Pressurized Water Reactor Power Plant Research Articles

A Computerized System for Continuous Monitoring of the Core Cooling Margin

A computerized system that assists pressurized water reactor (PWR) nuclear plant operators in controlling the saturation margin is described. This system continuously calculates and provides the temperature or pressure margin from saturation, and indicates the temperature and pressure values used for its calculation. The reactor core heatup and cooldown rates are also continuously provided by this computer-based system. The saturation curve is represented by a polynomial approximation as a function of temperature or pressure. Twenty-one analog signals of a PWR power plant are acquired by the system for calculating the margin and temperature variation rates. The man/machine interface is done through the visualization panel, which displays the above-mentioned values. Compared to the temperature and pressure values displayed in the visualization panel, the error of margin is always <0.5%.

Suboptimal Control of Pressurized Water Reactor Power Plant Using Approximate Model-Following Method

Suboptimal Control of Pressurized Water Reactor Power Plant Using Approximate Model-Following Method

Stability of Cement-Glass Packages Containing Sodium Borate Salt Generated from Pressurized Water Reactor Power Plants

A new solidification technique using cement-glass, which is a mixture of sodium silicate and cement, was studied for solidification of sodium borate salt of liquid waste generated from pressurized water reactor plants. When the sodium borate salt was solidified with the cement-glass, the resulting package contained eight times more sodium borate than was found in cement because it did not interact with sodium borate. The leaching ratio of cesium ion from the cement-glass package was one-tenth that of cement. Its low leaching ratio was due to the high cesium adsorption ability of cement-glass. The ratio could be theoretically evaluated by considering the cesium adsorption-desorption equilibrium.

Suboptimal control of pressurized water reactor power plant using approximate model-following method.

We attempted to develop an effective control system that can successfully manage the nuclear steam supply (NSS) system of a PWR power plant in an operational mode requiring relatively small variations of power. A procedure is proposed for synthesizing control system that is a simple, yet practiced, suboptimal control system. The suboptimal control system is designed in two steps; application of the optimal control theory, based on the linear state-feedback control and the use of an approximate model-following method. This procedure can appreciably reduce the complexity of the structure of the controller by accepting a slight deviation from the optimality and by the use of the output-feedback control. This eliminates the engineering difficulty caused by an incompletely state-feedback that is sometimes encountered in practical applications of the optimal state-feedback control theory to complex large-scale dynamical systems. Digital simulations and graphical studies based on the Bode-diagram demonstrate the effectiveness of the suboptimal control, and the applicability of the proposed design method as well.

Load Follow-Up Control of a Pressurized Water Reactor Power Plant by Using an Approximate Noninteractive Control

The present paper describes an attempt to apply an approximate noninteractive control to the load-following operation of the nuclear steam supply (NSS) system of a pressurized water reactor power plant. A control strategy is proposed for maximizing the unique merit of the noninteractive control in advancing the operational performance of the NSS system. An noninteractive load follow-up control system is designed based on the idea of approximate model-following. The authors make the design method more flexible and widely applicable to more general control problems by introducing some modifications to the method proposed by Cook. Digital simulations and graphical studies based on the Bode-diagram demonstrate the effectiveness of the noniniteractive load follow-up control as well as the applicability of the proposed design method.

MAXI-LILLIPUT: A simplified model of a French pressurized water reactor power plant

This paper describes a simplified model of a French pressurized water reactor (PWR) power plant. The goal of this model is the simulation of the xenon effect and his consequence on the core axial power distribution with a reactor control by means of grey rods during load-following operations and remote control (e.g. from an automatic dispatch center). We study the effects of principal design parameters on the dynamics and controlled performances of a power plant : a study of load-following capability. LILLIPUT is a good simplified model with primary and secondary parts, but the core is punctual. MAXI-LILLIPUT is an extension of LILLIPUT with an axial core representation.

A Bond Graph Description of U-Tube Steam Generator Dynamics

Bond graph methods are used to derive a nonlinear model of a U-tube steam generator like those used in pressurized water reactor (PWR) power plants. A major advantage of bond graph modeling is the ease with which different subsystem models can be interconnected; this feature is demonstrated in the steam generator modeling. Individual models of primary water cooling, generator tube heat capacity, secondary and downcomer fluid mass and energy flows, the feedwater supply system, and the main steam control valve are developed. The complete bond graph model is validated using data from a test reactor steam generator.

A Real-Time Method for Analyzing Nuclear Power Plant Transients

A method has been developed to aid pressurized water reactor (PWR) power plant operators in monitoring the transition of the plant from normal full-power conditions to expected post-reactor-trip conditions. Using a pressure-temperature plot or display, the operator can determine in real time whether the transition is normal or if a fault has occurred that has caused a loss of adequate reactor coolant system saturation margin, a loss of reactor coolant system heat sink, or excessive cooldown of the reactor coolant system. The method allows rapid differentiation among these faults and also has the capability of detecting multiple abnormalities. Evaluations of the technique have been performed with operators using manual plots in a classroom environment plotting data from plant transients and computer simulations. This method uses information that is currently available in PWR control rooms and that can be displayed on a cathode ray tube in the required format for analysis. This method appears to be very effective in communicating plant status to operators during a post-reactor-trip condition; application to other plant conditions is being investigated.

Determination of the Sources of the Airborne Physico-Chemical Iodine-131 Species in a Pressurized Water Reactor Power Plant

In a 1300-MW(electric) pressurized water reactor power plant, the sources of the airborne /sup 131/I species were determined over a period of five months. During power operation, the main source of the radiologically decisive elemental /sup 131/I was the exhaust from the hoods in which samples from the primary coolant are taken and processed. During refueling outage, elemental /sup 131/I was mainly contributed by the containment purge air. By efficient filtration of these exhausts, a reduction of the ingestion dose, caused by the total /sup 131/I stack release, by a factor of nearly 4 during power operation and of possibly 10 during refueling outage can be accomplished.

Fault Trees for Diagnosis of System Fault Conditions

Methods for generating repair checklists on the basis of fault tree logic and probabilistic importance are presented. A one-step-ahead optimization procedure, based on the concept of component criticality, minimizing the expected time to diagnose system failure is outlined. Options available to the operator of a nuclear power plant when system fault conditions occur are addressed. A low-pressure emergency core cooling injection system, a standby safeguard system of a pressurized water reactor power plant, is chosen as an example illustrating the methods presented.

加圧水形原子力プラントにおける最近の技術動向(&lt;小特集&gt;蒸気動力プラントの最近の進歩)

加圧水形原子力プラントにおける最近の技術動向(&lt;小特集&gt;蒸気動力プラントの最近の進歩)

Unitized concept for earthquake-resistant nuclear power plants

This paper discusses a unitized concept for an earthquake-resistant nuclear power plant which can withstand major earthquake shaking and fault slips without releasing radioactive material into the atmosphere. A 1000 MWe pressurized water reactor power plant of recent design is adapted to a unitized concept, and cost studies are made for the incremental cost.

Fracture behavior of reactor pressure vessel steel in the frangible, transitional and tough regimes

One of primary objectives of the USAEC's Heavy Section Stell Technology (HSST) program is to develop a technology for assessing the margins of safety against failure of the large heavy section pressure vessels typical of the commercial boiling water and pressurized water reactor power plants. It is the intent of this paper to review the several test programs which have been carried out on large specimens up to 12-in thick and to summarize how the results from such tests relate to the establishment of quantitative methods for making calculations with a practical degree of accuracy for fracture throughout the entire gamut of behaviors from frangible and transitional (mixed-mode) to tough fractures. Extensive test results have been obtained on dynamic tear specimens and compact tension specimens in thicknesses up to 12 in and at temperatures resulting in frangible through tough behaviors (that is, from considerably below room temperature up to 550°F). Large flawed 6-in thick tensile specimens have also been tested from below room temperature to over 200°F. Several series of small flawed pressure vessels have been tested and a series is scheduled for early 1972 which features vessels up to 6-in thick. Linear elastic fracture mechanics as well as elastic-plastic methods of analyzing for other than frangible behavior are reviewed and the data obtained to date are interpreted on the basis of these methods. The methods included are (1) the transition temperature approach, (2) the gross strain approach, (3) the crack opening displacement approach and (4) the equivalent energy approach. Special attention will be devoted to the equivalent energy approach since it has been used quite successfully in predicting and interpreting the fracture behavior of the thick-section tests performed under the HSST program.

Coolant instrumentation for pressurized water reactor power plants

Nuclear power has become a reality. The theory and classical experiments of half a century have been assembled and within the last decade have provided mankind with a new source of power. The problem of handling and controlling this new process is a monument in itself to technical ingenuity of the American engineer. At this stage of the art one cannot judge the nuclear power plant as a simple system, but with new technique and every day improvements this new industry will be established on an economical basis.

Main coolant instrumentation for high-performance pressurized water reactor power plants

New pressurized water reactor designs are obtaining higher efficiencies and smaller physical size by operating primary coolant loops at higher temperatures and pressures. Since the ultimate temperatures and pressures are limited by the critical point of water and instrument errors, highly accurate process-type instrumentation is required and the instrument engineer becomes an important member of the reactor design team.