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David Guzonas - One of the best experts on this subject based on the ideXlab platform.
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Material research for the supercritical Water-Cooled Reactor—summary and open issues
Nuclear Corrosion, 2020Co-Authors: R. Novotny, David GuzonasAbstract:Abstract The supercritical Water-Cooled Reactor (SCWR) is one of the six Reactor concepts included under the umbrella of the Generation IV International Forum. Research on materials and chemistry for SCWRs dates back to the 1960s. A number of Reactor concepts using Water at supercritical temperature but subcritical pressures (nuclear steam) were studied. Significant experience was also obtained from the operation of supercritical fossil-fired power plants. In this chapter, the material requirements of the various SCWR concepts are introduced, with a focus on the European Union’s pressure vessel concept and the Canadian’s pressure tube concept. Furthermore, the key material degradation mechanisms relevant for SCWRs are summarized, state-of-the-art understanding of the key materials degradation issues is reviewed, and knowledge gaps are identified.
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Fission Product Release Under Supercritical Water-Cooled Reactor Conditions
Journal of Nuclear Engineering and Radiation Science, 2016Co-Authors: David Guzonas, Liyan Qiu, S. Livingstone, S. RousseauAbstract:Most supercritical Water-Cooled Reactor (SCWR) concepts being considered as part of the Generation IV initiative are direct cycle. In the event of a fuel defect, the coolant will contact the fuel pellet, potentially releasing fission products and actinides into the coolant and transporting them to the turbines. At the high pressure (25 MPa) in an SCWR, the coolant does not undergo a phase change as it passes through the critical temperature in the core, and nongaseous species may be transported out of the core and deposited on out-of-core components, leading to increased worker dose. It is therefore important to identify species with a high risk of release and develop models of their transport and deposition behavior. This paper presents the results of preliminary leaching tests in SCW of U-Th simulated fuel pellets prepared from natural U and Th containing representative concentrations of the (inactive) oxides of fission products corresponding to a fuel burnup of 60 GWd/ton. The results show that Sr and Ba are released at relatively high concentrations at 400°C and 500°C.
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Assessment of Candidate Fuel Cladding Alloys for the Canadian Supercritical Water-Cooled Reactor Concept
Journal of Nuclear Engineering and Radiation Science, 2015Co-Authors: David Guzonas, M. Edwards, Wenyue ZhengAbstract:Selecting and qualifying a fuel cladding material for the Canadian supercritical Water-Cooled Reactor (SCWR) concept remains the most significant materials challenge to be overcome. The peak cladding temperature in the Canadian SCWR concept is predicted to be as high as 800°C. While advanced materials show promise for future deployment, currently, the best options available are austenitic stainless steels and nickel-based alloys. Many of these alloys were extensively studied for use as fuel cladding materials in the 1960s, as part of programs to develop nuclear superheated steam Reactors. After extensive out-of-pile testing and consideration of the existing data, five alloys (347 SS, 310 SS, Alloy 800H, Alloy 625, and Alloy 214) were selected for more detailed assessment using a combination of literature surveys and targeted testing to fill in major knowledge gaps. Wherever possible, performance criteria were developed for key materials properties. This paper summarizes the methodology used for the assessment and presents the key results, which show that 310 SS, Alloy 800H, and Alloy 625 would all be expected to give acceptable performance in the Canadian SCWR concept.
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Supercritical Water-Cooled Reactor materials – Summary of research and open issues
Progress in Nuclear Energy, 2014Co-Authors: David Guzonas, R. NovotnyAbstract:Abstract The Supercritical Water Reactor (SCWR) is one of the six Reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). Research on materials and chemistry for supercritical Water-Cooled Reactors dates back to the 1960s when a number of Reactor concepts using Water at supercritical temperatures but sub-critical pressures (nuclear steam) were studied. There is also significant experience available from the operation of supercritical fossil-fired power plants. In this paper, the materials requirements of the various SCWR concepts are introduced, with a focus on the European Union pressure vessel concept and the Canadian pressure tube concept. The current understanding of the key materials degradation issues is reviewed, and knowledge gaps identified.
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supercritical Water Cooled Reactor materials summary of research and open issues
Progress in Nuclear Energy, 2014Co-Authors: David Guzonas, R. NovotnyAbstract:Abstract The Supercritical Water Reactor (SCWR) is one of the six Reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). Research on materials and chemistry for supercritical Water-Cooled Reactors dates back to the 1960s when a number of Reactor concepts using Water at supercritical temperatures but sub-critical pressures (nuclear steam) were studied. There is also significant experience available from the operation of supercritical fossil-fired power plants. In this paper, the materials requirements of the various SCWR concepts are introduced, with a focus on the European Union pressure vessel concept and the Canadian pressure tube concept. The current understanding of the key materials degradation issues is reviewed, and knowledge gaps identified.
Seiichi Koshizuka - One of the best experts on this subject based on the ideXlab platform.
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Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor
2002Co-Authors: Seiichi Koshizuka, Yuki Ishiwatari, Akifumi YamajiAbstract:The paper describes elements of design consideration of supercritical-pressure, light Water Cooled Reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the Reactors are 1. Compact Reactor and turbines due to high specific enthalpy of supercritical Water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as Reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest Reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water Cooled Reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast Reactor core due to small coolant flow rate, potentially easy shift to fast breeder Reactor without changing coolant technology. 10.Potential ofmore » producing energy products such as hydrogen and high quality hydro carbons. (authors)« less
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supercritical pressure once through cycle light Water Cooled Reactor concept
Journal of Nuclear Science and Technology, 2001Co-Authors: Seiichi KoshizukaAbstract:The purpose of the study is to develop new Reactor concepts for the innovation of light Water Reactors (LWR) and fast Reactors. Concept of the once-through coolant cycle, supercritical-pressure light Water Cooled Reactor was developed. Major aspects of Reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast Reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the Reactor is monitoring coolant flow rate instead of Water level of LWR. The Reactor system is compact and simple because of high specific enthalpy of supercritical Water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the hig...
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control of a fast Reactor Cooled by supercritical light Water
Nuclear Technology, 1998Co-Authors: Tohru Nakatsuka, Yoshiaki Oka, Seiichi KoshizukaAbstract:The plant system of a supercritical-Water-Cooled Reactor is the once-through direct-cycle type, where steam-Water separators and coolant recirculation systems are not necessay. It is different from those of a boiling Water Reactor (BWR) and a pressurized Water Reactor. The supercritical-Water-Cooled Reactor is sensitive to perturbations of the feedWater flow rate because all of the core coolant, driven by the feed Water pumps, flows to the turbines without recirculating core flow. The axial coolant density change is three times larger than that of a BWR. It is necessary to analyze the controllability of the Reactor against coolant flow and pressure perturbations to assess the technical feasibility of the Reactor. The behaviors of a fast Reactor Cooled by supercritical Water are analyzed for three principal perturbations; change of the control rod position, the feedWater flow rate, and the turbine control valve opening. Based on the step responses to the perturbations, the Reactor control system is designed such that the pressure is controlled by the turbine control valves, the main steam temperature is controlled by the feedWater flow rate, and the core power is control the pressure by the feedWater flow rate like in a supercritical fossil-fired power plant because of the nuclear thermal-hydraulic coupling. Parameters of the control system are selected by the test calculations to satisfy both fast convergence and stability criteria. Reactor behaviors with the designed control system are stable against the perturbations, although because the plant is the once-through direct-cycle type, the coolant inventory is small. Reactors Cooled by supercritical light Water are controllable with the described control system.
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Safety of a light Water Cooled Reactor operating at supercritical pressure
1997Co-Authors: Seiichi KoshizukaAbstract:The concept of supercritical Water Cooled Reactors is developed for the innovation toward cost reduction. Design and safety features of the Reactor and the plant system are described. The Reactor coolant system is the once-through direct-cycle. It is different from those of BWR and PWR. It is the simplest, but no natural circulation is established when main feedWater pumps are stopped. Safety system was designed based on the transient and accident analyses. Turbine driven auxiliary feedWater systems are provided for fast core cooling at LOSP. This reduces the capacity of emergency diesel generators in spite of the high system pressure. Accumulators are required for cooling the tight lattice core at LOCA. Safety of the Reactor is evaluated by simplified PSA. The core damage frequency is maintained as the same level of Japanese conventional BWR due to the diversity of feedWater systems in the direct-cycle Reactors. 13 refs., 8 figs., 6 tabs.
R. Novotny - One of the best experts on this subject based on the ideXlab platform.
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Material research for the supercritical Water-Cooled Reactor—summary and open issues
Nuclear Corrosion, 2020Co-Authors: R. Novotny, David GuzonasAbstract:Abstract The supercritical Water-Cooled Reactor (SCWR) is one of the six Reactor concepts included under the umbrella of the Generation IV International Forum. Research on materials and chemistry for SCWRs dates back to the 1960s. A number of Reactor concepts using Water at supercritical temperature but subcritical pressures (nuclear steam) were studied. Significant experience was also obtained from the operation of supercritical fossil-fired power plants. In this chapter, the material requirements of the various SCWR concepts are introduced, with a focus on the European Union’s pressure vessel concept and the Canadian’s pressure tube concept. Furthermore, the key material degradation mechanisms relevant for SCWRs are summarized, state-of-the-art understanding of the key materials degradation issues is reviewed, and knowledge gaps are identified.
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Supercritical Water-Cooled Reactor materials – Summary of research and open issues
Progress in Nuclear Energy, 2014Co-Authors: David Guzonas, R. NovotnyAbstract:Abstract The Supercritical Water Reactor (SCWR) is one of the six Reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). Research on materials and chemistry for supercritical Water-Cooled Reactors dates back to the 1960s when a number of Reactor concepts using Water at supercritical temperatures but sub-critical pressures (nuclear steam) were studied. There is also significant experience available from the operation of supercritical fossil-fired power plants. In this paper, the materials requirements of the various SCWR concepts are introduced, with a focus on the European Union pressure vessel concept and the Canadian pressure tube concept. The current understanding of the key materials degradation issues is reviewed, and knowledge gaps identified.
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supercritical Water Cooled Reactor materials summary of research and open issues
Progress in Nuclear Energy, 2014Co-Authors: David Guzonas, R. NovotnyAbstract:Abstract The Supercritical Water Reactor (SCWR) is one of the six Reactor concepts being investigated under the framework of the Generation IV International Forum (GIF). Research on materials and chemistry for supercritical Water-Cooled Reactors dates back to the 1960s when a number of Reactor concepts using Water at supercritical temperatures but sub-critical pressures (nuclear steam) were studied. There is also significant experience available from the operation of supercritical fossil-fired power plants. In this paper, the materials requirements of the various SCWR concepts are introduced, with a focus on the European Union pressure vessel concept and the Canadian pressure tube concept. The current understanding of the key materials degradation issues is reviewed, and knowledge gaps identified.
Wargha Peiman - One of the best experts on this subject based on the ideXlab platform.
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Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel Supercritical Water-Cooled Reactor
Journal of Nuclear Engineering and Radiation Science, 2015Co-Authors: Marija Miletic, Wargha Peiman, Amjad Farah, Jeffrey Samuel, Alexey DragunovAbstract: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.
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Study on Neutronics and Thermalhydraulics Characteristics of 1200-MWel Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: Marija Miletic, Wargha Peiman, Amjad Farah, Jeffrey Samuel, Alexey DragunovAbstract: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.Copyright © 2014 by ASME
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Pressure Drop Analysis of a Re-Entrant Fuel Channel in a Pressure-Channel Type SuperCritical Water-Cooled Reactor
Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition, 2014Co-Authors: S. Maghsoudi, Wargha Peiman, Igor Pioro, K. GabrielAbstract:Pressure drop calculation and temperature profiles associated with fuel and sheath are important aspects of a nuclear Reactor design. The main objective of this paper is to determine the pressure drop in a fuel channel of a SuperCritical Water-Cooled Reactor (SCWR). One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions. The total pressure drop due to all these parameters was between 108 and 121 kPa.
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Pressure Drop Analysis of a Pressure-Channel Type SuperCritical Water-Cooled Reactor
Volume 15: Safety Reliability and Risk; Virtual Podium (Posters), 2013Co-Authors: Alexey Dragunov, Wargha PeimanAbstract:Pressure drop calculation and temperature profiles associated with fuel and sheath are important aspects of a nuclear Reactor design. The main objective of this paper is to determine the pressure drop in a fuel channel of a SuperCritical Water-Cooled Reactor (SCWR) and to calculate the temperature profile of the sheath and the fuel bundles. One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions.Copyright © 2013 by ASME
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Power Distribution in a Pressure-Channel SuperCritical Water-Cooled Reactor (SCWR)
Volume 6: Beyond Design Basis Events; Student Paper Competition, 2013Co-Authors: Wargha Peiman, Igor Pioro, Kamiel GabrielAbstract:SuperCritical Water-Cooled nuclear Reactor (SCWR) is one of the six nuclear-Reactor concepts being developed under the Generation IV International Forum (GIF) initiative. A generic 1200-MWel pressure-channel SCWR operates at a pressure of 25 MPa with coolant inlet and outlet temperatures of 350°C and 625°C, respectively. High coolant outlet temperature allows for high thermal efficiencies within the range of 45–50%. On the other hand, the high operating temperature of SCWR in turn results in high fuel centerline and sheath temperatures. Hence, it is necessary to determine a power distribution inside a core of a Reactor in order to ensure that a fuel and a fuel-bundle design comply with their corresponding temperature limits.The main objective of this paper is to determine a power distribution inside the core of a generic SCWR by using a lattice code DRAGON and a diffusion code DONJON. As a result of these calculations, heat-flux profiles in all fuel channels were determined. Consequently, the heat-flux profile in a channel with the maximum thermal power was used as an input into a thermalhydraulic code, which was developed in MATLAB in order to calculate a fuel centerline temperature of UO2 and UC nuclear fuels and a sheath temperature of a new fuel-bundle design. Results of this analysis showed that the fuel centerline temperature of the UC fuel was significantly lower than that of the UO2. This paper also proposes four energy groups for further neutronic studies related to SCWRs.Copyright © 2013 by ASME
Marc A Rosen - One of the best experts on this subject based on the ideXlab platform.
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performance analysis of a supercritical Water Cooled nuclear Reactor integrated with a combined cycle a cu cl thermochemical cycle and a hydrogen compression system
Applied Energy, 2017Co-Authors: Maan Alzareer, Ibrahim Dincer, Marc A RosenAbstract:Abstract A novel integration is proposed and analyzed of a thermochemical Water decomposition cycle with a supercritical Water-Cooled nuclear Reactor, a combined cycle, and a hydrogen compression system. The supercritical Water-Cooled Reactor in the integrated system has been investigated extensively in Canada. The integrated system uses a compression system to compress the product hydrogen. The hydrogen is produced via a hybrid thermochemical and electrical Water decomposition cycle that utilizes the chemical couple of copper and chlorine. The integrated system is modeled and simulated on Aspen Plus, except for the steam circuit, which is simulated on Aspen Hysys. The hydrogen production rate from the proposed system is 3.56 kg/s. Both energy and exergy analyses are performed of the integrated system, and its overall energy and exergy efficiencies are, in this regard, found to be 16.9% and 27.8%, respectively.