simulations have shown that proper modeling of the recirculation loop, which shares the upper and lower plena pressures with the reactor core, is essential for accurate stability calculations. Purpose of this paper is to show experimental evidence, obtained from a recent series of stability tests performed at the Browns Ferry-1 BWR, demonstrating the important role of momentum dynamics in BWR neutronic stability.
Date: January 1, 1985
Creator: March-Leuba, J. & Otaduy, P.J.
Partner: UNT Libraries Government Documents Department

Development of an automated diagnostic system for BWR stability measurements

Description: An algorithm capable of automatically evaluating BWR stability has been developed. Main advantages are: Conservative estimate (asymptotic), adjusts to solve difficult conditions, confidence level, and error estimate. The apparent decay ratio (DR) is not a conservative estimate of the reactor stability. The asymptotic decay ratio must be used. Long enough record lengths must be used to reduce the uncertainty of the estimated DR.
Date: October 1, 1984
Creator: March-Leuba, J. & Smith, C.M.
Partner: UNT Libraries Government Documents Department

Advanced Neutron Source Dynamic Model (ANSDM) code description and user guide

Description: A mathematical model is designed that simulates the dynamic behavior of the Advanced Neutron Source (ANS) reactor. Its main objective is to model important characteristics of the ANS systems as they are being designed, updated, and employed; its primary design goal, to aid in the development of safety and control features. During the simulations the model is also found to aid in making design decisions for thermal-hydraulic systems. Model components, empirical correlations, and model parameters are discussed; sample procedures are also given. Modifications are cited, and significant development and application efforts are noted focusing on examination of instrumentation required during and after accidents to ensure adequate monitoring during transient conditions.
Date: August 1, 1995
Creator: March-Leuba, J.
Partner: UNT Libraries Government Documents Department

A parametric analysis of decay ratio calculations in a boiling water reactor model

Description: The results of an investigation of the effects of several parameters on the reactivity instability of a Boiling Water Reactor (BWR) calculational model are summarized. Calculations were performed for a typical BWR operated at low flow conditions, where reactivity instabilities are more likely to occur. The parameters investigated include the axial power shape (characterized by two separate parameters), the core pressure, and operating flow. All calculations were performed using the LAPUR code which was developed at the Oak Ridge National Laboratory for the dynamic modeling of large BWR's. 4 refs., 8 figs.
Date: January 1, 1989
Creator: Blakeman, E.D. & March-Leuba, J.
Partner: UNT Libraries Government Documents Department

A study of out-of-phase power instabilities in boiling water reactors

Description: This paper presents a study of the stability of subcritical neutronic modes in boiling water reactors that can result in out-of-phase power oscillations. A mechanism has been identified for this type of oscillation, and LAPUR code has been modified to account for it. Numerical results show that there is a region in the power-flow operating map where an out-or-phase stability mode is likely even if the core-wide mode is stable. 4 refs., 7 figs.
Date: June 20, 1988
Creator: March-Leuba, J. & Blakeman, E.D.
Partner: UNT Libraries Government Documents Department

Influence of fuel vibration on PWR neutron noise associated with core barrel motion

Description: Ex-core neutron detector noise has been utilized to monitor core support barrel (CSB) vibrations. In order to observe long-term changes, noise signals at Sequoyah-1 were monitored continuously during the whole first fuel cycle and part of the second cycle. Results suggest that neutron noise measurements performed infrequently may not provide adequate surveillance of the CSB because it may be difficult to separate noise amplitude changes due solely to CSB motion from changes caused by fuel motion and burnup. (DLC)
Date: January 1, 1984
Creator: Sweeney, F.J. & March-Leuba, J.
Partner: UNT Libraries Government Documents Department

Physical model of nonlinear noise with application to BWR stability

Description: Within the framework of the present model it is shown that the BWR reactor cannot be unstable in the linear sense, but rather it executes limited power oscillations of a magnitude that depends on the operating conditions. The onset of these oscillations can be diagnosed by the decrease in stochasticity in the power traces and by the appearance of harmonics in the PSD.
Date: January 1, 1983
Creator: March-Leuba, J. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

NERI PROJECT 99-119. TASK 1. ADVANCED CONTROL TOOLS AND METHODS. FINAL REPORT

Description: Nuclear plants of the 21st century will employ higher levels of automation and fault tolerance to increase availability, reduce accident risk, and lower operating costs. Key developments in control algorithms, fault diagnostics, fault tolerance, and communication in a distributed system are needed to implement the fully automated plant. Equally challenging will be integrating developments in separate information and control fields into a cohesive system, which collectively achieves the overall goals of improved performance, safety, reliability, maintainability, and cost-effectiveness. Under the Nuclear Energy Research Initiative (NERI), the U. S. Department of Energy is sponsoring a project to address some of the technical issues involved in meeting the long-range goal of 21st century reactor control systems. This project, ''A New Paradigm for Automated Development Of Highly Reliable Control Architectures For Future Nuclear Plants,'' involves researchers from Oak Ridge National Laboratory, University of Tennessee, and North Carolina State University. This paper documents a research effort to develop methods for automated generation of control systems that can be traced directly to the design requirements. Our final goal is to allow the designer to specify only high-level requirements and stress factors that the control system must survive (e.g. a list of transients, or a requirement to withstand a single failure.) To this end, the ''control engine'' automatically selects and validates control algorithms and parameters that are optimized to the current state of the plant, and that have been tested under the prescribed stress factors. The control engine then automatically generates the control software from validated algorithms. Examples of stress factors that the control system must ''survive'' are: transient events (e.g., set-point changes, or expected occurrences such a load rejection,) and postulated component failures. These stress factors are specified by the designer and become a database of prescribed transients and component failures. The candidate control systems ...
Date: September 9, 2002
Creator: March-Leuba, J.A.
Partner: UNT Libraries Government Documents Department

NERI PROJECT 99-119."A NEW PARADIGM FOR AUTOMATIC DEVELOPMENT OF HIGHLY RELIABLE CONTROL ARCHITECTURES FOR NUCLEAR POWER PLANTS."PHASE-1 PROGRESS REPORT

Description: This report describes the tasks performed and the progress made during Phase 1 of the DOE-NERI project number 99-119 entitled ''Automatic Development of Highly Reliable Control Architecture for Future Nuclear Power Plants''. This project is a collaboration effort between the Oak Ridge National Laboratory (ORNL,) The University of Tennessee, Knoxville (UTK) and the North Carolina State University (NCSU). ORNL is the lead organization and is responsible for the coordination and integration of all work. This research focuses on the development of methods for automated generation of control systems that can be traced directly to the design requirements for the life of the plant. Our final goal is to ''capture'' the design requirements inside a ''control engine'' during the design phase. This control engine is, then, not only capable of designing automatically the initial implementation of the control system, but it also can confirm that the original design requirements are still met during the life of the plant as conditions change. This control engine captures the high-level requirements and stress factors that the control system must survive (e.g. a list of transients, or a requirement to withstand a single failure). The control engine, then, is able to generate automatically the control-system algorithms and parameters that optimize a design goal and satisfy all requirements. As conditions change during the life of the plant (e.g. component degradation, or subsystem failures) the control engine automatically ''flags'' that a requirement is not satisfied, and it can even suggest a modified configuration that would satisfy it. This control engine concept is shown schematically in Fig. 1. The implementation of this ''control-engine'' design methodology requires the following steps, which are described in detail in the attachments to this report: (1) Selection of Design Requirements Related to Control System Performance; (2) Implementation of Requirements in Mathematical Form; (3) ...
Date: August 29, 2000
Creator: March-Leuba, J.A.
Partner: UNT Libraries Government Documents Department

Limit cycles and bifurcations in nuclear systems

Description: This work provides a basis for scoping calculations to determine the dynamic behavior - both linear and nonlinear - of BWRs. Additional work is now underway to establish the feasibility of routine operation of nuclear systems in the nonlinear (limit-cycle) regime.
Date: January 1, 1986
Creator: Cacuci, D.G.; March-Leuba, J. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Contribution of fuel vibrations to ex-core neutron noise during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor

Description: Noise measurements were performed during the first and second fuel cycles of the Sequoyah-1 pressurized water reactor (PWR) to observe long-term changes in the ex-core neutron signatures. Increases in the ex-core neutron noise amplitude were observed throughout the 0.1- to 50.0-Hz range. In-core noise measurements indicate that fuel assembly vibrations contribute significantly to the ex-core neutron noise at nearly all frequencies in this range, probably due to mechanical or acoustic coupling with other vibrating internal structures. Space-dependent kinetics calculations show that ex-core neutron noise induced by fixed-amplitude fuel assembly vibrations will increase over a fuel cycle because of soluble boron and fuel concentration changes associated with burnup. These reactivity effects can also lead to 180/sup 0/ phase shifts between cross-core detectors. We concluded that it may be difficult to separate the changes in neutron noise due to attenuation (shielding) effects of structural vibrations from changes due to reactivity effects of fuel assembly motion on the basis of neutron noise amplitude or phase information. Amplitudes of core support barrel vibrations inferred from ex-core neutron noise measurements using calculated scale factors are likely to have a high degree of uncertainty, since these scale factors usually do not account for neutron noise generated by fuel assembly vibrations. Modifications in fuel management or design may also lead to altered neutron noise signature behavior over a fuel cycle.
Date: January 1, 1984
Creator: Sweeney, F.J.; March-Leuba, J. & Smith, C.M.
Partner: UNT Libraries Government Documents Department

CARDIOGRAMA: a stochastic, semi-empirical methodology for power-reactor surveillance and diagnostics

Description: The utilization of stochastic methods (reactor noise) for power reactor diagnostics and surveillance applications is by now a relatively well-established technique. In this technique, the power spectral density (PSD) of the fluctuations of a specified state variable is often used to define the reactor's signature at a given configuration. The purpose of the present work is to address the problem of handling efficiently the substantial amount of information involved in the application of reactor surveillance and diagnostics methods. Specifically, a methodology is described for: (a) representing the PSDs parametrically, and (b) detecting changes from the reactor's baseline PSD (normal signature).
Date: January 1, 1981
Creator: March-Leuba, J.A.; deSaussure, G. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Performance requirements of the advanced neutron source reactor protection system

Description: Research reactors often have protection-systems performance requirements very different from those of commercial reactors. This paper discusses the special characteristics of the Advanced Neutron Source (ANS) reactor that control these requirements, and it presents some calculations used to quantify this performance.
Date: April 1, 1995
Creator: March-Leuba, J. & Battle, R.E.
Partner: UNT Libraries Government Documents Department

Interpretation of subcriticality measurements with strong spatial effects

Description: A methodology has been developed to account for spatial effects in subcriticality measurements. Using experimental data, this new analysis methodology allows estimation of model contamination without previous knowledge about the system, neither in the form of neutronic or geometric factor calculations. 5 refs., 1 fig.
Date: January 1, 1987
Creator: March-Leuba, C.; March-Leuba, J. & Difilippo, F.C.
Partner: UNT Libraries Government Documents Department

Optimal filtering, parameter tracking, and control of nonlinear nuclear reactors

Description: This paper presents a new formulation of a class of nonlinear optimal control problems in which the system's signals are noisy and some system parameters are changing arbitrarily with time. The methodology is validated with an application to a nonlinear nuclear reactor model. A variational technique based on Pontryagin's Maximum Principle is used to filter the noisy signals, estimate the time-varying parameters, and calculate the optimal controls. The reformulation of the variational technique as an initial value problem allows this microprocessor-based algorithm to perform on-line filtering, parameter tracking, and control.
Date: June 24, 1988
Creator: March-Leuba, C.; March-Leuba, J. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Are limit cycle calculations a stochastic process

Description: Stochasticity is typically associated with processes that produce uncertain results which, in many cases, are due to process nonlinearities and/or extreme sensitivity to initial conditions. By its name, a stochastic process should have a probabilistic or random nature; however, it is well known that many if not all, of the processes that behave stochasticly are indeed deterministic. This is the case with computer calculations to predict the stability of boiling water reactors (BWRs). This paper attempts to introduce the reader to some of the stochastic'' uncertainties involved in this topic, and in particular the errors introduced by the approximations used to integrate numerically the solutions in the time domain. The knowledge of this type of errors is relevant not only in BWR stability calculations but also in time domain calculations involving nonlinear or stochastic processes.
Date: January 1, 1991
Creator: March-Leuba, J.
Partner: UNT Libraries Government Documents Department

Nonlinear dynamics of boiling water reactors

Description: Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters.
Date: January 1, 1983
Creator: March-Leuba, J.; Cacuci, D.G. & Perez, R.B.
Partner: UNT Libraries Government Documents Department

Excitation sources for fuel assembly vibrations in a PWR

Description: Noise measurements from in-core neutron detectors have been utilized previously to monitor in-plant vibrations of pressurized water reactor (PWR) fuel assemblies. Fuel assembly resonant frequencies and mode shapes were obtained from these in-core measurements by observing resonances in the neutron noise power spectral density (PSD) and then plotting the axial dependence of the root mean square (rms) neutron noise over frequency ranges containing these resonances. In order to determine the fuel assembly mode shapes and their relationship to core barrel motion, we performed simultaneous measurements of in-core and ex-core neutron noise at the Sequoyah-1 reactor, a Westinghouse 1150 MW(e) PWR. Analysis of this data indicates that there are two different sources of vibrational excitation for the fuel in the 6- to 8-Hz frequency range. 7 refs., 2 figs.
Date: January 1, 1985
Creator: Sweeney, F.J.; March-Leuba, J.; Wood, R.T. & McNew, C.O.
Partner: UNT Libraries Government Documents Department

Application of noise-analysis methods to monitor stability of boiling water reactors

Description: The dynamic stability of Boiling Water Reactors (BWR's) is influenced by the reactor control system and its interaction with external load demand, channel thermal hydraulic properties, and coupled neutronic-thermal-hydraulic dynamics. The latter aspect of BWR stability which is affected by void reactivity feedback, coolant flow rate and fuel-to-coolant heat transfer characteristics is studied in this paper using the normal fluctuation data. The feasibility of overall core stability trend monitoring using neutron noise and the relatonship between stability and two-phase flow velocity in a fuel channel are studied. Time series modeling of the average power range monitor (APRM) detector signal, and bivariate analysis of adjacent local power range monitor (LPRM) detector signals are used to determine the neutron impulse response, spectral characteristics and two-phase flow velocity using data from an operating BWR. The results of analysis show that the APRM noise signal can be used to monitor changes in the closed-loop output stability of BWRs (but not the absolute stability as determined by the reactivity-to-neutron power transfer function), and that a positive correlation exists between stability and two-phase flow velocity in a fuel channel.
Date: January 1, 1981
Creator: Upadhyaya, B.R.; March-Leuba, J.; Fry, D.N. & Kitamura, M.
Partner: UNT Libraries Government Documents Department

A new approach to controlling the water level of U-tube steam generators

Description: Automatic water level control in steam generators is currently achieved via a three-element controller. This algorithm is based on the measurements of level, steam flow, and feedwater flow. Unfortunately, at low power the feedwater flow signal is highly unreliable, forcing the transfer to manual control. A large number of reactor trips occur under these conditions, and the nuclear industry has shown a concern for this problem. This paper proposes and validates an alternative automatic control algorithm. This new algorithm does not rely on flow signals; it uses instead the pressure measurement in the steam header. A level set point modulation is introduced that allows the algorithm to compensate for shrink and swell phenomena. The standard {Delta}-P algorithm to control feedwater pump speed also has been modified to achieve improved performance and integration. 2 refs., 12 figs.
Date: January 1, 1989
Creator: March-Leuba, C.; March-Leuba, J. & Wilson, T.L.
Partner: UNT Libraries Government Documents Department

Sensitivity of BWR stability calculations to numerical integration techniques

Description: Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters, modeling assumptions, and numerical integration techniques. Following the 1988 LaSalle instability event, a significant industry-wide effort was invested in identifying these sensitivities. One major conclusion from these studies was that existing time-domain codes could best predict BWR stability by using explicit methods for the energy equation with a Courant number as close to unity as possible. This paper presents a series of sensitivity studies using simplified models, which allow us to determine the effect that different numerical integration techniques have on the results of stability calculations. The present study appears to indicate that, even though using explicit integration with a Courant number of one is adequate for existing codes using time-integration steps of less than 10 ms, second-order solution techniques for the time integration can result in significant improvements in the accuracy of linear (i.e., decay ratio) stability calculations.
Date: November 1, 1997
Creator: Peiro, D. G. & March-Leuba, J.
Partner: UNT Libraries Government Documents Department