Nuclear Power Plants With Closed-Cycle Helium Turbine for Industrial Energy Supply

1971 ◽  
Vol 93 (1) ◽  
pp. 156-161 ◽  
Author(s):  
K. Bammert ◽  
E. Boehm ◽  
R. Buende
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Expansion Ratio ◽  
Inlet Temperature ◽  
Closed Cycle ◽  
Steam Generation ◽  
Regenerative Heat ◽  
Cycle Number ◽  
Industrial Energy ◽  
Gas Turbine Cycle

All the possible arrangements of closed-cycle helium turbines for combined power and steam generation can be reduced to three variations of one single arrangement. A method for the optimum layout of this arrangement is described. Optimum data are determined for the expansion ratio and the regenerative heat exchange as functions of general parameters—electric power output and quantity and quality of the steam produced—and as functions of special parameters of the gas turbine cycle—number of intercoolers, cooling and feedwater temperatures, turbine inlet temperature and pressure drop—to obtain the best possible efficiencies.

Comparison of Nuclear Power Plants With Closed-Cycle Helium Turbine and With Steam Turbine Cycle for Combined Power and Steam Generation

1973 ◽  
Vol 95 (1) ◽  
pp. 11-18 ◽  
Author(s):  
K. Bammert ◽  
R. Buende
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Closed Cycle ◽  
Steam Generation ◽  
Single Cycle ◽  
Cycle System ◽  
System A ◽  
Secondary Steam

The heat of a helium-cooled reactor can be used for combined power and steam generation either in a closed-cycle helium turbine system, the so-called single-cycle system, or in a two-cycle system which consists of a helium cycle and a secondary steam turbine cycle. The optimum data for the two systems are determined within the same range of general parameters—electric power output and quantity and quality of the steam produced—as functions of the special parameters of each particular cycle system. A method of comparing different power plant systems is shown. With this method it is possible to determine those ranges in which the efficiencies achieved with one system are higher than those obtained with the other. It is described in which way the dividing line between such ranges depends on the special parameters of the cycles. The comparison shows that the single-cycle system offers advantages.

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).

An Efficient, Flexible Arrangement to Generate High Quality Process or Domestic Heat in Recuperative Gas Cycle Power Plants: Application to Helium Direct Cycle Nuclear Power Plants

1979 ◽  
Vol 101 (1) ◽  
pp. 130-140 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre ◽  
P. F. Jude
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Hot Water ◽  
Primary Circuit ◽  
Inlet Temperature ◽  
Total Efficiency ◽  
High Quality ◽  
Quality Process ◽  
Direct Cycle

The advantages of gas turbine power plants in general and closed cycle systems under gas pressure in particular for waste heat recovery are well known. A satisfactory efficiency for electric power generation and good conditions to obtain a significant amount of hot water above 100°C lead to a high fuel utilization. However, as in most of projects, it is not much possible to produce high temperature steam or water without significantly decreasing the electricity production. A new method for an additional generation of high quality process or domestic heat is proposed. The basic feature of this method lies in arranging one or two steam generators or preheaters in parallel with the low pressure side of the recuperator. The high total efficiency and the noteworthy flexibility of this system are emphasized. This arrangement is suitable for any kind of heat source, but the applications presented in this paper are related to helium direct cycle nuclear power plants the main features of which are a single 600 MW(e) turbomachine, a turbine inlet temperature of 775°C, no or one intermediate cooling and a primary circuit fully integrated in a pre-stressed concrete reactor vessel.

Study on Primary and Secondary Heat-Transport Systems for Sodium-Cooled Fast Reactor

2013 ◽  
Author(s):  
Alexey Dragunov ◽  
Eugene Saltanov ◽  
Igor Pioro ◽  
Glenn Harvel ◽  
Brian Ikeda
Keyword(s):  
Heat Exchanger ◽  
Gas Turbine ◽  
Heat Transport ◽  
Fast Reactor ◽  
Thermal Efficiency ◽  
Nuclear Power ◽  
Power Plants ◽  
Power Conversion ◽  
Gas Turbine Cycle ◽  
Steam Cycle

One of the current engineering challenges is to design next generation (Generation IV) Nuclear Power Plants (NPPs) with significantly higher thermal efficiencies (43–55%) compared to those of current NPPs to match or at least to be close to the thermal efficiencies reached at fossil-fired power plants (55–62%). The Sodium-cooled Fast Reactor (SFR) is one of the six concepts considered under the Generation IV International Forum (GIF) initiative. The BN-600 reactor is a sodium-cooled fast-breeder reactor built at the Beloyarsk NPP in Russia. This concept is the only one from the Generation IV nuclear-power reactors, which is actually in operation (since 1980’s). At the secondary side, it uses a subcritical-pressure Rankine-steam cycle with heat regeneration. The reactor generates electrical power in the amount of 600 MWel. The reactor core dimensions are 0.75 m (height) by 2.06 m (diameter). The UO2 fuel enriched to 17–26% is utilized in the core. There are 2 loops (circuits) for sodium flow. For safety reasons, sodium is used both in the primary and the intermediate circuits. Therefore, a sodium-to-sodium heat exchanger is used to transfer heat from the primary loop to the intermediate one. In this work major parameters of the reactor are listed. The actual scheme of the power-conversion heat-transport system is presented; and the results of the calculation of thermal efficiency of this scheme are analyzed. Details of the heat-transport system, including parameters of the sodium-to-sodium heat exchanger and main coolant pump, are presented. In this paper two possibilities for the SFR in terms of the power-conversion cycle are investigated: 1. a subcritical-pressure Rankine-steam cycle through a heat exchanger (current approach in Russian and Japanese power reactors); 2. a supercritical-pressure CO2 Brayton gas-turbine cycle through a heat exchanger (US approach). With the advent of modern super-alloys, the Rankine-steam cycle has progressed into the supercritical region of the coolant and is generating thermal efficiencies into the mid 50% range. Therefore, the thermal efficiency of a supercritical Rankine-steam cycle is also briefly discussed in this paper. According to GIF, the Brayton gas-turbine cycle is under consideration for future nuclear power reactors. The supercritical-CO2 cycle is a new approach in the Brayton gas-turbine cycle. Therefore, dependence of the thermal efficiency of this SC CO2 cycle on inlet parameters of the gas turbine is also investigated.

Efficient Power Generation From Large 750°C Heat Sources: Application to Coal-Fired and Nuclear Power Stations

1980 ◽  
Author(s):  
Z. P. Tilliette ◽  
B. Pierre
Keyword(s):  
Power Generation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Inlet Temperature ◽  
Heat Sources ◽  
Turbine Inlet Temperature ◽  
Power Stations ◽  
Efficient Power ◽  
Steam Cycle

Considering the concern about a more efficient, rational use of heat sources, and a greater location flexibility of power plants owing to cooling capability, closed gas cycles can offer new solutions for fossil or nuclear energy. An efficient heat conversion into power is obtained by the combination of a main non-intercooled helium cycle with a flexible, superheated, low-pressure bottoming steam cycle. Emphasis is placed on the matching of the two cycles and, for that, a recuperator bypass arrangement is used. The operation of the main gas turbocompressor does not depend upon the operation of the small steam cycle. Results are presented for a conservative turbine inlet temperature of 750 C. Applications are made for a coal-fired power plant and for a nuclear GT-HTGR. Overall net plant efficiencies of 39 and 46 percent, respectively, are projected.

scholarly journals Experimental Investigation of a Small-Scale ORC Power Plant Using a Positive Displacement Expander with and without a Regenerator

Energies ◽  
2019 ◽  
Vol 12 (8) ◽  
pp. 1452 ◽  
Author(s):  
Collings ◽  
Mckeown ◽  
Wang ◽  
Yu
Keyword(s):  
Heat Exchanger ◽  
Power Plants ◽  
Large Scale ◽  
Internal Heat ◽  
Expansion Ratio ◽  
Operating Conditions ◽  
Working Fluid ◽  
Small Scale ◽  
Regenerative Heat ◽  
Positive Displacement

While large-scale ORC power plants are a relatively mature technology, their application to small-scale power plants (i.e., below 10 kW) still encounters some technical challenges. Positive displacement expanders are mostly used for such small-scale applications. However, their built-in expansion ratios are often smaller than the expansion ratio required for the maximum utilisation of heat sources, leading to under expansion and consequently higher enthalpy at the outlet of the expander, and ultimately resulting in a lower thermal efficiency. In order to overcome this issue, one possible solution is to introduce an internal heat exchanger (i.e., the so-called regenerator) to recover the enthalpy exiting the expander and use it to pre-heat the liquid working fluid before it enters the evaporator. In this paper, a small-scale experimental rig (with 1-kW rated power) was designed and built that is capable of switching between regenerative and non-regenerative modes, using R245fa as the working fluid. It has been tested under various operating conditions, and the results reveal that the regenerative heat exchanger can recover a considerable amount of heat when under expansion occurs, increasing the cycle efficiency.

Research on Thermal Efficiencies of Various Power Cycles for GFRs and VHTRs

2018 ◽  
Author(s):  
Mohammed Mahdi ◽  
Roman Popov ◽  
Igor Pioro
Keyword(s):  
Gas Turbine ◽  
Heavy Water ◽  
Nuclear Power ◽  
Power Plants ◽  
Supercritical Pressure ◽  
Combined Cycle ◽  
Power Cycles ◽  
Power Cycle ◽  
Combined Cycles ◽  
Gas Turbine Cycle

The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs. Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle. Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N2 mixture, application of SuperCritical (SC)-CO2 cycles or combined cycles. The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO2 is provided.

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