scholarly journals Flowing Refractometer for Feed Water State Control in the Second Loop of Nuclear Reactor

Energies ◽  
2022 ◽  
Vol 15 (2) ◽  
pp. 457
Author(s):  
Vadim Davydov ◽  
Irena Gureeva ◽  
Roman Davydov ◽  
Valentin Dudkin
Keyword(s):  
Heat Capacity ◽  
Refractive Index ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Electrical Energy ◽  
Physical Structure ◽  
Heat Capacity Change ◽  
Feed Water ◽  
State Control ◽  
Water State

The necessity to control the feed water state in the second loop of a nuclear power plant nuclear reactor is justified. The different methods of the state control of flowing water in the pipeline are reviewed. It has been established that controlling the feed water state should not result in irreversible changes in its chemical composition and physical structure. A change in the composition or structure of feed water leads to a change in its heat capacity. The heat transfer deteriorates, the production of electrical energy in the installation decreases, and the additional release of heat into the atmosphere increases. This process also occurs during the heat capacity changes for other reasons. The method for controlling the feed water heat capacity by measuring the value of the refractive index n is developed. The design of a flow-through refractometer based on the total internal reflection for control of the feed water state in the stream is made. The dependence of the heat capacity change of feed water from the refractive index is established. The results of research on different types of water are presented.

Supercritical Water-Cooled Nuclear Reactors (SCWRs): Current and Future Concepts — Steam Cycle Options

2008 ◽  
Author(s):  
R. B. Duffey ◽  
I. Pioro ◽  
X. Zhou ◽  
U. Zirn ◽  
S. Kuran ◽  
...  
Keyword(s):  
Steam Turbine ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Hydrogen Generation ◽  
Electrical Energy ◽  
Reactor Outlet ◽  
Cycle Efficiency ◽  
Direct Cycle

One of the six Generation IV nuclear reactor concepts is a SuperCritical Water-cooled nuclear Reactor (SCWR), which is currently under development. The main objectives for developing and utilizing SCWRs are to increase the thermal efficiency of Nuclear Power Plants (NPPs), to decrease electrical energy costs, and possibility for co-generation, including hydrogen generation. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing pressure-tube SCWR concepts. The targeted steam parameters at the reactor outlet are approximately 25 MPa and 625°C. This paper presents a survey on modern SuperCritical (SC) steam turbine technology and a study on potential steam cycles for the SCWR plants. The survey reveals that by the time the Gen IV SCWRs are market-ready, the required steam turbine technology will be well proven. Three potential steam cycles in an SCWR plant are presented: a dual-cycle with steam reheat, a direct cycle with steam reheat, and a direct cycle with a Moisture Separator and Reheater (MSR). System thermal-performance simulations have been performed to determine the overall cycle efficiency of the proposed cycles. The results show that the direct cycle with steam reheat has the highest efficiency. The direct cycle with MSR is an alternative option, which will simplify the reactor design at the penalty of a slightly lower cycle efficiency.

Choices

1997 ◽  
Author(s):  
Robert Pool
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Liquid Metals ◽  
Reactor Core ◽  
Electrical Energy ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Fuel Oil ◽  
Organic Liquids ◽  
Speed Up

During the 1940s and early 1950s, when atomic energy was new, it was common to hear reactors described as nuclear “furnaces.” They “burned” their nuclear fuel and left behind nuclear “ash.” Technically, of course, none of these terms made sense, since burning is a chemical process and a reactor gets its energy from fission, but journalists liked the terminology because it was easy and quick. One loaded fuel into the reactor, flipped a switch, and things got very hot. If that wasn’t exactly a furnace, it was close enough. And actually, the metaphor was pretty good—up to a point. The basement furnace burns one of several different fuels: natural gas or fuel oil or even, in some ancient models, coal. Nuclear reactors can be built to use plutonium, natural uranium, or uranium that has been enriched to varying degrees. Home furnaces have a “coolant”—the air that is circulated through the furnace and out through the rest of the house, carrying heat away from the fire. Reactors have a coolant, too—the liquid or gas that carries heat away from the reactor core to another part of the plant, where heat energy is transformed into electrical energy. There, however, the metaphor sputters out. In a nuclear reactor, the coolant not only transfers heat to a steam generator or a turbine, but it also keeps the fuel from overheating. The coolant in a furnace does nothing of the sort. And most reactors use a moderator to speed up the fission reaction. The basement burner has nothing similar. But the most importance weakness of the furnace metaphor is that it obscured just how many varieties of reactors were possible—and, consequently, obscured the difficult choice facing the early nuclear industry: Which reactor type should become the basis for commercial nuclear power? The possibilities were practically unlimited. The fuel selection was wide. The coolant could be nearly anything that has good heat-transfer properties: air, carbon dioxide, helium, water, liquid metals, organic liquids, and so on.

Thermal-Design Options for Pressure-Channel SCWRS With Cogeneration of Hydrogen

2008 ◽  
Vol 131 (1) ◽  
Author(s):  
Maria Naidin ◽  
Sarah Mokry ◽  
Farina Baig ◽  
Yevgeniy Gospodinov ◽  
Udo Zirn ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Electrical Energy ◽  
Thermal Design ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Development Costs ◽  
Pressure Channel ◽  
Steam Cycle

Currently there are a number of Generation IV supercritical water-cooled nuclear reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are (1) to increase the gross thermal efficiency of current nuclear power plants (NPPs) from 33–35% to approximately 45–50% and (2) to decrease the capital and operational costs and, in doing so, decrease electrical-energy costs (approximately US$ 1000∕kW or even less). SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25MPa and outlet temperatures of up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical cogeneration of hydrogen through thermochemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of a SCW NPP and to increase its reliability, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature supercritical (SC) fossil power plants (including their SC turbine technology). On this basis, several conceptual steam-cycle arrangements of pressure-channel SCWRs, their corresponding T‐s diagrams and steam-cycle thermal efficiencies are presented in this paper together with major parameters of the copper-chlorine cycle for the cogeneration of hydrogen. Also, bulk-fluid temperature and thermophysical properties profiles were calculated for a nonuniform cosine axial heat-flux distribution along a generic SCWR fuel channel, for reference purposes.

Thermodynamic Considerations for a Single-Reheat Cycle SCWR

2009 ◽  
Author(s):  
M. C. Naidin ◽  
R. Monichan ◽  
U. Zirn ◽  
K. Gabriel ◽  
I. Pioro
Keyword(s):  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Electrical Energy ◽  
Performance Simulation ◽  
Development Costs ◽  
And Performance ◽  
Steam Cycle

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) Increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 30 – 35% to approximately 45 – 50%, and 2) Decrease capital and operational costs and, in doing so, decrease electrical-energy costs. SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of a SCW NPP, to increase its reliability, and to achieve similar high thermal efficiencies as the advanced fossil steam cycles it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil-fired thermal power plants (including their SC-turbine technology). The state-of-the-art SC-steam cycles at fossil-fired power plants are designed with a single-steam reheat and regenerative feedwater heating. Due to that, they reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43% on a Higher Heating Value (HHV) Basis). This paper analyzes main parameters and performance in terms of thermal efficiency of a SCW NPP concept based on a direct regenerative steam cycle. To increase the thermal efficiency and to match current SC-turbine parameters, the cycle also includes a single steam-reheat stage. The cycle is comprised of: an SCWR, a SC turbine, which consists of one High-Pressure (HP) cylinder, one Intermediate-Pressure (IP) cylinder and two Low-Pressure (LP) cylinders, one deaerator, ten feedwater heaters, and pumps. Since this option includes a “nuclear” steam-reheat stage, the SCWR is based on a pressure-tube design. A thermal-performance simulation reveals that the overall thermal efficiency is approximately 50%.

On the Use of Binary Importance Decision for Risk-Informed Categorization of SSCs

2009 ◽  
Author(s):  
Wei Zeng ◽  
Hongxing Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Water System ◽  
Injection System ◽  
Daya Bay ◽  
Feed Water ◽  
The Difference ◽  
The Relationship

The categorization of Structures, Systems and Components (SSCs) is one of the foundations in the design of the nuclear reactor. The regulation requirements based on the current deterministic categorization are over-conservative in some aspects; on the other hand, some requirements for the nonsafety-related components are too loose. In order to make the requirements reasonable, the risk-informed significance categorization of SSCs has been presented. Risk-informed significance categorization (RISC) is to categorize structures, systems, or components (SSCs) of a nuclear power plant (NPP) into two or more groups, according to their significance using both probabilistic and deterministic insights. The SSCs are quantitatively categorized by their importance measures; Fussell–Vesely (FV) and Risk Achievement Worth (RAW) are widely used. But in conventional methods for the RISC, first, a component will be categorized as significance once the value of FV or RAW is over the thresholds. So the significant components are treated equally regardless of the difference between FV and RAW, that is not suitable. Second, the component RAW derived from the sum or maximum of the basic events is not realistic. Third, the categorization threshold for FV is not uniform, different reactors have the different thresholds. The three key problems will be researched in this paper, the quartered way will be presented base on the discussion about combination of the two importance measures (binary importance decision). And then through transferring the additivity from FV to RAW, the realistic component RAW derived from FV will be got. Finally, according to the relationship between FV and RAW, the threshold for FV will be gained similar to RAW. In the study, the author will use the new method to make a practice on the Daya Bay Nuclear Power Plant, 43% components of the low pressure safety injection system which are safety-related are categorized as low significant; 2.6% components which are nonsafety-related of the auxiliary feed water system are categorized as significant.

General Layouts of Supercritical-Water NPPs

2010 ◽  
Author(s):  
I. Pioro ◽  
M. Naidin ◽  
S. Mokry ◽  
Eu. Saltanov ◽  
W. Peiman ◽  
...  
Keyword(s):  
Supercritical Water ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Electrical Energy ◽  
Development Costs ◽  
Flow Circuit ◽  
Steam Cycle

Currently, there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) Increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 30–35% to approximately 45–50%, and 2) Decrease capital and operational costs and, in doing so, decrease electrical-energy costs. SuperCritical Water (SCW) NPPs will have much higher operating parameters compared to current NPPs (i.e., steam pressures of about 25 MPa and steam outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of an SCW NPP, to increase its reliability, and to achieve similar high thermal efficiencies as the advanced fossil-fired steam cycles, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil-fired thermal power plants (including their SC-turbine technology). The state-of-the-art SC-steam cycles at fossil-fired power plants are designed with a single-steam reheat and regenerative feedwater heating. Due to this, they reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43–50% on a Higher Heating Value (HHV) basis). This paper presents several possible general layouts of SCW NPPs, which are based on a regenerative-steam cycle. To increase the thermal efficiency and to match current SC-turbine parameters, the cycle also includes a single steam-reheat stage. Since these options include a nuclear steam-reheat stage, the SCWR is based on a pressure-tube design.

scholarly journals SIMULATION OF THERMAL PROCESSES OF A NPP STEAM GENERETOR FOR INFORMATION TECHNOLOGY OPTIMIZED CONTROL

2021 ◽  
pp. 56-61
Author(s):  
Olena Nikulina ◽  
Valerii Severyn ◽  
Nina Kotsiuba ◽  
Anton Bubnov
Keyword(s):  
Heat Transfer ◽  
Information Technology ◽  
Differential Equations ◽  
Steam Generator ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Feed Water ◽  
Thermal Processes ◽  
State Variables ◽  
System Of Differential Equations

Mathematical models of thermal processes in the form of Cauchy in the state space with relative variables of the steam generator PGV-1000 of the power unit of a nuclear power plant with a nuclear reactor VVER-1000 have been developed for the using of models in information technology for optimizing the control of a steam generator. The working thermal processes in the PGV-1000 steam generator associated with the supply of feed water to it from the water treatment system and the coolant from the nuclear reactor and the removal of vapors to the main steam header are considered. The design diagram of the steam generator is presented, which reflects the working processes in it under the evaporation mirror and above it. On the basis of differential equations of the heat balance of the heat carrier in the steam generator and in the metal heat exchange tubes, the simulation of heat transfer from the heat carrier to the feed water in the steam generator is carried out. The heat transfer model in the form of a linear system of differential equations in relative state variables is developed. The processes of vaporization during heating of feed water by the heat transfer surface are considered. Differential equations of material and heat balances of dynamic processes of vaporization in a steam generator are compiled, which are not equations in the Cauchy form. Transformations of the differential equations of material and heat balances in the steam generator to the Cauchy form are carried out. A nonlinear system of differential equations for the balance of vaporization in relative state variables is obtained. The values of the constant parameters of the models for the steam generator PGV-1000 are calculated. The mathematical model of thermal processes in the PGV-1000 steam generator, which is presented in the form of a system of differential equations and includes the processes of heat transfer and steam generation, will make it possible to identify and optimize the steam generator control system with the help of information optimization technology. Keywords: nuclear power plant, steam generator, thermal processes, mathematical model, differential equations, optimization, control, information technology.

Conceptual Thermal-Design Options for Pressure-Tube SCWRs With Thermochemical Co-Generation of Hydrogen

2008 ◽  
Author(s):  
Sarah Mokry ◽  
Maria Naidin ◽  
Farina Baig ◽  
Yevgeniy Gospodinov ◽  
Udo Zirn ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Electrical Energy ◽  
Thermal Design ◽  
Pressure Tube ◽  
Fluid Temperature ◽  
Bulk Fluid ◽  
Development Costs ◽  
Steam Cycle

Currently there are a number of Generation IV SuperCritical Water-cooled nuclear Reactor (SCWR) concepts under development worldwide. The main objectives for developing and utilizing SCWRs are: 1) To increase gross thermal efficiency of current Nuclear Power Plants (NPPs) from 33–35% to approximately 45–50%, and 2) To decrease the capital and operational costs and, in doing so, decrease electrical-energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to current NPPs (i.e., pressures of about 25 MPa and outlet temperatures up to 625°C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated. Furthermore, SCWRs operating at higher temperatures can facilitate an economical co-generation of hydrogen through thermo-chemical cycles (particularly, the copper-chlorine cycle) or direct high-temperature electrolysis. To decrease significantly the development costs of a SCW NPP and to increase its reliability, it should be determined whether SCW NPPs can be designed with a steam-cycle arrangement that closely matches that of mature SuperCritical (SC) fossil power plants (including their SC turbine technology). The state-of-the-art SC steam cycles in fossil power plants are designed with a single-steam reheat and regenerative feedwater heating and reach thermal steam-cycle efficiencies up to 54% (i.e., net plant efficiencies of up to 43% on a Higher Heating Value Basis). It would be beneficial if SCWRs could involve a regenerative feedwater heating and nuclear steam reheat to be able to adapt the current SC turbine technology and to achieve similar high thermal efficiencies as the advanced fossil steam cycles. The nuclear steam reheat is easier to implement inside pressure-tube or pressure-channel reactors compared to pressure-vessel reactors. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing concepts of the pressure-tube SCWRs. Therefore, no-reheat, single-reheat, and double-reheat cycles of future SCW NPPs were analyzed in terms of their thermal efficiencies. On this basis, several conceptual steam-cycle arrangements of pressure-tube SCWRs, their corresponding T-s diagrams and steam-cycle thermal efficiencies are presented in this paper together with major parameters of the copper-chlorine cycle for the co-generation of hydrogen. Also, bulk-fluid temperature and thermophysical properties profiles were calculated for a non-uniform cosine Axial Heat-Flux Distribution (AHFD) along a generic SCWR fuel channel, for reference purposes.

scholarly journals Temperature Dependence of Thermophysical Properties of Full-Scale Corium of Fast Energy Reactor

2017 ◽  
Vol 2017 ◽  
pp. 1-7
Author(s):  
Mazhyn K. Skakov ◽  
Nurzhan Ye. Mukhamedov ◽  
Alexander D. Vurim ◽  
Ilya I. Deryavko
Keyword(s):  
Heat Capacity ◽  
Thermal Diffusivity ◽  
Fast Reactor ◽  
Thermophysical Properties ◽  
Heat Conductivity ◽  
Nuclear Reactor ◽  
Reactor Vessel ◽  
Full Scale ◽  
Temperature Fields ◽  
First Time

For the first time the paper determines thermophysical properties (specific heat capacity, thermal diffusivity, and heat conductivity) of the full-scale corium of the fast energy nuclear reactor within the temperature range from ~30°С to ~400°С. Obtained data are to be used in temperature fields calculations during modeling the processes of corium melt retention inside of the fast reactor vessel.

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