A Supercritical CO2 Brayton Cycle Power Converter for a Sodium-Cooled Fast Reactor Small Modular Reactor

Author(s):  
James J. Sienicki ◽  
Anton Moisseytsev ◽  
Lubomir Krajtl

Although a number of power conversion applications have been identified or have even been developed (e.g., waste heat recovery) for supercritical carbon dioxide (S-CO2) cycles including fossil fuel combustors, concentrated solar power (i.e., solar power towers), and marine propulsion, the benefits of S-CO2 Brayton cycle power conversion are especially prominent for applications to nuclear power reactors. In particular, the S-CO2 Brayton cycle is well matched to the Sodium-Cooled Fast Reactor (SFR) nuclear power reactor system and offers significant benefits for SFRs. The recompression closed Brayton cycle is highly recuperated and wants to operate with an approximate optimal S-CO2 temperature rise in the sodium-to-CO2 heat exchangers of about 150 °C which is well matched to the sodium temperature rise through the core that is also about 150 °C. Use of the S-CO2 Brayton cycle eliminates sodium-water reactions and can reduce the nuclear power plant cost per unit electrical power. A conceptual design of an optimized S-CO2 Brayton cycle power converter and supporting systems has been developed for the Advanced Fast Reactor – 100 (AFR-100) 100 MWe-class (250 MWt) SFR Small Modular Reactor (SMR). The AFR-100 is under ongoing development at Argonne National Laboratory (ANL) to target emerging markets where a clean, secure, and stable source of electricity is required but a large-scale power plant cannot be accommodated. The S-CO2 Brayton cycle components and cycle conditions were optimized to minimize the power plant cost per unit electrical power (i.e., $/kWe). For a core outlet temperature of 550 °C and turbine inlet temperature of 517 °C, a cycle efficiency of 42.3 % is calculated that exceeds that obtained with a traditional superheated steam cycle by one percentage point or more. A normal shutdown heat removal system incorporating a pressurized pumped S-CO2 loop slightly above the critical pressure on each of the two intermediate sodium loops has been developed to remove heat from the reactor when the power converter is shut down. Three-dimensional layouts of S-CO2 Brayton cycle power converter and shutdown heat removal components and piping have been determined and three-dimensional CAD drawings prepared. The S-CO2 Brayton cycle power converter is found to have a small footprint reducing the space requirements for components and systems inside of both the turbine generator building and reactor building. The results continue to validate earlier notions about the benefits of S-CO2 Brayton cycle power conversion for SFRs including higher efficiency, improved economics, elimination of sodium-water reactions, load following, and smaller footprint.

Author(s):  
Chenggang Yu ◽  
Michael A. Smith ◽  
Earl E. Feldman ◽  
Won Sik Yang ◽  
James J. Sienicki

A scoping design study has been carried out of the feasibility of a small, 25 MWt (∼10 MWe), modular lead-cooled fast reactor coupled to an advanced power converter consisting of a gas turbine Brayton cycle that utilizes supercritical carbon dioxide as the working fluid. Major constraints of the study are an ultralong 20 year core lifetime, near zero reactivity burnup swing over the core lifetime, Pb primary coolant natural circulation heat transport, road transportability of plant modular assemblies including the reactor and guard vessels, and high Brayton cycle power conversion efficiency. It is found that the goal of a near zero reactivity burnup swing implies a low core power density that results in an unacceptably low discharge burnup.


2021 ◽  
Vol 345 ◽  
pp. 00032
Author(s):  
Michal Volf ◽  
Martin Pelikán ◽  
Pavel Žitek

The article focuses on a power conversion system for a gas-cooled fast reactor working with helium. The power conversion system, i.e., secondary and possible tertiary system of a power plant, is used to convert heat generated by nuclear fission into electrical energy. The presented research deals with the conceptual design of this system, mainly its secondary circuit, which is assumed to be a Brayton cycle. Several concepts are evaluated, including single and staged compression and possible heat regeneration. The goal of the work is to select the main parameters of such a cycle that would not only be ideal in terms of efficiency, but would also allow decay heat to be used and further converted into electricity. In this way, the secondary cycle could be used as an additional safety system for the nuclear power plant.


Author(s):  
Tom G. Lewis ◽  
Edward J. Parma ◽  
Steven A. Wright ◽  
Milton E. Vernon ◽  
Darryn D. Fleming ◽  
...  

The advanced nuclear concept group at Sandia National Laboratories has been investigating two advance right size reactors (RSR); this paper will discuss one of these two systems. The supercritical carbon dioxide (S-CO2), direct cycle gas fast reactor (SC-GFR) concept was developed to determine the feasibility of a RSR type concept using S-CO2 as the working fluid in a direct cycle fast reactor. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this paper show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The SC-GFR concept is a relatively small (200 MWth) fast reactor that is cooled with CO2 at a pressure of 20 MPa. The CO2 flows out of the reactor vessel at ∼650°C directly into a turbine-generator unit to produce electrical power. The thermodynamic cycle that is used for the power conversion is a supercritical gas Brayton cycle with CO2 as the working fluid. With the CO2 gas near the critical point after the heat rejection portion of the cycle, it can be compressed with less power as compared to a standard gas Brayton cycle, thereby allowing for a higher thermal efficiency at the same turbine inlet temperature. A cycle efficiency of 45–50% is theoretically achievable for an optimized configuration. The major advantages of the concept include the following: • High thermal efficiency at relatively low reactor outlet temperatures; • Compact, cost-effective, power conversion system; • Non-flammable, stable, inert, non-toxic, inexpensive, and well-characterized coolant; • Potential long-life core and closed fuel cycle; • Small void reactivity worth from loss of coolant; • Natural convection decay heat removal; • Feasible design using today’s technologies. The goal of this work was to develop a SC-GFR concept and perform scoping analyses, including a review of other concepts that are similar in nature, to determine concept feasibility, advantages, disadvantages, and issues requiring further investigation. Overall, the SC-GFR concept as described in this paper appears feasible and warrants further study.


2011 ◽  
Vol 133 (05) ◽  
pp. 30-33 ◽  
Author(s):  
Lee S. Langston

This article explores the increasing use of natural gas in different turbine industries and in turn creating an efficient electrical system. All indications are that the aviation market will be good for gas turbine production as airlines and the military replace old equipment and expanding economies such as China and India increase their air travel. Gas turbines now account for some 22% of the electricity produced in the United States and 46% of the electricity generated in the United Kingdom. In spite of this market share, electrical power gas turbines have kept a much lower profile than competing technologies, such as coal-fired thermal plants and nuclear power. Gas turbines are also the primary device behind the modern combined power plant, about the most fuel-efficient technology we have. Mitsubishi Heavy Industries is developing a new J series gas turbine for the combined cycle power plant market that could achieve thermal efficiencies of 61%. The researchers believe that if wind turbines and gas turbines team up, they can create a cleaner, more efficient electrical power system.


2011 ◽  
Vol 99-100 ◽  
pp. 350-353
Author(s):  
Xiao Bing Sun ◽  
Xu Bin Qiao

As the largest unit capacity of nuclear power plant at present, the flow conduit of circulating water pump in EPR1750 nuclear power plant is a volute conduit, which is a cast-in-situ conceret structure with complexly gradual change cavity. Therefore, the hydraulic efficiency of circulating water pump is not only related with the design of pump leaves, but also closely related to the design of volute and the complicated spatial type of intake and outtake conduits. With the pump leaves and the intake and outtake conduits of conceret volute as the research model, based on computational fluid dynamics (CFD)and the three dimensional Reynolds averaged Navier-Stokes equations, an analytic model suitable for computation is established to simulate the three-dimensional steady flow in the whole pumping system under different operating modes. By use of the commercial fluid-computation softer ANSYS, the distribution of basic physic quantities in the fluid field inside the pump and the conduits is obtained. The analysis and prediction of the performance of pump system are made, and the spatial type design of intake and outtake conduits is evaluated. The calculation results can be referenced to improve the design of pump systems in the similar projects.


Author(s):  
Lei Wan ◽  
Guiyong Li ◽  
Min Rui ◽  
Yongkang Liu ◽  
Jue Yang

A floating nuclear power plant (FNPP) with small modular reactor (SMR) is a combination of a civilian nuclear infrastructure and an offshore installation, which is defined as a floating nuclear facility. The article draws the lessons from studying of the engineer combination like Floating Production Storage and Offloading (FPSO) under the regulation of several government departments. It puts forward recommendations for license application and government regulation as follows in consideration with current license application for nuclear power plant and ship survey. A FNPP shall follow the requirements of construction, fueling and operation for civil nuclear installation combined with ship survey. Application is submitted to nuclear safety regulator for construction permit, while the design drawings shall be submitted to department of ship survey which checks the drawings whether meet the requirements of ship survey, considering some nuclear safety needs. The result of ship survey shall be represented in the safety analysis reports. The construction and important devices manufacturing shall be under the supervision of nuclear installation regulators and ship survey departments. In conclusion, National Nuclear Safety Administration (NNSA) and Maritime Safety Administration of the People’s Republic of China (MSA) shall establish united supervisory system for SMR on sea in China. It is suggested that NNSA is in charge of the overall safety of a FNPP, while MSA is responsible of the ship survey. The operator shall undertake obligation of a FNPP and evaluate the ship cooperating with experienced agency. It is suggested that government departments build the mutual recognition agreement of safety review. It is better to solve the vague questions by coordination.


Author(s):  
Jing Chen ◽  
Dalin Zhang ◽  
Suizheng Qiu ◽  
Kui Zhang ◽  
Mingjun Wang ◽  
...  

As the first developmental step of the sodium-cooled fast reactor (SFR) in China, the pool-type China Experimental Fast Reactor (CEFR) is equipped with the openings and inter-wrapper space in the core, which act as an important part of the decay heat removal system. The accurate prediction of coolant flow in the reactor core calls for complete three-dimensional calculations. In the present study, an investigation of thermal-hydraulic behaviors in a 180° full core model similar to that of CEFR was carried out using commercial Computational Fluid Dynamics (CFD) software. The actual geometries of the peripheral core baffle, fluid channels and narrow inter-wrapper gap were built up, and numerous subassemblies (SAs) were modeled as the porous medium with appropriate resistance and radial power distribution. First, the three-dimensional flow and temperature distributions in the full core under normal operating condition are obtained and quantitatively analyzed. And then the effect of inter-wrapper flow (IWF) on heat transfer performance is evaluated. In addition, the detailed flow path and direction in local inter-wrapper space including the internal and outlet regions are captured. This work can provide some valuable understanding of the core thermal-hydraulic phenomena for the research and design of SFRs.


2018 ◽  
Vol 4 (4) ◽  
pp. 251-256 ◽  
Author(s):  
Sergey Shcheklein ◽  
Ismail Hossain ◽  
Mohammad Akbar ◽  
Vladimir Velkin

Bangladesh lies in a tectonically active zone. Earlier geological studies show that Bangladesh and its adjoining areas are exposed to a threat of severe earthquakes. Earthquakes may have disastrous consequences for a densely populated country. This dictates the need for a detailed analysis of the situation prior to the construction of nuclear power plant as required by the IAEA standards. This study reveals the correlation between seismic acceleration and potential damage. Procedures are presented for investigating the seismic hazard within the future NPP construction area. It has been shown that the obtained values of the earthquake’s peak ground acceleration are at the level below the design basis earthquake (DBE) level and will not lead to nuclear power plant malfunctions. For the most severe among the recorded and closely located earthquake centers (Madhupur) the intensity of seismic impacts on the nuclear power plant site does not exceed eight points on the MSK-64 scale. The existing predictions as to the possibility of a super-earthquake with magnitude in excess of nine points on the Richter scale to take place on the territory of the country indicate the necessity to develop an additional efficient seismic diagnostics system and to switch nuclear power plants in good time to passive heat removal mode as stipulated by the WWER 3+ design. A conclusion is made that accounting for the predicted seismic impacts in excess of the historically recorded levels should be achieved by the establishment of an additional efficient seismic diagnostics system and by timely switching the nuclear power plants to passive heat removal mode with reliable isolation of the reactor core and spent nuclear fuel pools.


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