The Experts below are selected from a list of 8208 Experts worldwide ranked by ideXlab platform
Yunqing Bai - One of the best experts on this subject based on the ideXlab platform.
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preliminary design and analysis on the cogeneration system for small modular lead cooled Fast Reactor
Applied Thermal Engineering, 2020Co-Authors: Ming Jin, Tao Zhou, Fanli Kong, Sheng Gao, Yunqing BaiAbstract:Abstract Small Modular Lead-Cooled Fast Reactor (SMLFR) is characterized by high safety, high efficiency, adaptability and versatility. The integrated system design adapts to more site conditions. Cogeneration is an advanced form of energy utilization based on the principle of energy cascade utilization, in which energy is mainly used to produce electricity and waste heat is used for heating at the same time. In this paper, the preliminary design of cogeneration system for SMLFR with the extraction-condensing and back-pressure cogeneration (EBC) type is proposed based on mathematical modeling and thermodynamically analysis. Based on a 35 MWth SMLFR, the effects of the main parameters and mass flow rate of heating steam on the power generation and efficiency are analyzed. A certain urban area in northern China with about 800,000 m2 heating area is chosen to analyze the heating capacity, power generation capacity and economy under different cogeneration types by thermodynamic calculation. The work aims to point out the optimization direction of system integration for the Lead-Cooled Fast Reactor (LFR) cogeneration system by the thermodynamic analysis. This cogeneration system can provide a reference for the secondary loop design of Generation-IV (GEN-IV) Reactors.
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analysis of core blockage scenarios during pump shutdown accidents for small size lead cooled Fast Reactor using relap5 hd
Progress in Nuclear Energy, 2019Co-Authors: Xiaoliang Zou, Tao Zhou, Guangyu Zhang, Ming Jin, Yunqing BaiAbstract:Abstract Two models of a designed small size Lead-Cooled Fast Reactor (LFR) were built by RELAP5-HD to simulate the behavior of the Reactor system during pump shutdown accidents, along with different levels of core blockage. Two typical pump shutdown accidents of pump rotor seizure accident and loss-of-pump-power accident were selected to do the simulation and analysis of the behavior of the nuclear power Reactor during pump shutdown accidents, when all the safety systems were assumed to be unavailable. A one dimensional vessel-one dimensional core model was built to simulate the long term cooling condition of the pump shutdown accidents under the Reactor shutdown or operation conditions. The simulation results would help to analyze whether pump accidents could cause core damage. In order to show the heat removing capability of the coolant in the Reactor system, a more detailed one dimensional vessel-three dimensional core model was built to simulate the long term cooling condition of the pump shutdown accidents along with different levels of core blockage accidents.
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development strategy and conceptual design of china lead based research Reactor
Annals of Nuclear Energy, 2016Co-Authors: Yunqing Bai, Yong Song, Qunying Huang, Zhumin ZhaoAbstract:Abstract Chinese Academy of Sciences (CAS) launched an engineering project to develop an Accelerator Driven System (ADS) for nuclear waste transmutation since 2011, and China LEAd-based Reactor (CLEAR) proposed by Institute of Nuclear Energy Safety Technology (INEST) is selected as the ADS reference Reactor. In this paper, the development strategy and conceptual design of China Lead-based Research Reactor are proposed. The Chinese ADS development program consists of three stages, and during the first stage, a 10 MW th lead-based research Reactor named CLEAR-I will be built with subcritical and critical dual-mode operation capability for validation of ADS transmutation system and lead cooled Fast Reactor technology. Major design principles of CLEAR-I are oriented at technology feasibility, safety reliability, experiment flexibility and technology continuity. Followed by the development strategy and design principles, CLEAR-I design options and conceptual design scenarios are presented.
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assessment of rvacs performance for small size lead cooled Fast Reactor
Annals of Nuclear Energy, 2015Co-Authors: Ming Jin, Jiayue Chen, Yunqing BaiAbstract:Abstract Reactor Vessel Air Cooling System (RVACS), adopted by China Lead-based Research Reactor (CLEAR-I), was considered as a promising solution to passively remove the decay heat under accident conditions. In order to assess CLEAR-I RVACS performance, the investigation of CLEAR-I RVACS performance has been conducted by using the RELAP5 code. The RELAP5 model of CLEAR-I RVACS was established and verified by comparison with three-dimensional computational fluid dynamics commercial code CFX calculation. The maximum heat removal capability of CLEAR-I RVACS was investigated. Several design parameters, such as presence of air cool tubes, vessel size and chimney height, have been studied to evaluate their effect on the heat removal capability. It is suggested that CLEAR-I RVACS had the capability to remove the CLEAR-I decay heat and could apply for the 25 MW pool type Fast Reactors as an independent decay heat removal system. The heat removal capability of the RVACS with air channel was 7.5% bigger than that of the RVACS with air tubes for the same main vessel and air thermal cycle. A 10 m increase in chimney height resulted in a 5 °C decrease in the main and safety vessel temperatures. A 10% increase in the whole size of CLEAR-I RVACS resulted in a 20 °C decrease in the main and safety vessel temperatures.
James J Sienicki - One of the best experts on this subject based on the ideXlab platform.
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lead cooled system design and challenges in the frame of generation iv international forum
Journal of Nuclear Materials, 2011Co-Authors: Luciano Cinotti, Craig F Smith, Hiroshi Sekimoto, L Mansani, Marco Reale, James J SienickiAbstract:Abstract The Generation IV International Forum (GIF) Technology Roadmap identified the Lead-Cooled Fast Reactor (LFR) as a technology well suited for electricity generation, hydrogen production and actinide management in a closed fuel cycle. One of the most important features of the LFR is the fact that lead is a relatively inert coolant, a feature that conveys significant advantages in terms of safety, system simplification, and the consequent potential for economic performance. In 2004, the GIF LFR Provisional System Steering Committee was organized and began to develop the LFR System Research Plan. The committee selected two pool-type Reactor concepts as candidates for international cooperation and joint development in the GIF framework: these are the Small Secure Transportable Autonomous Reactor (SSTAR); and the European Lead-Cooled System (ELSY). The high boiling point (1745 °C) of lead has a beneficial impact to the safety of the system, whereas its high melting point (327.4 °C) requires new engineering strategies, especially for In-Service-Inspection and refuelling. Lead, especially at high temperatures, is also relatively corrosive towards structural materials. This necessitates that coolant purity and the level of dissolved oxygen be carefully controlled, in addition to the proper selection of structural materials. For the GIF LFR concepts, lead has been chosen as the coolant rather than Lead–Bismuth Eutectic primarily because of its greatly reduced generation of the alpha-emitting 210 Po isotope formed in the coolant. This results in significantly reduced levels of radioactive contamination of the coolant while minimizing the effect of decay power in the coolant from such contaminants; an additional consideration is the desire to eliminate dependence on bismuth which might be a limited resource. This paper provides an overview of the historical development of the LFR, a summary of the advantages and challenges associated with heavy liquid metal coolants, and an update of the current status of development of LFR concepts under consideration. The main characteristics of the SSTAR and ELSY systems are summarized, and the current status of design of each system is presented. Because of the significant recent efforts in the ELSY system design, greater emphasis is placed on the ELSY plant, with focus on the technological development and design provisions intended to overcome or alleviate recognized drawbacks to the use of heavy liquid metal coolants. In the case of the SSTAR system for which development has proceeded more slowly, a more limited summary is provided. It is noted that both systems share many of the same research needs and objectives thus providing a strong basis for international collaboration.
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core design investigation for a superstar small modular lead cooled Fast Reactor demonstrator
Nuclear Engineering and Design, 2011Co-Authors: Sara Bortot, James J Sienicki, Anton Moisseytsev, C ArtioliAbstract:Abstract In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration has been developed meeting the foremost requirements of limiting the reactivity swing over the core lifetime to about 1 $ and flattening the radial power profiles, as demanded by the choice of wrapper-less (i.e. without flow ducts) fuel assemblies and by the stringent technological constraints imposed by the requirement of short-term deployment. Reactivity coefficients and kinetic parameters have been evaluated for the reference beginning-of-life, middle-of-life and end-of-life core configurations. Furthermore, the results of thermal-hydraulic analyses of the primary loop have confirmed that the system can be effectively cooled by natural circulation heat transport, all the technological constraints being respected even when incorporating peaking factors.
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transient accident analysis of a supercritical carbon dioxide brayton cycle energy converter coupled to an autonomous lead cooled Fast Reactor
Nuclear Engineering and Design, 2008Co-Authors: Anton Moisseytsev, James J SienickiAbstract:Abstract The supercritical carbon dioxide (S-CO2) Brayton cycle is a promising advanced alternative to the Rankine steam cycle and recuperated gas Brayton cycle for the energy converters of specific Reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the S-CO2 Brayton cycle coupled to an autonomous, natural circulation Lead-Cooled Fast Reactor (LFR). The plant dynamics code was used to simulate the whole-plant response to accident conditions. The specific design features of the Reactor concept influencing passive safety are discussed and accident scenarios are identified for analysis. Results of calculations of the whole-plant response to loss-of-heat sink, loss-of-load, and pipe break accidents are demonstrated. The passive safety performance of the Reactor concept is confirmed by the results of the plant dynamics code calculations for the selected accident scenarios.
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sstar the us lead cooled Fast Reactor lfr
Journal of Nuclear Materials, 2008Co-Authors: Craig F Smith, James J Sienicki, William G Halsey, Neil W Brown, Anton Moisseytsev, D C WadeAbstract:Abstract It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10–700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized Reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-Cooled systems offer several key advantages in meeting these goals. The small Lead-Cooled Fast Reactor concept known as the small secure transportable autonomous Reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV Lead-Cooled Fast Reactor system.
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status of development of the small secure transportable autonomous Reactor sstar for worldwide sustainable nuclear energy supply
2008Co-Authors: James J Sienicki, Anton Moisseytsev, D C Wade, A NikiforovaAbstract:Significant progress and improvements have been made on development of a pre-conceptual design of the Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) concept since it was last reported on at ICAPP 05. SSTAR is a small, 20 MWe (45 MWt), exportable, natural circulation, Fast Reactor plant concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety. Customers of SSTAR include: (1) clients looking for energy security at small capital outlay; (2) cities in developing nations; and (3) deregulated independent power producers in developed nations. The SSTAR pre-conceptual design integrates three major features: primary coolant natural circulation heat transport; lead (Pb) coolant; and transuranic nitride fuel in a pool vessel configuration. The Pb coolant flows upward through the core which is an open-lattice of large-diameter (2.5 centimeter) fuel pins containing transuranic nitride pellets clad bonded with liquid Pb to silicon-enhanced ferritic/martensitic (F/M) stainless steel arranged on a triangular pitch with spacing maintained by grid spacers; the core does not incorporate removable fuel assemblies as one means of restricting accessmore » to the fuel. The whole core is a single removable assembly with a long lifetime (30 years) at which time refueling equipment is brought onsite. Conversion of the core thermal energy to electricity is accomplished using a supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle energy converter providing higher plant efficiencies and lower balance of plant costs than the traditional Rankine steam cycle operating at the same Reactor core outlet temperature. A control strategy has been developed for automatic control of the S-CO{sub 2} Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power whereby the Reactor core power adjusts itself to the heat removal from the Reactor system to the power converter through the large reactivity feedback of the Fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the Reactor to the grid load. A safety design approach has been formulated for SSTAR based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials. The inherent safety features of the lead coolant (T{sub boil} = 1740 C, lack of chemical reaction of Pb with the CO{sub 2} working fluid, low absorption of neutrons by Pb, and the heavy Pb), nitride fuel (high thermal conductivity, transuranic nitride decomposition temperature {approx} 1300 C, compatibility with cladding, low volumetric swelling and fission gas release), Fast neutron spectrum core, pool vessel configuration, natural circulation, and containment enable the requirements for each level of protection to be readily met or exceeded. The interest in higher plant efficiencies has heretofore driven interest in operation of SSTAR at higher Pb temperatures to take advantage of the increase in plant efficiency with temperature of the S-CO{sub 2} Brayton cycle. A peak cladding temperature of 650 C has been used as a goal; at this temperature, a Reactor core outlet temperature of 564 C is achieved resulting in a Brayton cycle efficiency of 44.2 % and a net plant efficiency of 43.8 %. It has always been recognized that this would require the development of cladding and structural materials for long-term service in Pb coolant up to 650 C peak cladding temperature with corrosion protection provided by active maintenance and control of the dissolved oxygen potential in the coolant giving rise to the formation of protective oxide layers on the steel cladding and structures. SSTAR development has been supported by the testing in the DELTA loop at Los Alamos National Laboratory of alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR operates. The focus of LFR development in the U.S. is now shifting towards the development of a near-term deployable LFR test demonstrator and a near-term deployable small exportable LFR. Both Reactors would operate at lower temperatures enabling the use of existing materials such as T91 or HT9 F/M stainless steel that is already incorporated into the ASME codes and have been shown to have corrosion resistance to lead-bismuth eutectic with active oxygen control at temperatures below about 550 C in experiments carried out in the DELTA loop and elsewhere.« less
C Artioli - One of the best experts on this subject based on the ideXlab platform.
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conceptual core design study for a high flux lfr demonstrator
Progress in Nuclear Energy, 2012Co-Authors: Sara Bortot, Patrizio Console Camprini, Giacomo Grasso, C ArtioliAbstract:A preconceptual core design study for a pool-type Lead-Cooled Fast Reactor (LFR) demonstrator (DEMO) has been developed in the frame of an I-NERI between the Italian National Agency for the New Technologies, Energy and Sustainable Economic Development (ENEA) and Argonne National Laboratory (ANL), based on the European Lead-Cooled System (ELSY) reference concept. A demonstration Reactor is expected to prove the viability of technology to be implemented in the first-of-a-kind industrial power plant. DEMO specifications as a nuclear power facility demonstrating ELSY main features and performance, besides validating design methodology and tools, have been defined. Suitable design parameters have been set to meet the foremost objective of reaching a high Fast neutron flux while respecting all technological constraints. Preliminary thermal-hydraulic analyses have been carried out to verify safety limits were not exceeded.
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core design investigation for a superstar small modular lead cooled Fast Reactor demonstrator
Nuclear Engineering and Design, 2011Co-Authors: Sara Bortot, James J Sienicki, Anton Moisseytsev, C ArtioliAbstract:Abstract In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration has been developed meeting the foremost requirements of limiting the reactivity swing over the core lifetime to about 1 $ and flattening the radial power profiles, as demanded by the choice of wrapper-less (i.e. without flow ducts) fuel assemblies and by the stringent technological constraints imposed by the requirement of short-term deployment. Reactivity coefficients and kinetic parameters have been evaluated for the reference beginning-of-life, middle-of-life and end-of-life core configurations. Furthermore, the results of thermal-hydraulic analyses of the primary loop have confirmed that the system can be effectively cooled by natural circulation heat transport, all the technological constraints being respected even when incorporating peaking factors.
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a new paradigm for core design aimed at the sustainability of nuclear energy the solution of the extended equilibrium state
Annals of Nuclear Energy, 2010Co-Authors: C Artioli, Giacomo Grasso, C PetrovichAbstract:The future expansion of nuclear energy, a technology identified as one of the main candidates for reducing the world’s dependence on fossil fuels, requires a thorough analysis of the sustainability of this energy source for long-term supply. Generation-IV nuclear systems could represent a turning point for energy production by minimizing the environmental footprint of the fuel cycle. A new paradigm is thus required for Reactor design, focusing, at the core design level, on both the closure of the fuel cycle and the effective utilization of natural resources. Within this framework, the so-called “adiabatic core” concept represents a particularly interesting solution. It is based on the idea of ensuring by design a condition of equilibrium in the fuel cycle (i.e., an equilibrium “fuel vector”), foreseeing nuclear power systems able to maintain a constant total amount of both plutonium and minor actinides (TRU), consuming only uranium (either natural or depleted), while discharging to the environment only fission products and reprocessing losses. Under such a hypothesis, all actinides can be continuously recycled in the same system, reducing both the waste volume and its long-term radiotoxicity, as well as utilizing effectively uranium resources. Two mathematical approaches have been devised to find the “extended” equilibrium solution for the fuel vector. These methods are compared, validated with the codes MCNPX and FISPACT and applied to the European Lead-Cooled Fast Reactor ELSY, confirming the potential of this approach (e.g., a reduction by two orders of magnitude of the TRU mass in the final waste in comparison with the fuel cycle of Light Water Reactors operated in a once-through scenario).
Anton Moisseytsev - One of the best experts on this subject based on the ideXlab platform.
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core design investigation for a superstar small modular lead cooled Fast Reactor demonstrator
Nuclear Engineering and Design, 2011Co-Authors: Sara Bortot, James J Sienicki, Anton Moisseytsev, C ArtioliAbstract:Abstract In this paper a preconceptual neutronics design study for a SUstainable Proliferation-resistance Enhanced Refined Secure Transportable Autonomous Reactor (SUPERSTAR) demonstrator is presented. The main goal of achieving the highest realistic power level limited by natural circulation and transportability, while providing energy security and proliferation resistance thanks to a long core lifetime design has been satisfactorily attained. A preliminary core configuration has been developed meeting the foremost requirements of limiting the reactivity swing over the core lifetime to about 1 $ and flattening the radial power profiles, as demanded by the choice of wrapper-less (i.e. without flow ducts) fuel assemblies and by the stringent technological constraints imposed by the requirement of short-term deployment. Reactivity coefficients and kinetic parameters have been evaluated for the reference beginning-of-life, middle-of-life and end-of-life core configurations. Furthermore, the results of thermal-hydraulic analyses of the primary loop have confirmed that the system can be effectively cooled by natural circulation heat transport, all the technological constraints being respected even when incorporating peaking factors.
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transient accident analysis of a supercritical carbon dioxide brayton cycle energy converter coupled to an autonomous lead cooled Fast Reactor
Nuclear Engineering and Design, 2008Co-Authors: Anton Moisseytsev, James J SienickiAbstract:Abstract The supercritical carbon dioxide (S-CO2) Brayton cycle is a promising advanced alternative to the Rankine steam cycle and recuperated gas Brayton cycle for the energy converters of specific Reactor concepts belonging to the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. A new plant dynamics analysis computer code has been developed for simulation of the S-CO2 Brayton cycle coupled to an autonomous, natural circulation Lead-Cooled Fast Reactor (LFR). The plant dynamics code was used to simulate the whole-plant response to accident conditions. The specific design features of the Reactor concept influencing passive safety are discussed and accident scenarios are identified for analysis. Results of calculations of the whole-plant response to loss-of-heat sink, loss-of-load, and pipe break accidents are demonstrated. The passive safety performance of the Reactor concept is confirmed by the results of the plant dynamics code calculations for the selected accident scenarios.
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status report on the small secure transportable autonomous Reactor sstar lead cooled Fast Reactor lfr and supporting research and development
2008Co-Authors: J J Sienicki, Anton Moisseytsev, D C Wade, A Nikiforova, W S Yang, P Hanania, H J Ryu, K P Kulesza, S J Kim, William G HalseyAbstract:This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, Fast Reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the Reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the Reactor core power adjusts itself to equal the heat removal from the Reactor system to the power converter through the large reactivity feedback of the Fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the Reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled Fast Reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics computer codes and methodologies applied to SSTAR. Another section of the report details the SSTAR safety design approach which is based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials and how the inherent safety features of the lead coolant, nitride fuel, Fast neutron spectrum core, pool vessel configuration, natural circulation, and containment meet or exceed the requirements for each level of protection. The report also includes recent results of a systematic analysis by LANL of data on corrosion of candidate cladding and structural material alloys of interest to SSTAR by LBE and Pb coolants; the data were taken from a new database on corrosion by liquid metal coolants created at LANL. The analysis methodology that considers penetration of an oxidation front into the alloy and dissolution of the trailing edge of the oxide into the coolant enables the long-term corrosion rate to be extracted from shorter-term corrosion data thereby enabling an evaluation of alloy performance over long core lifetimes (e.g., 30 years) that has heretofore not been possible. A number of candidate alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR would operate were analyzed following testing in the DELTA loop at LANL including steels that were treated by laser peening at LLNL; laser peening is an approach that alters the oxide-metal bonds which could potentially improve corrosion resistance. LLNL is also carrying out Multi-Scale Modeling of the Fe-Cr system with the goal of assisting in the development of cladding and structural materials having greater resistance to irradiation.
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sstar the us lead cooled Fast Reactor lfr
Journal of Nuclear Materials, 2008Co-Authors: Craig F Smith, James J Sienicki, William G Halsey, Neil W Brown, Anton Moisseytsev, D C WadeAbstract:Abstract It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10–700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized Reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-Cooled systems offer several key advantages in meeting these goals. The small Lead-Cooled Fast Reactor concept known as the small secure transportable autonomous Reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV Lead-Cooled Fast Reactor system.
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status of development of the small secure transportable autonomous Reactor sstar for worldwide sustainable nuclear energy supply
2008Co-Authors: James J Sienicki, Anton Moisseytsev, D C Wade, A NikiforovaAbstract:Significant progress and improvements have been made on development of a pre-conceptual design of the Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) concept since it was last reported on at ICAPP 05. SSTAR is a small, 20 MWe (45 MWt), exportable, natural circulation, Fast Reactor plant concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety. Customers of SSTAR include: (1) clients looking for energy security at small capital outlay; (2) cities in developing nations; and (3) deregulated independent power producers in developed nations. The SSTAR pre-conceptual design integrates three major features: primary coolant natural circulation heat transport; lead (Pb) coolant; and transuranic nitride fuel in a pool vessel configuration. The Pb coolant flows upward through the core which is an open-lattice of large-diameter (2.5 centimeter) fuel pins containing transuranic nitride pellets clad bonded with liquid Pb to silicon-enhanced ferritic/martensitic (F/M) stainless steel arranged on a triangular pitch with spacing maintained by grid spacers; the core does not incorporate removable fuel assemblies as one means of restricting accessmore » to the fuel. The whole core is a single removable assembly with a long lifetime (30 years) at which time refueling equipment is brought onsite. Conversion of the core thermal energy to electricity is accomplished using a supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle energy converter providing higher plant efficiencies and lower balance of plant costs than the traditional Rankine steam cycle operating at the same Reactor core outlet temperature. A control strategy has been developed for automatic control of the S-CO{sub 2} Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power whereby the Reactor core power adjusts itself to the heat removal from the Reactor system to the power converter through the large reactivity feedback of the Fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the Reactor to the grid load. A safety design approach has been formulated for SSTAR based upon defense-in-depth providing multiple levels of protection against the release of radioactive materials. The inherent safety features of the lead coolant (T{sub boil} = 1740 C, lack of chemical reaction of Pb with the CO{sub 2} working fluid, low absorption of neutrons by Pb, and the heavy Pb), nitride fuel (high thermal conductivity, transuranic nitride decomposition temperature {approx} 1300 C, compatibility with cladding, low volumetric swelling and fission gas release), Fast neutron spectrum core, pool vessel configuration, natural circulation, and containment enable the requirements for each level of protection to be readily met or exceeded. The interest in higher plant efficiencies has heretofore driven interest in operation of SSTAR at higher Pb temperatures to take advantage of the increase in plant efficiency with temperature of the S-CO{sub 2} Brayton cycle. A peak cladding temperature of 650 C has been used as a goal; at this temperature, a Reactor core outlet temperature of 564 C is achieved resulting in a Brayton cycle efficiency of 44.2 % and a net plant efficiency of 43.8 %. It has always been recognized that this would require the development of cladding and structural materials for long-term service in Pb coolant up to 650 C peak cladding temperature with corrosion protection provided by active maintenance and control of the dissolved oxygen potential in the coolant giving rise to the formation of protective oxide layers on the steel cladding and structures. SSTAR development has been supported by the testing in the DELTA loop at Los Alamos National Laboratory of alloy specimens with special treatments or coatings which might enhance corrosion resistance at the temperatures at which SSTAR operates. The focus of LFR development in the U.S. is now shifting towards the development of a near-term deployable LFR test demonstrator and a near-term deployable small exportable LFR. Both Reactors would operate at lower temperatures enabling the use of existing materials such as T91 or HT9 F/M stainless steel that is already incorporated into the ASME codes and have been shown to have corrosion resistance to lead-bismuth eutectic with active oxygen control at temperatures below about 550 C in experiments carried out in the DELTA loop and elsewhere.« less
Hongli Chen - One of the best experts on this subject based on the ideXlab platform.
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development and validation of a coupled neutron diffusion thermal hydraulic calculation procedure for Fast Reactor applications
Annals of Nuclear Energy, 2020Co-Authors: Xuebei Zhang, Qin Zeng, Hongli ChenAbstract:Abstract The neutron diffusion equation is defined based on the User Defined Function (UDF) and the User Defined Scalar (UDS) functions of the FLUENT. The neutron diffusion equation is solved iteratively by using the solver of the FLUENT with the Finite Volume Method (FVM). At the same time, the mass, momentum and energy equations are solved iteratively. At each iteration, the power distribution (flux distribution) obtained by the iteration of the neutron diffusion equation is transferred to the thermal-hydraulics calculation and is used as the heat source term. At the same time, the temperature distribution obtained from the thermal-hydraulics calculation is transferred to the neutron diffusion calculation and the macroscopic cross sections of the materials are corrected to realize the coupling calculation of the neutron diffusion and the thermal-hydraulics under the same solver of the FLUENT without needing to develop the interface program and the computational cost is saved. 2D-TWIGL benchmark problem is calculated by the FLUENT solver to verify the feasibility of this method to solve neutron diffusion equation. Through the modeling and calculation of the 5 × 5 PWR assembly model, the calculation results are compared with the results of other programs to verify the feasibility of the coupling method and the correctness of data transfer. Then this coupling method is applied to calculate the hot assembly of a modular Lead-Cooled Fast Reactor (M2LFR-1000) to verify that the thermal-hydraulics characteristics (the maximum fuel temperature and the maximum cladding outer surface temperature) are all within the corresponding thermal-hydraulics design limits.
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transient sub channel code development for lead cooled Fast Reactor using the second order upwind scheme
Progress in Nuclear Energy, 2019Co-Authors: Liankai Cao, Guangliang Yang, Hongli ChenAbstract:Abstract For its unique safety and economic advantages, the lead cooled Fast Reactor has become one of the most interesting candidate Reactors for the Generation IV nuclear system. To study the thermal-hydraulics of the core under transient conditions, KMC-SUBtra--a sub-channel code for transient thermal-hydraulic analysis of lead cooled Fast Reactor has been developed. The code uses a modified pressure gradient method to solve the simultaneous equations of the fluid mass, momentum and energy containing cross flow and turbulent mixing, in which the axial pressure gradients are solved as pending variables of the simultaneous equations. A staggered mesh scheme is used for scalar and vector quantities and the second-order upwind scheme is adopted in the discretization of the convection term. The code was validated and verified by the experimental data and CFD simulation results on the steady and transient conditions so its capability for lead cooled Reactors was confirmed. The transient flow and heat transfer in the fuel assembly with time-varying boundary conditions were studied, which revealed different transient thermal characteristics of the coolant and fuel rods. In addition, the results calculated using different order upwind schemes were compared and showed the second-order derivative of the axial flow is small.
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Transient safety analysis of M2LFR-1000 Reactor using ATHLET
Elsevier, 2019Co-Authors: Chong Shen, Liankai Cao, Xilin Zhang, Chi Wang, Hongli ChenAbstract:M2LFR-1000 is a medium-power modular Lead-Cooled Fast Reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a Reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of M2LFR-1000 Reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft für Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the M2LFR is safe enough under these three transients due to the good inherent safety features of the Reactor, without human intervention, the Reactor will reach a new steady state under UTOP condition. Keywords: Transient analysis, LFR, Forced circulation, ATHLE
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conceptual design of a small modular natural circulation lead cooled Fast Reactor snclfr 100
International Journal of Hydrogen Energy, 2016Co-Authors: Hongli Chen, Zhao Chen, Chong Chen, Xilin Zhang, Haoran Zhang, Pengcheng Zhao, Shuzhou Li, Jingchao Feng, Qin ZengAbstract:Abstract SNCLFR-100, a 100 MW th Lead-Cooled small modular Reactor with a passive cooling feature to both normal and abnormal operations, was proposed by University of Science and Technology of China (USTC). The Reactor is well suited as a remote power source because of its compact size, as well as because it has a refueling interval of 10 years without assembly reconfiguration. The Reactor is a typical pool-type Fast Reactor with an array of heterogeneous square fuel assemblies loaded with MOX fuels. In this paper, the overall design and neutronics features were illustrated and evaluated. The steady state thermal-hydraulic performance, mass flow distribution characteristics and sub-channel T/H features were analyzed and discussed. Two major accident scenarios including unprotected overpower transient (UTOP) and unprotected loss of heat sink transient (ULOHS) were selected for a first evaluation of its dynamic behavior. The results show that the safety criteria are satisfied and Reactor is tolerant to the UTOP and ULOHS transients. This implies that the conceptual design of SNCLFR-100 is acceptable and the Reactor has excellent inherent safety characteristics.