Jose A March-Leuba

A reduced-order model of boiling water reactor linear dynamics

The development of a reduced-order linear model of the linear dynamic behavior of boiling water reactors (BWRs) is reported. The model is based on a detailed study of the various physical dynamic p...

A MODAL-BASED REDUCED-ORDER MODEL OF BWR OUT-OF-PHASE INSTABILITIES

A low-order model of BWR out-of-phase instabilities is developed via modal analysis. The modal point kinetics equations are derived using the fundamental and first harmonic modes of the neutron diffusion equation. Parameter estimation in the frequency domain is used to determine the coefficients of the transfer functions (or differential equations) representing the necessary BWR feedback characteristics. Combined with well-known non-linear reactor kinetics, the resulting simple structure of the model provides physical insight into the mechanisms behind the out-of-phase coupled thermal-hydraulic/neutronic BWR instability. This investigation is specifically oriented toward the LaSalle Unit II BWR. An out-of-phase simulation of the LaSalle BWR illustrates the ability of these equations to effectively reproduce the phenomenon.

Nonlinear Dynamics and Chaos in Boiling Water Reactors

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 irr 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.

Development of an automated approach to control system design

Abstract A research effort to develop methods for automated generation of control systems that can be traced directly to the design requirements is documented. This research is being conducted under the Nuclear Energy Research Initiative for the U.S. Department of Energy. The final goal is to allow the control 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. The automated design approach also lends itself to a control system structure that captures the design requirements and permits the optimum control solution to be maintained during the plant life.

The Modeling of Advanced BWR Fuel Designs with the NRC Fuel Depletion Codes PARCS/PATHS

Abstract The PATHS (PARCS Advanced Thermal Hydraulic Solver) code was developed at the University of Michigan in support of U.S. Nuclear Regulatory Commission research to solve the steady-state, two-phase, thermal-hydraulic equations for a boiling water reactor (BWR) and to provide thermal-hydraulic feedback for BWR depletion calculations with the neutronics code PARCS (Purdue Advanced Reactor Core Simulator). The simplified solution methodology, including a three-equation drift flux formulation and an optimized iteration scheme, yields very fast run times in comparison to conventional thermal-hydraulic systems codes used in the industry, while still retaining sufficient accuracy for applications such as BWR depletion calculations. The capability to model advanced BWR fuel designs with part-length fuel rods and heterogeneous axial channel flow geometry has been implemented in PATHS, and the code has been validated against previously benchmarked advanced core simulators as well as BWR plant and experimental data. The modifications to the codes and the results of the validation are described in this paper.

Detector system for monitoring fissile mass flow in liquids and gases

A computer controlled detector/electronics module has been designed and constructed for the Oak Ridge National Laboratory fissile mass flow (FMF) monitor. The FMF monitor employs a host computer, a modulated neutron source (Cf-252) and four of the gamma-ray detector/electronics modules described here to non-intrusively measure liquid or gas fissile-mass-flow rates in a pipe. The gamma-ray detector/electronics modules each consist of a bismuth germanate/photomultiplier-tube scintillation detector, an integrating preamplifier, a bipolar spectroscopy amplifier, two single channel analyzers, a temperature sensor and an on-board network communications node. The host computer automatically calibrates amplifier gain and single-channel analyzer thresholds via the network using a characteristic emission peak of the material flowing in the pipe and corrects for detector gain variations due to the temperature coefficients of the bismuth germanate. The single channel analyzer output pulses are counted by a microprocessor on board the network node and reported to the host computer. The key components of the detector/electronics module are described and initial data obtained with the FMF monitor are presented.

Separation of spatial effects in frequency-domain subcriticality measurements☆

Abstract This paper documents a technique to isolate modal or spatial effects from frequency-domain quantities that are used for subcriticality measurements. This technique is an enhancement of the 252 Cf-source-driven noise analysis method for subcriticality measurements. A physical model was developed to account for the space-dependence of those frequency-domain quantities. The model has been implemented in a computer code to aid in interpretation of subcriticality experiments where spatial effects are diagnosed or suspected. The concept of fundamental-mode ratio is introduced, and the ratio is shown to be real and independent of frequency. Previously developed analysis techniques to obtain a value for the systems reactivity can be applied directly to the fitted fundamental ratio. Using data from previous experiments, several examples are presented to illustrate the applicability of this technique.

A Study of the Effect of Mixed Cores on the Stability of BWRs

Cores loaded with a mixture of fuel types are known to reduce stability margins. Mixed fuel cores have become more common as utilities change fuel suppliers, or when fuel vendors upgrade their fuel designs to take advantage of improved thermal and mechanical margins. This paper studies some of the physical processes that reduce the stability of mixed cores. A number of runs have been performed using the LAPUR6 stability code to evaluate the effect on mixed cores on the stability of a typical BWR. To this end, two fuel types have been set up with two different single-phase to two-phase pressure drop ratios by artificially adjusting the spacer and inlet orifice friction coefficients. The flow and pressure drop characteristics of both fuels have been matched at full flow, full power conditions. All manufacturers match the pressure drop of new fuels so that the flow distributions among the new and old fuel elements operating at the same power are approximately constant. The critical power ratio and thermo-mechanical criteria are typically limiting at full power; therefore matching the flow performance at full power maximizes the margin to these criteria. Stability is of concern at low flows, especially at natural circulation, where the thermal-hydraulic conditions are significantly different from full flow and power. Our simulations show that even if two fuel elements are perfectly matched at full flow, the axial void fraction distribution changes significantly when the flow is reduced to natural circulation conditions and the two fuel elements are not fully thermal-hydraulically compatible at the reduced flows. Basically, the two fuel types set up two separate natural circulation lines, and one of the fuel types essentially starves the other from flow. Since stability has such a strong dependence with channel flow, the reactor stability is controlled by the fuel type that has the smaller flow at natural circulation. A counterintuitive result of this study shows that, in general, loading a more stable fuel type into a mixed core has the opposite effect, and the stability margin of that mixed core is lower until the new, more stable fuel becomes dominant. Because of the burnable Gadolinium in most modern BWR fuels, the highest reactivity fuel elements are the once-burned. Loading a more stable fuel type starves the flow of the high-reactivity older fuel, reducing the stability margin.© 2010 ASME

Implementation of the Fissile Mass Flow Monitor Source Verification and Confirmation

This report presents the verification procedure for neutron sources installed in U.S. Department of Energy equipment used to measure fissile material flow. The Fissile Mass Flow Monitor (FMFM) equipment determines the {sup 235}U fissile mass flow of UF{sub 6} gas streams by using {sup 252}Cf neutron sources for fission activation of the UF{sub 6} gas and by measuring the fission products in the flow. The {sup 252}Cf sources in each FMFM are typically replaced every 2 to 3 years due to their relatively short half-life ({approx} 2.65 years). During installation of the new FMFM sources, the source identity and neutronic characteristics provided by the manufacturer are verified with the following equipment: (1) a remote-control video television (RCTV) camera monitoring system is used to confirm the source identity, and (2) a neutron detection system (NDS) is used for source-strength confirmation. Use of the RCTV and NDS permits remote monitoring of the source replacement process and eliminates unnecessary radiation exposure. The RCTV, NDS, and the confirmation process are described in detail in this report.

Fissile Mass Flow Monitor Source-Strength Calibration Using the ORNL Neutron Detector System

This paper describes the neutron detector system (NDS) developed to measure the neutron source strength of the 252Cf neutron sources used in the fissile mass flow monitor (FMFM). The FMFM measures the 235U fissile mass flow in the UF6 gas streams and uses 252Cf neutron sources for the fission activation of the UF6 gas. Four FMFMs are operational in three Russian facilities for the U.S. Department of Energys Highly Enriched Uranium Transparency Program. The 252Cf sources are replaced about every 2 years due to their short half-life (~2.65 years). During FMFM source replacement, the new 252Cf sources are calibrated and verified with the previously installed sources (i.e., a relative source mass check) to ensure proper and seamless FMFM performance. The NDS consists of a neutron detector (a commercially available high-efficiency 3He proportional counter) and the electronics, which are commercial nuclear instrument modules. A measurement time of 10 s is sufficient to take the data; the NDS yields about les 0.5% error from the average count taken up to 100 s. Measurement repeatability is good (< 1%), and it is not sensitive (< 0.5%) to the orientation of the source plug inside the NDS polyethylene source plug holder. The NDS is then calibrated for the measured source mass strength, and it is established that the sources can be measured with an overall uncertainty of 1.4% for a typical FMFM source mass of 3 mug. A detailed description of the NDS, its performance characteristics, and results of measurements performed on the latest FMFM sources are presented.