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Author: Raymond Michael Fanning
Publisher:
ISBN:
Category :
Languages : en
Pages : 38
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Book Description
"Next generation nuclear power plants, specifically small modular reactor designs, are the best alternative to fossil fuels for power generation due to their power density and low carbon emissions and constant awareness of safety concerns. A promising safety feature of new designs is the removal of heat by passive systems in accident scenarios. The passive systems require no moving parts and no intervention by personnel. These systems must be accurately simulated for better understanding of the heat transport phenomena: natural convection cooling. Due to the fact that most work developing these passive heat removal systems are proprietary information, a passive heat removal system for a small modular reactor was designed and simulated in Star CCM+ to evaluate the capability of natural convective flows to remove decay heat in a shutdown scenario. The size and dimensions of the heat exchanger are based on the Westinghouse-SMR design. The design of the passive heat removal system was a hexagonal lattice heat exchanger. The final design was projected to dissipate the 56MW of decay heat at the rate simulated in Star CCM+"--Abstract, page iv.
Author: Raymond Michael Fanning
Publisher:
ISBN:
Category :
Languages : en
Pages : 38
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Book Description
"Next generation nuclear power plants, specifically small modular reactor designs, are the best alternative to fossil fuels for power generation due to their power density and low carbon emissions and constant awareness of safety concerns. A promising safety feature of new designs is the removal of heat by passive systems in accident scenarios. The passive systems require no moving parts and no intervention by personnel. These systems must be accurately simulated for better understanding of the heat transport phenomena: natural convection cooling. Due to the fact that most work developing these passive heat removal systems are proprietary information, a passive heat removal system for a small modular reactor was designed and simulated in Star CCM+ to evaluate the capability of natural convective flows to remove decay heat in a shutdown scenario. The size and dimensions of the heat exchanger are based on the Westinghouse-SMR design. The design of the passive heat removal system was a hexagonal lattice heat exchanger. The final design was projected to dissipate the 56MW of decay heat at the rate simulated in Star CCM+"--Abstract, page iv.
Author: Yugao Ma
Publisher: Frontiers Media SA
ISBN: 2832547036
Category : Technology & Engineering
Languages : en
Pages : 133
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Book Description
Future space missions and deep-sea explorations will require small/micro nuclear reactors (kWe~MWe) for power generation. Compared with conventional energy systems such as storage batteries and fossil energy, nuclear reactors are featured higher energy intensity, higher reliability, and longer lifetime. According to the coolant, the candidate small/micro nuclear reactors include the heat pipe cooled reactor, liquid metal cooled reactor, and gas-cooled reactor, most of which are still in the conceptual design stage with numerical studies and experimental research. These emerging reactors have an entirely different core structure and working principle from the existing light water reactors, which has led to an increasing need for updated simulation methods and experimental studies.
Author: Joshua J. Whitman
Publisher:
ISBN:
Category :
Languages : en
Pages : 78
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Book Description
The lead-cooled flexible conversion ratio fast reactor shows many benefits over other fast-reactor designs; however, the higher power rating and denser primary coolant present difficulties for the design of a passive decay heat removal system. In order to achieve passive cooling, enhancements are needed over current designs, such as the S-PRISM and ABR, which utilize passive cooling through the reactor vessel to atmospheric air. Enhancements such as axial fins, a perforated plate, and round indentations, or dimples, were considered as additions to the hot air riser to increase heat transfer. Other enhancements include a liquid metal bond between the reactor and guard vessels, and a dual-level design which introduces ambient temperature air halfway up the vessel wall. A code was written in Java to simulate these conditions, leading to a promising case using dimples on the guard vessel wall as the primary mode of heat transfer enhancement, and including the dual-level design. A conservative estimate of dimple performance indicates that during a passive decay heat removal shutdown, bulk primary coolant temperature will peak at 713 'C, giving a 12 OC margin to clad failure. Attempts were made to refine the uncertainty within the calculations using a computational fluid dynamics code, Fluent, but these ultimately were unsuccessful. Additional studies were conducted on the static stress imparted on the vessel, and the dynamic stress caused by a seismic event. The static stress was found to be within ASME code limits. Seismic analysis determined that a seismic isolation scheme would be necessary in order to prevent damage to the vessel during an earthquake.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
Using a naturally circulating air stream to remove shutdown decay heat from a nuclear reactor vessel is a key feature of advanced liquid metal reactor (LMR) concepts developed by potential vendors selected by the Department of Energy. General Electric and Rockwell International continue to develop innovative design concepts aimed at improving safety, lowering plant costs, simplifying plant operation, reducing construction times, and most of all, enhancing plant licensability. The reactor program at Argonne National Laboratory (ANL) provides technical support to both organizations. The method of shutdown heat removal proposed employs a totally passive cooling system that rejects heat from the reactor by radiation and natural convection to air. The system is inherently reliable since it is not subject failure modes associated with active decay cooling systems. The system is designed to assure adequate cooling of the reactor under abnormal operating conditions associated with loss of heat removal through other heat transport paths.
Author: J. B. Heineman
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 172
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Book Description
Author: Angelo Frisani
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
The design of passive heat removal systems is one of the main concerns for the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The Reactor Cavity Cooling System (RCCS) is an important heat removal system in case of accidents. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The commercial Computational Fluid Dynamics (CFD) STAR-CCM+/ V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. Two models were developed to analyze heat exchange in the RCCS. Both models incorporate a 180 degree section resembling the VHTR RCCS bench table test facility performed at Texas A & M University. All the key features of the experimental facility were taken into account during the numerical simulations. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls temperature below design limits. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The models considered included the first-moment closure one equation Spalart-Allmaras model, the first-moment closure two-equation k-e and k-w models and the second-moment closure Reynolds Stress Transport (RST) model. For the near wall treatments, the low y+ and the all y+ wall treatments were considered. The two-layer model was also used to investigate the effect of near-wall treatment. The comparison of the experimental data with the simulations showed a satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. The tested turbulence models demonstrated that the Realizable k-e model with two-layer all y+ wall treatment performs better than the other k-e models for such a complicated geometry and flow conditions. Results are in satisfactory agreement with the RST simulations and experimental data available. A scaling analysis was developed to address the distortion introduced by the experimental facility and CFD model in simulating the physics inside the RCCS system with respect to the real plant configuration. The scaling analysis demonstrated that both the experimental facility and CFD model give a satisfactory reproduction of the main flow characteristics inside the RCCS cavity region, with convection and radiation heat exchange phenomena being properly scaled from the real plant to the model analyzed.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) has investigated the heat transfer performance of the GE/PRISM passive design. This initial series of experiments simulates the air-side geometry of the PRISM Radiant Reactor Vessel Auxiliary Cooling System (RVACS). The NSTF operates in either a uniform heat flux mode and a uniform temperature mode at the air/guard vessel interface. Analysis of the RVACS performance data indicates excellent agreement with pretest analytical predictions. Correlation analysis presents the heat transfer data in a form suitable for use in LMR design and verification of analytical studies.
Author: Nicolas Zweibaum
Publisher:
ISBN:
Category :
Languages : en
Pages : 231
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Book Description
The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). The capability to validate integral transient response models is a key issue for licensing new reactor designs. This dissertation presents the scaling strategy, design and fabrication aspects, and startup testing results from the Compact Integral Effects Test (CIET) facility at UCB, which reproduces the thermal hydraulic response of an FHR under forced and natural circulation operation. CIET provides validation data to confirm the performance of the direct reactor auxiliary cooling system (DRACS) in an FHR, used for natural-circulation-driven decay heat removal, under a set of reference licensing basis events, as predicted by best-estimate codes such as RELAP5-3D. CIET uses a simulant fluid, Dowtherm A oil, which at relatively low temperatures (50-120°C) matches the Prandtl, Reynolds, Froude and Grashof numbers of the major liquid salts simultaneously, at approximately 50% geometric scale and heater power under 2% of prototypical conditions. The studies reported here include isothermal pressure drop tests performed during startup testing of CIET, with extensive pressure data collection to determine friction losses in the system, as well as subsequent heated tests, from parasitic heat loss tests to more complex feedback control tests and natural circulation experiments. For initial code validation, coupled steady-state single-phase natural circulation loops and simple forced cooling transients were conducted in CIET. For various heat input levels and temperature boundary conditions, fluid mass flow rates and temperatures were compared between RELAP5- 3D results, analytical solutions when available, and experimental data. This study shows that RELAP5-3D provides excellent predictions of steady-state natural circulation and simple transient forced cooling in CIET. The code predicts natural circulation mass flow rates within 8%, and steady-state and transient fluid temperatures, under both natural and forced circulation, within 2°C of experimental data, suggesting that RELAP5-3D is a good EM to use to design and license FHRs. A key element in design and licensing of new reactor technology lies in the analysis of the plant response to a variety of potential transients. When applicable, this involves understanding of passive safety system behavior. This dissertation develops a framework to assess reliability and propose design optimization and risk mitigation strategies associated with passive decay heat removal systems, applied to the Mk1 PB-FHR DRACS. This investigation builds upon previous detailed design work for Mk1 components and the use of RELAP5-3D models validated for FHR natural circulation phenomenology. For risk assessment, reliability of the point design of the passive safety system for the Mk1 PB-FHR, which depends on the ability of various structures to fulfill their safety functions, is studied. Whereas traditional probabilistic risk assessment (PRA) methods are based on event and fault trees for components of the system that perform in a binary way - operating or not operating -, this study is mostly based on probability distributions of heat load compared to the capacity of the system to remove heat, as recommended by the reliability methods for passive safety functions (RMPS) that are used here. To reduce computational time, the use of response surfaces to describe the system in a simplified manner, in the context of RMPS, is also demonstrated. The design optimization and risk mitigation part proposes a framework to study the elements of the design of the reactor, and more specifically its passive safety cooling system, which can contribute to enhanced reliability of heat removal under accident conditions. Risk mitigation measures based on design, startup testing, in-service inspection and online monitoring are proposed to narrow probability distributions of key parameters of the system and increase reliability and safety. Another major aspect in the development of novel energy systems is the assessment of their impacts on the environment compared to current technologies. While most existing life cycle assessment (LCA) studies have been applied to conventional nuclear power plants, this dissertation proposes a framework to extend such studies to advanced reactor designs, using the example of the Mk1 PB-FHR. The Mk1 uses a nuclear air-Brayton combined cycle designed to produce 100 MWe of base-load electricity when operated with only nuclear heat, and 242 MWe using natural gas co-firing for peaking power. The Mk1 design provides a basis for quantities and costs of major classes of materials involved in building the reactor and fabricating fuel, and operation parameters. Existing data and economic input-output LCA models are used to calculate greenhouse gas emissions per kWh of electricity produced over the life cycle of the reactor. Baseline life cycle emissions from the Mk1 PB-FHR in base-load configuration are 26% lower than average Generation II light water reactors in the U.S., 98% lower than average U.S. coal plants and 96% lower than average U.S. natural gas combined cycle plants using the same turbine technology. In peaking configuration, due to its nuclear component and higher thermal efficiency, the Mk1 plant only produces 32% of the emissions of average U.S. gas turbine simple cycle peaking plants. One key contribution to life cycle emissions results from the amount and type of concrete used for reactor construction. This is an incentive to develop innovative construction methods using optimized steel-concrete composite wall modules and new concrete mixes to reduce life cycle emissions from the Mk1 and other advanced reactor designs.
Author: Tian Zhang
Publisher:
ISBN:
Category : Heat
Languages : en
Pages :
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Book Description
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages :
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Book Description
A facility designed and constructed to demonstrate the viability of natural convection passive heat removal systems as a key feature of innovative LMR Shutdown Heat Removal (SHR) systems is in operation at Argonne National Laboratory (ANL). This Natural Convection Shutdown Heat Removal Test Facility (NSTF) is being used to investigate the heat transfer performance of the GE/PRISM and the RI/SAFR passive designs. This paper presents a description of the NSTF, the pretest analysis of the Radiant Reactor Vessel Auxiliary Cooling System (RVACS) in support of the GE/PRISM IFR concept, and experiment results for the RVACS simulation. Preliminary results show excellent agreement with predicted system performance.
