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.

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.

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%.

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.

Thermal Design Options Using Uranium Carbide and Uranium Dicarbide in SCWR Uniformly-Heated Fuel Channel

2009 ◽  
Author(s):  
Leyland J. Allison ◽  
Lisa Grande ◽  
Sally Mikhael ◽  
Adrianexy Rodriguez Prado ◽  
Bryan Villamere ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Operating Conditions ◽  
Thermal Design ◽  
Nuclear Fuels ◽  
Viable Option ◽  
Bulk Fluid ◽  
Fuel Channel ◽  
Uranium Carbide

SuperCritical Water-cooled nuclear Reactor (SCWR) options are one of the six reactor options identified in Generation IV International Forum (GIF). In these reactors the light-water coolant is pressurized to supercritical pressures (up to approximately 25 MPa). This allows the coolant to remain as a single-phase fluid even under supercritical temperatures (up to approximately 625°C). SCW Nuclear Power Plants (NPPs) are of such great interest, because their operating conditions allow for a significant increase in thermal efficiency when compared to that of modern conventional water-cooled NPPs. Direct-cycle SCW NPPs do not require the use of steam generators, steam dryers, etc. allowing for a simplified NPP design. This paper shows that new nuclear fuels such as Uranium Carbide (UC) and Uranium Dicarbide (UC2) are viable option for the SCWRs. It is believed they have great potential due to their higher thermal conductivity and corresponding to that lower fuel centerline temperature compared to those of conventional nuclear fuels such as uranium dioxide, thoria and MOX. Two conditions that must be met are: 1) keep the fuel centreline temperature below 1850°C (industry accepted limit), and 2) keep the sheath temperature below 850°C (design limit). These conditions ensure that SCWRs will operate efficiently and safely. It has been determined that Inconel-600 is a viable option for a sheath material. A generic SCWR fuel channel was considered with a 43-element bundle. Therefore, bulk-fluid, sheath and fuel centreline and HTC profiles were calculated along the heated length of a fuel channel.

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.

Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat

2016 ◽  
Author(s):  
Yifeng Zhou ◽  
Paul Ponomaryov ◽  
Cristina Mazza ◽  
Igor Pioro
Keyword(s):  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Thermal Power ◽  
Difficult Problem ◽  
Combined Cycle ◽  
The Usa ◽  
Pressure Channel ◽  
High Neutron

Currently, i.e., in 2016, 4361 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%). Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of nuclear-power reactors. It should be noted that among six Generation-IV nuclear-reactor concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature. As an interim way of thermal-efficiency improvement for water-cooled NPPs nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat in subcritical-pressure experimental boiling reactors. Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel reactors with nuclear steam reheat. For this purpose the DE-TOP program has been used.

Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)

2014 ◽  
Author(s):  
Marija Miletic ◽  
Wargha Peiman ◽  
Amjad Farah ◽  
Jeffrey Samuel ◽  
Alexey Dragunov
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Electrical Energy ◽  
Rankine Cycle ◽  
Combined Cycle ◽  
Accurate Estimation ◽  
Pressure Channel

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical-energy generation. The largest group of operating Nuclear Power Plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) have gross thermal efficiencies ranging from 30% and up to 36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8 MPa / 257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fuel – natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., SuperCritical Water-cooled Reactors (SCWRs) have to be designed. This path of the thermal-efficiency increasing is considered as a conventional way through which coal-fired power plants gone more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel Pressure-Channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulic code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the CFD Fluent code has been used for better understanding of specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical-water loop and developing passive-safety systems.

Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor

2015 ◽  
Vol 1 (1) ◽  
Author(s):  
Marija Miletic ◽  
Wargha Peiman ◽  
Amjad Farah ◽  
Jeffrey Samuel ◽  
Alexey Dragunov ◽  
...  
Keyword(s):  
Heat Transfer ◽  
Supercritical Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Electrical Energy ◽  
Rankine Cycle ◽  
Combined Cycle ◽  
Accurate Estimation ◽  
Pressure Channel

Nuclear power becomes more and more important in many countries worldwide as a basis for current and future electrical energy generation. The largest group of operating nuclear power plants (NPPs) equipped with water-cooled reactors (96% of all NPPs) has gross thermal efficiencies ranging from 30–36%. Such relatively low values of thermal efficiencies are due to lower pressures/temperatures at the inlet to a turbine (4.5–7.8  MPa/257–293°C). However, modern combined-cycle power plants (Brayton gas-turbine cycle and subcritical-pressure steam Rankine cycle, fueled by natural gas) and supercritical-pressure coal-fired power plants have reached gross thermal efficiencies of 62% and 55%, respectively. Therefore, next generation or Generation IV NPPs with water-cooled reactors should have thermal efficiencies as close as possible to those of modern thermal power plants. A significant increase in thermal efficiencies of water-cooled NPPs can be possible only due to increasing turbine inlet parameters above the critical point of water, i.e., supercritical water-cooled reactors (SCWRs) have to be designed. This path of increasing thermal efficiency is considered as a conventional way that coal-fired power plants followed more than 50 years ago. Therefore, an objective of the current paper is a study on neutronics and thermalhydraulics characteristics of a generic 1200-MWel pressure-channel (PCh) SCWR. Standard neutronics codes DRAGON and DONJON have been coupled with a new thermalhydraulics code developed based on the latest empirical heat-transfer correlation, which allowed for more accurate estimation of basic characteristics of a PCh SCWR. In addition, the computational fluid dynamics (CFD) Fluent code has been used for better understanding of the specifics of heat transfer in supercritical water. Future studies will be dedicated to materials and fuels testing in an in-pile supercritical water loop and developing passive safety systems.

Power Cycles of Generation III and III+ Nuclear Power Plants

2014 ◽  
Author(s):  
Alexey Dragunov ◽  
Eugene Saltanov ◽  
Igor Pioro ◽  
Pavel Kirillov ◽  
Romney Duffey
Keyword(s):  
Renewable Energy ◽  
Natural Gas ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Renewable Energy Sources ◽  
Thermal Power ◽  
Electrical Energy ◽  
Energy Sources ◽  
Thermal Power Plants

It is well known that the electrical-power generation is the key factor for advances in any other industries, agriculture and level of living. In general, electrical energy can be generated by: 1) non-renewable-energy sources such as coal, natural gas, oil, and nuclear; and 2) renewable-energy sources such as hydro, wind, solar, biomass, geothermal and marine. However, the main sources for electrical-energy generation are: 1) thermal - primary coal and secondary natural gas; 2) “large” hydro and 3) nuclear. The rest of the energy sources might have visible impact just in some countries. Modern advanced thermal power plants have reached very high thermal efficiencies (55–62%). In spite of that they are still the largest emitters of carbon dioxide into atmosphere. Due to that, reliable non-fossil-fuel energy generation, such as nuclear power, becomes more and more attractive. However, current Nuclear Power Plants (NPPs) are way behind by thermal efficiency (30–42%) compared to that of advanced thermal power plants. Therefore, it is important to consider various ways to enhance thermal efficiency of NPPs. The paper presents comparison of thermodynamic cycles and layouts of modern NPPs and discusses ways to improve their thermal efficiencies.

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