Optimizing Feedwater Heater Operation in a Nuclear Power Plant

2005 ◽  
Author(s):  
Bill Fitzgerald
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Pneumatic Control ◽  
Cascading Effects ◽  
Control Scheme ◽  
Control Schemes ◽  
Operations And Maintenance ◽  
Field Device ◽  
The Individual

Feedwater heaters are a tough application even in a conventional power plant. Because of the complicated control scheme and the cascading effects between heaters, the levels in the heaters tend to cycle, reducing their ability to effectively transfer heat to the feedwater and wearing out many of the components that surround them. This situation is made even worse in a typical nuclear plant where the control schemes are normally built on obsolete local pneumatic control platforms that do not have the flexibility to be tuned for optimum performance. This paper will present the basics of a step by step analysis of a feedwater heater system in a Boiling Water Nuclear Reactor that is aimed at optimizing the individual field devices in the system and then coupling them up to a state of the art control system. It will illustrate the importance of field device selection and setup, and demonstrate several new tools available that will insure that the system is operating at peak efficiency. The summary will detail the resulting improvements in process control and their impact on reducing operations and maintenance costs, while driving up overall efficiency in the plant.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

2016 ◽  
Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Decisive Role ◽  
Reactor Power ◽  
Coupling Method ◽  
Outlet Temperature

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

scholarly journals Neutronic safety analysis of proposed reactor technologies for Ghana’s nuclear power plant using the MCNP code

2019 ◽  
Vol 34 (3) ◽  
pp. 238-242
Author(s):  
Rex Abrefah ◽  
Prince Atsu ◽  
Robert Sogbadji
Keyword(s):  
High Pressure ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Atomic Energy Commission ◽  
Clean Energy ◽  
Safety Analysis ◽  
Operating Conditions ◽  
The Government

In pursuance of sufficient, stable and clean energy to solve the ever-looming power crisis in Ghana, the Nuclear Power Institute of the Ghana Atomic Energy Commission has on the agenda to advise the government on the nuclear power to include in the country's energy mix. After consideration of several proposed nuclear reactor technologies, the Nuclear Power Institute considered a high pressure reactor or vodo-vodyanoi energetichesky reactor as the nuclear power technologies for Ghana's first nuclear power plant. As part of technology assessments, neutronic safety parameters of both reactors are investigated. The MCNP neutronic code was employed as a computational tool to analyze the reactivity temperature coefficients, moderator void coefficient, criticality and neutron behavior at various operating conditions. The high pressure reactor which is still under construction and theoretical safety analysis, showed good inherent safety features which are comparable to the already existing European pressurized reactor technology.

scholarly journals Thermodynamic analysis of high temperature nuclear reactor coupled with advanced gas turbine combined cycle

2019 ◽  
Vol 23 (Suppl. 4) ◽  
pp. 1187-1197 ◽  
Author(s):  
Marek Jaszczur ◽  
Michal Dudek ◽  
Zygmunt Kolenda
Keyword(s):  
High Temperature ◽  
Power Plant ◽  
Gas Turbine ◽  
Thermodynamic Analysis ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Coal Gasification ◽  
Nuclear Reactor ◽  
Combined Cycle ◽  
Oil Processing

One of the most advanced and most effective technology for electricity generation nowadays based on a gas turbine combined cycle. This technology uses natural gas, synthesis gas from the coal gasification or crude oil processing products as the energy carriers but at the same time, gas turbine combined cycle emits SO2, NOx, and CO2 to the environment. In this paper, a thermodynamic analysis of environmentally friendly, high temperature gas nuclear reactor system coupled with gas turbine combined cycle technology has been investigated. The analysed system is one of the most advanced concepts and allows us to produce electricity with the higher thermal efficiency than could be offered by any currently existing nuclear power plant technology. The results show that it is possible to achieve thermal efficiency higher than 50% what is not only more than could be produced by any modern nuclear plant but it is also more than could be offered by traditional (coal or lignite) power plant.

scholarly journals Determination of the safety rods (SA, SB) for optimized power reactor 1000 using the Core simulator OPR1000

2018 ◽  
Vol 1 (6) ◽  
pp. 177-184
Author(s):  
Son An Nguyen ◽  
Nguyen Trung Tran
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Boron Concentration ◽  
Power Reactor ◽  
Experimental Simulation ◽  
The Core ◽  
Operation Process

In order to operate a nuclear power plant, ensuring safety is the most important factor. The function of safety rods are to shut down the reactor in case of emergency. The purpose of this paper to show the result of research and determine the value of safety rods SA, SB. Determination of the Boron concentration corresponding to each group of safety rods of OPR1000 nuclear reactor ensures the safely in the whole operation process. Experimental simulation is carried out in the system simulating core reactor OP1R1000 (CoSi OPR1000). The expermental result corresponds with the theoretic calculated result of Sa and Sb with 1500 pcm, 4000 pcm. The concentrations of Boron appropriately are 134 ppm and 284 ppm, respectively.

scholarly journals Performance Analyses and Evaluation of Co2 and N2 as Coolants in a Recuperated Brayton Gas Turbine Cycle for a Generation IV Nuclear Reactor Power Plant

2020 ◽  
Vol 6 (2) ◽  
Author(s):  
Emmanuel O. Osigwe ◽  
Arnold Gad-Briggs ◽  
Theoklis Nikolaidis ◽  
Pericles Pilidis ◽  
Suresh Sampath
Keyword(s):  
Power Plant ◽  
Gas Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Power ◽  
Critical Conditions ◽  
Working Fluid ◽  
Closed Cycle ◽  
Gas Turbine Cycle

Abstract As demands for clean and sustainable energy renew interests in nuclear power to meet future energy demands, generation IV nuclear reactors are seen as having the potential to provide the improvements required for nuclear power generation. However, for their benefits to be fully realized, it is important to explore the performance of the reactors when coupled to different configurations of closed-cycle gas turbine power conversion systems. The configurations provide variation in performance due to different working fluids over a range of operating pressures and temperatures. The objective of this paper is to undertake analyses at the design and off-design conditions in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as the working fluid in the cycle. The analysis is demonstrated using an in-house tool, which was developed by the authors. The results show that the choice of working fluid controls the range of cycle operating pressures, temperatures, and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions. The analyses intend to aid the development of cycles for generation IV nuclear power plants (NPPs) specifically gas-cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs).

The Autoignition of Nuclear Reactor Power Plant Explosions

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactions ◽  
High Energy ◽  
Reactor Power ◽  
Common Cause ◽  
Fluid Transients ◽  
Accident Conditions

Abstract Explosive research proves that there is a common cause for most explosions in nuclear reactor power plants during normal operations and accident conditions. The autoignition of flammable hydrogen is a common cause for nuclear power plant explosions, where complex corrosion processes, nuclear reactions, and thermal-fluid transients autoignite explosions. Research evaluated increasingly complicated accidents. First, piping explosions occurred at Hamaoka and Brunsbuttel. Fluid transients compressed oxygen and flammable hydrogen to heat these gases to autoignition, where resultant explosions shredded steel pipes. This identical mechanism was responsible for pipe and pump damages to U.S. reactor systems since the 1950s, where water hammer alone has been assumed to cause damages. Small explosions inside the piping actually cause damages during nuclear reactor startups and flow rate changes. Second, explosions are caused by thermal-fluid transients during nuclear reactor restarts, following accidental nuclear reactor meltdowns. Disastrous explosions destroyed nuclear reactor buildings (RBs) at Fukushima Daiichi. Previously considered to be a fire, a 319 kilogram hydrogen explosion occurred at Three Mile Island (TMI). The explosion cause following each of these loss-of-coolant accidents was identical, i.e., after meltdowns, pump operations heated gases, which in turn acted as the heat source to autoignite sequential hydrogen explosions in reactor systems to ignite RBs. Third, the Chernobyl explosion followed a reactor meltdown that was complicated by a high energy nuclear criticality. The hydrogen ignition and explosion causes are more complicated as well, where two sequential hydrogen explosions were ignited by high-temperature reactor fuel.

Controlling and Supervising the Nuclear (Core) Power of a PWR in Relation to Nuclear Safety and Economics

2009 ◽  
Author(s):  
P. Wouters ◽  
W. Van Rompay ◽  
F. Bertels ◽  
W. Van Hove ◽  
E. Gorleer ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Total Error ◽  
Payback Period ◽  
Measuring System ◽  
Steam Generators ◽  
Nuclear Core ◽  
Ultrasonic Flow Meter

Knowing exactly the nuclear core power of a nuclear reactor is one of the most important parameters for the operator; it is vital for safety as well as for economical matters. The secondary calorimetric is the only one where one can pilot on; it is a combination of measured parameters, of which the feedwater (FW) flow towards the steam generators is the most significant one. This feedwater flow can be measured by means of an ultrasonic flow meter, “LEFM CheckPlus™ system” instead of the commonly used venturis or diaphragms. In the Belgian Nuclear Power Plant (NPP) Doel 4, a new ultrasonic “LEFM CheckPlus™” feedwater flow measuring system has been installed in April 2008. The paper describes the consequences of the installation, as the total error on the secondary calorimetric decreases from the previous 1,3% to the current 0,8% with a possibility of further reduction to 0,4%. Additionally, the economical effects of the installation are calculated for a 1000 MWe power plant with venturi meters undergoing fouling. For the NPP Doel 4 it was an economically interesting investment since the payback period was only 45 days. Finally, the possibility of consuming the margin on the secondary calorimetric for a mini-power uprate is inspected, technically and economically. It is concluded that such a mini-power uprate is an interesting option for the NPP owner.

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