Pressurized Water Reactor Power Plant Research Articles

Predicting the wall thinning engendered by erosion–corrosion using CFD methodology

Erosion–corrosion (EC) is a serious degradation mechanism of piping, especially for nuclear power plants since it may result in the piping damage, plant shutdown, or personnel injury. The majority of this paper investigates the dependence of wall thinning on the hydrodynamic characteristics using the computational fluid dynamics (CFD) methodology. Four piping systems in a pressurized water reactor (PWR) power plant are selected in this investigation. Based on the plots showing the measured wall thinning with the calculated hydrodynamic parameters, the relationship between them is clearly revealed. Utilizing the characteristics of near-wall turbulence kinetic energy, an envelopment model is proposed herein to conservatively predict the amount of wall thinning distributed on the pipe wall. This estimation model can simply predict the possible distributions of severe EC wear sites and subsequently assist the plant staff to schedule the pipe wall monitoring program in the measured range of pipe wall for the fittings.

Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

Thermoeconomic optimization of a hybrid pressurized water reactor (PWR) power plant coupled to a multi effect distillation desalination system with thermo-vapor compressor (MED-TVC)

Thermoeconomic optimization of a typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant coupled to a Multi Effect Distillation (MED) desalination system with thermo-vapor compressor (TVC) is performed. A thermodynamic modeling based on the energy and exergy analysis is performed while economic modeling is developed based on the Total Revenue Requirement (TRR) method. The objective function based on the thermoeconomic analysis is obtained. The proposed cogeneration plant, for simultaneous production of power and fresh water, including sixteen decision variables is proposed for thermoeconomic optimization in which the goal is minimizing the cost of system product (including the cost of generated electricity and fresh water). The optimization process is performed using a stochastic/deterministic optimization approach namely as Genetic Algorithm. It is found that thermoeconomic optimization aims at reduction of sub-components total costs by reducing either the cost of inefficiency or the cost of owning the components, whichever is dominant. For some components such as evaporators, the improvement is obtained by reducing the owning cost of the sub-system at the cost of reduction of the thermodynamic efficiency. For components like as TVC + de-superheater, improvement is achieved by increasing the thermodynamic efficiency or decreasing the inefficiency cost.

Multi-objective thermoeconomic optimization of coupling MSF desalination with PWR nuclear power plant through evolutionary algorithms

This paper deals with the application of an evolutionary algorithm to multi-objective thermoeconomic optimization of coupling a multi stage flash desalination (MSF) plant with a pressurized water reactor (PWR) nuclear power plant. The thermodynamic simulation of this initial PWR plant has been performed in a Thermoflex simulator. An Excel add-in called Thermoflex Link has been developed to calculate the exergy of each stream from Thermoflex simulation results. Meanwhile, a computer code has been developed for thermoeconomic and improved combined pinch–exergy analysis in Matlab environment. Both the design configuration (feed water heater structure) and the process variables are optimized simultaneously. The optimization algorithm can choose among several design options included in a superstructure of the feed water heater in dual purpose plant. For the assumptions and simplifications made in this study, a 3000 MW (thermal) PWR power plant such as the Bushehr power plant has been considered. A detailed exergy and exergoeconomic analysis of selected final optimal design identifies the magnitude, location and causes of the thermodynamic inefficiencies. Improved combined pinch and exergy analysis has been applied to display the system information graphically for one to visualize the performance of the system in the initial and final case.

Aplicação de um modelo fuzzy DEA para priorizar modos de falha em sistemas nucleares

Neste artigo apresenta-se a aplicação de um modelo híbrido, baseado na análise envoltória de dados e em conjuntos fuzzy, para estabelecer uma priorização dos modos de falha levantados numa análise dos modos e efeitos de falha. O objetivo principal é ilustrar a aplicabilidade da modelagem para analisar sistemas complexos, como o sistema de controle químico e volumétrico de uma usina nuclear do tipo água pressurizada, onde uma grande quantidade de modos de falha deve ser avaliada. O modelo apresentado possibilita tratar variáveis linguísticas, i.e., permite que se tratem as opiniões de especialistas de modo mais efetivo num contexto de análise de riscos industriais quando há poucos dados disponíveis, como pode ocorrer em projetos com inovação tecnológica. Os resultados obtidos demonstram o potencial e a viabilidade de se utilizar uma modelagem híbrida para se estabelecer uma ordem de prioridade entre os diferentes modos de falha associados a um sistema complexo.

Control rod drop analysis by finite element method using fluid–structure interaction for a pressurized water reactor power plant

The control rod drop analysis is very important for safety analysis. For seismic and loss of coolant accident event, the control rod assemblies shall be capable of traveling from a fully withdrawn position to 90% insertion without any blockage and within specified time and displacement limits. The analysis has been executed by analytical method using in-house code. In this method, several field data are needed. These data are obtained from nuclear, thermal–hydraulic and mechanical design groups, peculiar codes, those work groups need to cooperate together. Following the enhancement of a computer and development of the multi-physics analysis code, a new method for the control rod drop analysis is proposed by finite element method. This analysis model incorporates the structure and fluid parts, termed as a fluid and structure interaction (FSI). Because a control rod is submerged inside a guide tube of a fuel assembly, the FSI boundary condition is applied. In this model, it is assumed that the fluid is incompressible laminar flow. The structures are modeled with the solid elements because there is no deformation due to the fluid flow. The analysis two-dimensional plane model is created in the analysis with considering an axi-symmetric geometry. Therefore, the proposed analysis model will be very simple and the design data from other fields will be unnecessary. The analysis results are compared with those of the in-house code, which have been used for a commercial design. After validation, it is found that the present analysis gives a useful tool in the design of the control rod and fuel assembly.

E213 PLANT PERFORMANCE AND ECONOMIC STUDY ON OXY FUEL GAS TURBINE POWER PLANT UTILIZING NUCLEAR STEAM GENERATOR(Power System-3)

In this paper we propose a new plant of a closed cycle oxy-fuel gas turbine power plant combining a nuclear heat generator. A pressurized water reactor (PWR) is designed to supply saturated steam into an oxy-fuel gas turbine power plant for specific power output increase. The saturated steam from the reactor can be under lower pressure and temperature than those of an existing PWR power plant. In this study we estimated plant performance from the heat balance model based on a conceptual design of a hybrid plant and generating costs of the proposed plant calculated from Japanese cost data of an existing PWR plant and an LNG combined cycle gas turbine plant. The generating efficiency of an oxy-fuel gas turbine plant without nuclear steam generator is estimated to be less than 35 [%]. Based on this efficiency of the whole plant performance, the corresponding generating efficiency of nuclear steam generator contributing to the specific power output of the proposed hybrid plant is estimated to be around 45 [%] even if the steam conditions are lower than an existing PWR plant. The generating costs of the hybrid plant are estimated for some different power capacities from 200 to 1000 [MWe] with gas turbine combustor inlet pressure of 5.00 [MPa]. The generating costs are 15-20 [%] lower than those calculated from the weighted heat performance of both an oxy-fuel gas turbine plant without nuclear steam generator and an existing PWR plant.

Optimization of the coupling of pressurized water nuclear reactors and multistage flash desalination plant by evolutionary algorithms and thermoeconomic method

Thermodynamic simulation programs are widely used for designing complex thermal systems, but most of them do not incorporate second law optimization techniques. In this study, an efficient optimization strategy is presented, which integrates three optimization techniques with a professional power plant and a cogeneration simulator so as to perform exergoeconomic optimization of complex thermal systems and generate combined pinch and exergy representations. This paper deals with the application of an evolutionary algorithm based on NSGA-II to multi-objective thermoeconomic optimization of coupling desalination plant with pressurized water reactor (PWR). In addition, one-objective thermoeconomic optimization through genetic algorithm and mixed integer non-linear mathematical programming methods has been applied for evaluation of multi-objective optimization. The thermodynamic simulation of this plant has been performed in the THERMOFLEX simulator. An Excel Add-in called THERMOFLEX link has been developed to calculate the exergy of each stream from THERMOFLEX simulation results. In addition, a computer code has been developed for thermoeconomic and improved combined pinch–exergy analysis in the MATLAB environment. Also, multi-objective and one-objective evolutionary algorithm optimization has been performed in MATLAB and one-objective mathematical programming has been performed in LINGO software. Both the design configuration and the process variables are optimized simultaneously. The optimization algorithm can choose among several design options included in a superstructure of the feed water heaters and multistage flash desalination in a dual-purpose plant. For the assumptions and simplifications made in this study, a 3000 MWh PWR power plant similar to Bushehr power plant has been considered. Copyright © 2008 John Wiley & Sons, Ltd.

Various approaches in optimization of a typical pressurized water reactor power plant

A typical 1000 MW pressurized water reactor ( PWR) nuclear power plant is considered for optimization. The thermodynamic modeling is performed based on the energy and exergy analysis, while an economic model is developed according to the total revenue requirement ( TRR) method. The objective functions based on the thermodynamic and thermoeconomic analysis are obtained. The optimization process with ten degrees of freedom for plant under consideration is performed using the genetic algorithm ( GA). Three levels of optimization including thermodynamic single-objective, thermoeconomic single-objective and multi-objective optimizations are performed. In multi-objective optimization, both thermodynamics and thermoeconomic objectives are considered, simultaneously. A series of optimum solutions namely as Pareto frontier are obtained. In the case of multi-objective optimization, an example of decision-making process for selection of the final optimal solution from the Pareto frontier is introduced. The results obtained using the various optimization approaches are compared and discussed.

Selection of a steam generator pressure control program for a 500 MWe pressurized heavy water reactor power plant through transient analysis

A personal computer based 500 MWe pressurized heavy water reactor (PHWR) power plant transient analysis code has been developed for analysing plant operational transients, for sorting out design issues and for taking decisions on new operating procedures and safety-related transient events. This code is based on the coupled solution of the conservation equations of mass, momentum and energy and the equation of the fluid state for simulation of process transients/dynamics. It includes mathematical models for all the major equipment/subsystems, i.e. the reactor core heat transfer, steam generators (SGs), surge tank (pressurizer), bleed condenser, etc., of the 500 MWe PHWR power plant. All the associated control systems on the primary and secondary sides are also incorporated. These codes are also validated against the available plant data. This code was also used to resolve the issues related to SG pressure control program selection. Two fundamental types of SG pressure control program are available for nuclear power plants, i.e. the constant boiler pressure program (CBPP) and the variable boiler pressure program (VBPP) for power manoeuvring from zero to 100 per cent of full power (FP), and vice versa. With CBPP the secondary steam cycle design and operation are simplified, at the cost of more load on the primary heat transport (PHT) pressure control system, whereas VBPP reduces the load on the PHT pressure controller. Parametric studies were carried out by taking into account the reactor trip, turbine trip and the IRV sizing design basis transient for comparison of these SG pressure control programs, i.e. (1) the earlier old design VBPP with constant PHT temperature 25 per cent of FP onwards, (2) CBPP and (3) VBPP similar to the 220 MWe PHWR SGPC program. After examining the two extremes in cases 1 and 2, the third scheme similar to the 220 MWe boiler pressure control (BPC) program was found to be most suitable. Schemes 2 and 3 put more load on the PHT pressure control and make use of the capacities enhanced by the pressurizer. The dip in the pressurizer level is acceptable in the CBPP, i.e. in the second case also. All the pros and cons of the selection of a BPC program were clearly quantified and resolved by considering a wide range of transients. All the related transient studies are discussed in this paper.

A PSA-based vital area identification methodology development

Abstract This paper presents a vital area identification method based on the current probabilistic safety assessment (PSA) techniques. The vital area identification method in this paper is focused on core melt rather than radioactive material release. Furthermore, it describes a conceptual framework with which the risk from sabotage-induced events could be assessed. Location minimal cut sets (MCSs) are evaluated after developing a core melt location fault tree (LFT). LFT is a fault tree whose basic events are sabotage-induced damages on the locations within which various safety-related components are located. The core melt LFT is constructed by combining all sequence LFTs of various event trees with OR gates. Each sequence LFT is constructed by combining the initiating event LFT and the mitigating event LFTs with an AND gate. The vital area could be identified by using the location importance measures on the core melt location MCSs. An application was made to a typical 1000 MWe pressurized water reactor power plant located at the Korean seashore. The methodology suggested in the present paper is believed to be very consistent and most complete in identifying the vital areas in a nuclear power plant because it is based on the well-proven PSA technology.

Optimal power control system of a research nuclear reactor

The power control system is a key control system for a nuclear reactor, which directly concerns the safe operation of a nuclear reactor. Much attention is paid to the power control system performance of nuclear reactor in engineering. The designers put a high value upon design of an optimal power control system. In this paper, a design method is applied to the design of power control system. According to the optimal control theory, an objective function, quadratic performance index with weight factors is proposed. Then, the objective function is transformed into frequency domain form by use of Paserval's theorem. In frequency domain, an optimal transfer function can be obtained at the lowest value of objective function. The system with optimal transfer function has an optimal performance. The transfer function of the power control system is derived from a typical research nuclear reactor. Using the state feedback theory, the transfer function is synthesized to the optimal transfer function. The simulative results with the optimal controller and with a conventional controller show that the performance of the optimal power control system is largely improved on dynamic characters. The method applied here not only can be used for research nuclear reactor but also can be easily extended to pressurized water reactor power plant and other fields.

A Pressurized Water Reactor Design for Plutonium Incineration: Fuel Cycle Options

The possibility exists of utilizing pressurized water reactor (PWR) power plants of current technology for efficient and cost-competitive incineration of excess plutonium. Several plutonium-based fuel cycle options were considered, i.e., pure 239Pu or reactor-grade plutonium as a fissile component and natural uranium or thorium as a fertile component of the fuel.A typical PWR was chosen as the base for detailed analysis and comparison of all investigated fuel cycle options. A series of calculations was carried out for each of the fuel cycle options generating “equilibrium” cycles of equal length.Results of the design analysis and comparison of main performance parameters were used to compare different fuel options. Material mass balances were calculated to evaluate the plutonium incineration potential of the considered options. A potential of efficient reduction of excess plutonium was demonstrated for all considered fuel options. The thorium-based fuel cycles were found especially effective for destruction of fissile isotopes of plutonium (>1000 kg/yr). This was partially compensated by the buildup of 233U isotope. One of the important conclusions of this work is that significant amounts of fissile plutonium may be incinerated in thorium-based cycles and that 233U may be denatured by addition of modest amounts of natural uranium. Preliminary economic evaluations indicate that plutonium incineration may be carried out in existing PWRs without economic penalty and may, therefore, present a viable alternative to other plutonium disposition methods.

Pump and Valve Fastener Serviceability in PWR Nuclear Facilities

The results of several studies conducted on corrosion of carbon and low-alloy steels in borated water have shown that impingement of borated steam on ferritic steels or contact with a moist paste of boric acid can lead to high corrosion rates due to high local concentrations of boric acid on the surface. The corrosion process of the flange fasteners of pumps and valves is considered a material compatibility and equipment maintenance problem. Therefore, the nuclear utilities of pressurized water reactor (PWR) power plants can prevent this damage by implementing appropriate fastener steel replacement and extended inspections to detect and correct the cause of leakage. A 3-phase corrosion protection program is presented for implementation based on system operability, outage-related accessibility, and cost of fastener replacement versus maintenance frequency increase. A selection criterion for fastener material is indicated based on service limitation: preloading and metal temperature.

Fault-tolerant control and diagnostics for large-scale systems

The integration of control design, signal and command validations, and diagnostics is the primary objective of this study. The fault-tolerant control principle consists of parallel control and command validation functions. The parallel control includes a number of control algorithms, each designed to accomplish the same control task using different sets of plant signals. The purpose of command validation is to predict the expected control actions using independent approaches, and to cross-check the control actions. The fault-tolerant control and diagnostics system (FCDS) was demonstrated with application to the regulation of the feedwater system in a four-loop pressurized water reactor (PWR) power plant. Because of a high level of reliability that can be achieved by this tier system for control design approach, this is referred to as a "smart" controller.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Concentration of Radiocarbon and Its Chemical Forms in Gaseous Effluents, Environmental Air, Nuclear Waste and Primary Water of a Pressurized Water Reactor Power Plant in Hungary

We measured airborne releases of 14C from the Paks Pressurized Water Reactor (PWR) Nuclear Power Plant (NPP). Two continuous stack samplers collect 14C in 14CO2 and 14CnHm chemical forms. 14C activities were measured using two techniques; environmental air samples of lower activities were analyzed by proportional counting, stack samples were measured by liquid scintillation counting. 14C concentration of air in the stack varies between 80 and 200 Bqm−3. The average normalized yearly discharge rates for 1988–1993 were 0.74 TBqGW−1ey−1 for hydrocarbons and 0.06 TBqGW−1ey−1 for CO2. The discharge rate from Paks Nuclear Power Plant is about four times higher than the mean discharge value of a typical Western European PWR NPP. The higher 14C production may be apportioned to the higher level of nitrogen impurities in the primary coolant. Monitoring the long-term average excess from the NPP gave D14C = 3.5‰ for CO2 and D14C = 20‰ for hydrocarbons. We determined 14C activity concentration in the primary coolant to be ca. 4 kBq liter−1. The 14C activity concentrations of spent mixed bed ion exchange resins vary between 1.2 and 5.3 MBqkg−1 dry weight.

Examples of the evolution of materials for nuclear applications: metallurgical improvement of 16MND5 steel and new technologies for manufacturing heavy components

Creusot-Loire Industrie has been constantly improving and adjusting the quality of its products (heavy forgings, castings or plates) based on knowledge of their in-service behaviour. The development of steel manufacturing techniques has been oriented to increasing the quality of steel, in particular by reducing drastically the impurity content (S, P), by better control of the basic range of chemical composition, and by decreasing the gas content (N, H, O). Changes in ingot shapes have been made to optimize the quality of the steel with respect to segregation (development of hollow and LSD ingots to reduce the effect of irradiation embrittlement at the inner surface and sub-surface area). The effects of these developments on the toughness of nuclear materials are illustrated by two examples: 16MND5 steel used for pressurized water reactors; ASTM A350LF5 steel used for cask bodies for the transport of irradiated fuel elements. Modifications in hot working (mainly forging) procedures have enabled the development of products with shape adapted to easier construction and/or best quality and/or reduction in the number of welds, thus decreasing the construction and in-service inspection costs. New forging products to be used in nuclear pressurized water reactor power plants are presented.

An experimental study on drift flux parameters for two-phase flow in vertical round tubes

Correlations which can correctly express the velocity difference between steam and water of two-phase flow are required for accurate predictions of the thermal-hydraulic transient behaviour for a pressurized water reactor (PWR) power plant during a small break loss-of-coolant accident (LOCA). A drift flux model is one of the typical models which describes the relative velocity between the two phases. A number of empirical expressions of the parameters for the drift flux model have been proposed based on experiments. It is necessary to confirm whether those formulas are applicable to the simulation of two-phase flow behaviour in a large scale apparatus such as the actual PWR plant. In this study, the experiments of cocurrent and counter-current steam-water two-phase flow using a large scale apparatus with 0.1023 m diameter tubes, and small scale apparatus with 0.0197 m diameter tubes were carried out. This paper presents the tube diameter effects on the drift flux parameters ( C o = distribution parameter and V gj = drift velocity). In addition, new empirical correlations of the drift flux parameters for the two-phase flow in large diameter tubes are proposed. A root mean square error of the predicted void fraction is 16.8% compared with 1353 experimental data.

Analytical technique of lighting surges induced on grounding mesh of PWR nuclear power plant

It is noted that an analytical technique is needed to make a qualitative and predictive evaluation of transient voltages induced on local grounding meshes and instrumentation cables by a lightning strike on a lightning rod in a PWR (pressurized water reactor) power plant. An experiment with lightning surge impulses in a PWR power plant was set up to observe lightning-caused transient voltages. Experimental data are compared with EMTP (electromagnetic transient program) simulation results, and an improved simulation method is developed. The improved method provides a good estimation of induced voltages on grounding meshes and instrumentation cables. An accuracy factor of 3-5 was reached. >

Assessment of Long-Term Reliability of Pressurized Water Reactor Plants Based on Power Generation Statistics (Up to 1987)

The performance of pressurized water reactor (PWR) power plants is evaluated through analysis of power generation records, and directions are given in which measures might be effectively sought for further improvement of plant performance and productivity. The boundary between what has already been achieved in the performance of such plants and what remains to be done in development and demonstration is clearly identified.To supplement the traditionally adopted “capacity factor” (average power/nominal capacity) for assessing the improvements gained in the performance of uprated fuels and reactor cores, additional new yardsticks are proposed to represent the productivity of nuclear fuel and the reactor core. For evaluating the performance of fuel, the proposed variable is based on the correlation between specific power and annual core average burnup, i.e., thermal power and annual heat generation per unit mass of fuel. Similarly for the reactor core, the variable is based on the correlation between thermal power and heat generation per unit core volume and per unit area of core cross section. The advantages of adopting the proposed variables are discussed.Operating experience with PWR plants indicates that the relatively short service life of nuclear fuel, compared with the core structure and other reactor components, has permitted reliable, effective service of uprated fuel in high-performance plants to be demonstrated over periods extending beyond the service life of individual fuels. This is not the case, however, with the core structure and other reactor components: Most existing plants have been in service for less than half of the expected service life of these components, therefore data available today are insufficient for evaluation of their long-term performance. An analysis is presented on possible repercussions to be expected from the current trend in development, which tends toward higher core power rating, and it is pointed out that certain plant components will possibly come to be exposed to increasingly severe thermal conditions, which calls for further efforts in development and demonstration to ensure their continued reliability in service.