John E. Meyer

Automatic controller for steam generator water level during low power operation

Abstract This research proposes a new controller which ensures a satisfactory automatic control for the steam generator water level from low power to full power. It is premised that the current analog control loop is replaced with digital computer control thus expanding the range of possible solutions. The proposed approach is to compensate the level measurement for thermal shrink and swell effects which cause complications in level control during low power operation. A non-linear digital predictor is a part of the controller and is used to estimate shrink and swell effects. The predictor is found to be stable and on-line applicable with micro-processors. The controller is evaluated by calculations in which it controls an existing non-linear digital computer model of a steam generator. For a multi-ramp power increase from low power to full power, the proposed controller shows good performances for the entire range. Water level settles down within 3 min after a single ramp increase (5% power increase in a minute) without any stability problem. Even at very low power, the maximum overshoot is judged to be acceptable.

Design and Evaluation of Model-Based Compensators for the Control of Steam Generator Level

The design and evaluation of a controller that uses modelbased compensators to offset the inverse response behavior of water level in the steam generators of nuclear power plants is described. Included is a review of steam generator level dynamics, the development of a model that replicates the effects of feedwater and steam flowrate as well as temperature on steam generator level, the design of both the compensators and the overall controller, and the results of simulation studies in which the performance of this modelbased controller and existing analog ones were compared. The compensator-based controller is stable and its use significantly improves the controllability of steam generator level.

Flow stability analysis of a particle bed reactor fuel element

This paper describes an investigation of thermal hydraulic flow stability in a particle bed reactor fuel element. The work starts from an adaptation of the stability criterion used by Bussard and DeLauer (1958 and 1965). A one‐dimensional evaluation was then performed, using the Ergun relation (1952) to evaluate numerically the pressure drop. If one considers the entire element from cold frit to hot frit, the analysis would indicate flow stability for all conditions. When one uses the pressure drop in the bed for the criterion, the resultant criterion curve has the same shape as the curve obtained by using the Bussard‐DeLauer methodology, but predicts higher flow rates are required for stability. A two‐dimensional thermal hydraulic study used to verify the one‐dimensional analysis yields consistent results. Finally, results from an evaluation to determine if the heat deposition shape within the bed could impact the stability show that, for a chosen average element operating condition, a flat deposition sh...

An economically optimum PWR reload core for a 36-month cycle

Thesis (S.M. and Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1998.

Studies on the closed-loop digital control of multi-modular reactors

This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

Heat transfer conditions in water-cooling of a fusion reactor divertor

The diverters of planned fusion reactors such as ITER require removal of heat fluxes that are large in magnitude and are spatially peaked. Single-side energy deposition on the plasma-facing surface is transferred by conduction and provides an axially and circumferentially non-uniform heat flux on the wall of the cooling channel. Promising candidate cooling systems are based on the use of highly subcooled water (subcooling more than 100/spl deg/C) flowing at very high velocities (more than 5 m/s). These conditions are quite different from those used to obtain existing correlations for single-phase convection and subcooled nucleate boiling. Through experimental assessments this paper suggests that existing correlations can be applied for single-phase convection. However, available data in the range of interest for the subcooled nucleate boiling show that extrapolations of existing correlations cannot be used. This is due to the suppression of nuclear boiling, resulting in high wall temperatures, sometimes far above expected and very close to the temperature of homogeneous nucleation (THN). Based on experimental data, a new correlation is proposed for subcooled nucleate boiling region. The range of application for the new correlation is defined as follows pressure 2-3 MPa, bulk water temperatures 19-25/spl deg/C, heat fluxes up to 25 MW/m/sup 2/, flow velocities from 3 to 15 m/s.

Shield water system design options to improve the safety of fusion reactor blankets

Abstract This paper examines the potential of the shield and the shield cooling system of a fusion reactor to be an effective heat sink inside the reactor vacuum vessel, thereby improving the performance of the fusion reactor blanket during an undercooling accident such as a loss of cooling accident (LOCA) or loss of flow accident (LOFA). The paper demonstrates that having a shield cooling system capable of natural circulation during such an accident can make the difference between a blanket surviving the accident or not. The paper also demonstrates that it is inexpensive to add natural circulation capability to a shield cooling system.

Mechanisms for extreme heat transfer conditions in water-cooling of fusion reactor components

Abstract In existing fusion reactors conceptual designs, water-cooled impurity control components employ operating heat fluxes up to 15 MW m −2 , coolant velocities above 5 m −1 , subcoolings more than 100 K, and pressures below 5 MPa. These conditions are quite different from those most used to obtain existing correlations for subcooled nucleate boiling and critical heat flux. In addition, some available data in the range of interest for fusion reactor components show that extrapolations of the correlations cannot be used. This paper suggests that two heat transfer mechanisms must be incorporated in developing correlations for fusion reactor high heat flux components. First, boiling can be suppressed, resulting in observed non-boiling wall temperatures far above those expected for nucleate boiling. Second, critical heat flux (CHF) can apparently occur with no prior boiling, when the wall temperature reaches the temperature of homogeneous nucleation (THN) while a single phase liquid is adjacent to the wall. A limit of this nature can apparently occur only under conditions of a very high mass flux and a very high bulk subcooling. These two mechanisms must be incorporated in fusion thermal-hydraulics analysis: suppression of nucleate boiling and critical heat flux caused by homogeneous nucleation.

Parity simulation of single phase thermal hydraulics

The analysis of transients in nuclear power plants is a complex problem normally requiring use of simulation tools. Although analog computers have been used for dynamic simulation, the most common approach involves use of a digital computer. An alternative method to attack the same problem, known as parity simulation, is described. Parity simulation, which originated in the study of electronic network transients, exploits the concept of electrical analogs of a physical system. Electrical analogs of the components of a system are constructed and interconnected in a highly user-oriented facility known as a parity simulator. The application of parity simulation to transient thermal-hydraulic single-phase flow is described. The development of a single-phase incompressible flow element is described. The governing mass, momentum, and energy equations along with other conditions are applied to a pipe section. The resulting model is then used to construct a circuit analog. The proposed circuit analog requires nonstandard components, the design and implementation of which is discussed. Subsequently, a formulation for single-phase compressible flow is given. Results obtained for different cases are presented. Comparison with reference numerical solutions show general agreement.

Studies on the closed-loop digital control of multi-modular reactors. Final report

This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)