110 research outputs found
Design of an Organic Simplified Nuclear Reactor
No full textNumerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs. Keywords: Advanced Concepts, Graphite Moderated, Nuclear Design, Organic Reactor, Santowax, SM
CIRIS
No full textThesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.Cataloged from PDF version of thesis.Includes bibliographical references (p. 199-202).The International Reactor Innovative and Secure (IRIS) is an advanced medium size, modular integral light water reactor design, rated currently at 1000 MWt. IRIS design has been under development by over 20 organizations from nine countries, led by Westinghouse. IRIS has a standard Westinghouse PWR core, but is an integral reactor, which means the reactor vessel contains all pumps, steam generators, pressurizer and control rod drive mechanisms. This work assesses possible improvements of the plant economics by the allowable power in the IRIS vessel, while maintaining the same or better safety limits, through increasing the power density in the core and heat exchangers. IRIS was designed with 8 Once-Through Helically coiled Steam Generators (OTHSG), located in an annulus near the vessel in the region above the core. The unit size dictates the vessel diameter, or limits the core size for fixed vessel diameter, and thus the reactor power rating. The Printed Circuit Heat Exchangers (PCHE) of HEATRICIm are compact heat exchangers that can provide high power density along with low pressure drop. They are proposed here as replacement for the OTHSG. The PCHEs experience is mostly for single phase heat transfer. A model is developed for the two phase fluid heat transfer in the small horizontal PCHE flow channels. The PCHE performance under IRIS conditions was modeled by a one dimensional nodal code. For the same power output, the PCHEs are found to safely reduce the IRIS vessel diameter by as much as 1.5 m and reduce the pressure drop in the SG by 30 %. The Internally and eXternally cooled Annular Fuel (IXAF) had been investigated as part of MIT's Advanced Fuel Project. It was found to maintain the current operating MDNBR margin under steady state IRIS conditions at 150% of nominal power density when the flow rate can be proportionally increased. The MDNBR in the inner channels was sensitive to flow changing flow conditions. A complete RELAP5 model of the IRIS reactor, along with PCHE and IXAF design representation, was developed. The PCHE RELAP model was first benchmarked against the stand-alone code and their agreement was demonstrated successfully. The short and long term responses of IRIS with PCHE and IXAF were analyzed for a Loss Of Flow Accident (LOFA) and a Loss Of FeedWater Accident (LOFWA). Under LOFA the MDNBR margins were found to be acceptable with added inertia to current IRIS pumps configuration. Therefore, the pressure vessel size can be reduced by implementing the PCHE instead of the OTHSG, and IXAF instead of solid fuel rods in addition to increasing the power rating of the reactor by 50% for the same vessel size. The results indicate that a large potential exist to reduce the cost per kilowatt and increase the attractiveness of the IRIS reactor design.by Koroush Shirvan.S.M
Analysis methods for high power density BWRs
No full textThesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2013.Cataloged from PDF version of thesis.Includes bibliographical references (p. 263-268).Increasing the economic competitiveness of nuclear energy is vital to its future. Improving the economics of BWRs is the main goal of this work, focusing on designing cores with higher power density, to reduce the BWR capital cost. Generally, the core power density in BWRs is limited by the thermal Critical Power of its assemblies, below which heat removal can be accomplished with low fuel and cladding temperatures. The present study investigates both increases in the heat transfer area between the fuel and coolant and changes in operating parameters to achieve higher power levels while meeting the appropriate thermal as well as materials and neutronic constraints. A scoping study is conducted under the constraints of using fuel with cylindrical geometry, traditional materials and enrichments below 5% to enhance its licensability. The reactor vessel diameter is limited to the largest proposed thus far. The BWR with High power Density (BWR-HD) is found to have a power level of 5000 MWth, equivalent to 26% uprated ABWR, resulting into 20% cheaper O&M and Capital costs. This is achieved by utilizing the same number of assemblies, but with wider 16x1 6 assemblies and 50% shorter active fuel than that of the ABWR. The fuel rod diameter and pitch are reduced to just over 45% of the ABWR values. Traditional cruciform form control rods are used, which restricts the assembly span to less than 1.2 times the current GE 14 design due to limitation on shutdown margin. Thus, it is possible to increase the power density and specific power by 65%, while maintaining the nominal ABWR Minimum Critical Power Ratio (MCPR) margin. The optimum core pressure is the same as the current 7.1 MPa. The core exit quality is increased to 19% from the ABWR nominal exit quality of 15%. The pin linear heat generation rate is 20% lower, and the core pressure drop and mass of uranium are 30% lower. The BWR-HD's fuel, modelled with FRAPCON 3.4, showed similar performance to the ABWR pin design. The fuel cycle is only 12 month long, but on the per kWhr, the new design operates with 14% lower fuel cycle front-end costs and similar total fuel cycle cost to the 18 month ABWR fuel cycle. The plant systems outside the vessel are assumed to be the same as the ABWR-1I design, utilizing a combination of active and passive safety systems. Safety analyses applied a void reactivity coefficient calculated by SIMULATE-3 for an equilibrium cycle core that showed a 15% less negative coefficient for the BWR-HD compared to the ABWR. The feedwater temperature was kept the same for the BWR-HD and ABWR which resulted in 4 °K cooler core inlet temperature for the BWR-HD given that its feedwater makes up a larger fraction of total core flow. The stability analysis using the STAB and S3K codes showed satisfactory results for the hot channel, coupled regional out-of-phase and coupled core-wide in-phase modes. A RELAP5 model of the ABWR system was constructed and applied to six transients for the BWR-HD and ABWR. The AMCPRs during all the transients were found to be equal or less for the new design and the core remained covered for both. The lower void coefficient along with smaller core volume proved to be advantages for the simulated transients. Helical Cruciform Fuel (HCF) rods were proposed in prior MIT studies to enhance the fuel surface to volume ratio. In this work, higher fidelity models (e.g. CFD instead of subchannel methods for the hydraulic behaviour) are used to investigate the resolution needed for accurate assessment of the HCF design. For neutronics, conserving the fuel area of cylindrical rods results in a different reactivity level with a lower void coefficient for the HCF design. In single-phase flow, for which experimental results existed, the friction factor is found to be sensitive to HCF geometry and cannot be calculated using current empirical models. A new approach for analysis of flow crisis conditions for HCF rods in the context of Departure from Nucleate Boiling (DNB) and dryout using the two phase interface tracking method was proposed and initial results are presented. It is shown that the twist of the HCF rods promotes detachment of a vapour bubble along the elbows which indicates no possibility for an early DNB for the HCF rods and in fact a potential for a higher DNB heat flux. Under annular flow conditions, it was found that the twist suppressed the liquid film thickness on the HCF rods, at the locations of the highest heat flux, which increases the possibility of reaching early dryout. It was also shown that modeling the 3D heat and stress distribution in the HCF rods is necessary for accurate steady state and transient analyses. The safety analysis of the 20% uprated HCF design in the context of a BWR/4 RPV showed satisfactory AMCHFR performance only if CR is estimated by the EPRI- 1 correlation.by Koroush Shirvan.Ph.D
Development of Cr cold spray–coated fuel cladding with enhanced accident tolerance
No full textAccident-tolerant fuels (ATFs) are currently of high interest to researchers in the nuclear industry and in governmental and international organizations. One widely studied accident-tolerant fuel concept is multilayer cladding (also known as coated cladding). This concept is based on a traditional Zr-based alloy (Zircaloy-4, M5, E110, ZIRLO etc.) serving as a substrate. Different protective materials are applied to the substrate surface by various techniques, thus enhancing the accident tolerance of the fuel. This study focuses on the results of testing of Zircaloy-4 coated with pure chromium metal using the cold spray (CS) technique. In comparison with other deposition methods, e.g., Physical vapor deposition (PVD), laser coating, or Chemical vapor deposition techniques (CVD), the CS technique is more cost efficient due to lower energy consumption and high deposition rates, making it more suitable for industry-scale production. The Cr-coated samples were tested at different conditions (500°C steam, 1200°C steam, and Pressurized water reactor (PWR) pressurization test) and were precharacterized and postcharacterized by various techniques, such as scanning electron microscopy, Energy-dispersive X-ray spectroscopy (EDX), or nanoindentation; results are discussed. Results of the steady-state fuel performance simulations using the Bison code predicted the concept's feasibility. It is concluded that CS Cr coating has high potential benefits but requires further optimization and out-of-pile and in-pile testing. Keywords: Accident-Tolerant Fuel, Chromium, Cladding, Coating, Cold Spray, Nuclear Fue
Assessment of reinforcement learning algorithms for nuclear power plant fuel optimization
No full textThe nuclear fuel loading pattern optimization problem belongs to the class of large-scale combinatorial optimization. It is also characterized by multiple objectives and constraints, which makes it impossible to solve explicitly. Stochastic optimization methodologies including Genetic Algorithms and Simulated Annealing are used by different nuclear utilities and vendors but hand-designed solutions continue to be the prevalent method in the industry. To improve the state-of-the-art, Deep Reinforcement Learning (RL), in particular, Proximal Policy Optimization is leveraged. This work presents a first-of-a-kind approach to utilize deep RL to solve the loading pattern problem and could be leveraged for any engineering design optimization. This paper is also to our knowledge the first to propose a study of the behavior of several hyper-parameters that influence the RL algorithm. The algorithm is highly dependent on multiple factors such as the shape of the objective function derived for the core design that behaves as a fudge factor that affects the stability of the learning. But also an exploration/exploitation trade-off that manifests through different parameters such as the number of loading patterns seen by the agents per episode, the number of samples collected before a policy update , and an entropy factor that increases the randomness of the policy during training. We found that RL must be applied similarly to a Gaussian Process in which the acquisition function is replaced by a parametrized policy. Then, once an initial set of hyper-parameters is found, reducing and until no more learning is observed will result in the highest sample efficiency robustly and stably. This resulted in an economic benefit of 535,000 - 642,000 $/year/plant. Future work must extend this research to multi-objective settings and comparing them to state-of-the-art implementation of stochastic optimization methods
Superheated Water-Cooled Small Modular Underwater Reactor Concept
No full textAbstractA novel fully passive small modular superheated water reactor (SWR) for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF). The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required
Design of an Organic Simplified Nuclear Reactor
No full textAbstractNumerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs
PESA: Prioritized experience replay for parallel hybrid evolutionary and swarm algorithms - Application to nuclear fuel
No full textWe propose a new approach called PESA (Prioritized replay Evolutionary and Swarm Algorithms) combining prioritized replay of reinforcement learning with hybrid evolutionary algorithms. PESA hybridizes different evolutionary and swarm algorithms such as particle swarm optimization, evolution strategies, simulated annealing, and differential evolution, with a modular approach to account for other algorithms. PESA hybridizes three algorithms by storing their solutions in a shared replay memory, then applying prioritized replay to redistribute data between the integral algorithms in frequent form based on their fitness and priority values, which significantly enhances sample diversity and algorithm exploration. Additionally, greedy replay is used implicitly to improve PESA exploitation close to the end of evolution. PESA features in balancing exploration and exploitation during search and the parallel computing result in an agnostic excellent performance over a wide range of experiments and problems presented in this work. PESA also shows very good scalability with number of processors in solving an expensive problem of optimizing nuclear fuel in nuclear power plants. PESA's competitive performance and modularity over all experiments allow it to join the family of evolutionary algorithms as a new hybrid algorithm; unleashing the power of parallel computing for expensive optimization
Achieving Sizable Power-Uprate for Existing Fleet Through LEU+ and ATF
No full textNuclear energy produces more carbon-free electricity than any other source, accounting for ~20% of U.S. electricity generation. With the net-zero emission goals of the upcoming decades, there is a consensus and recognition of the vital importance in continual operation of existing reactor fleet. Over the last 25 years, the U.S. NRC has approved over 130 power-uprates for the existing fleet ranging from ~0.5-20%. Given that the rise of cheap natural gas has challenged the economic viability of adding new nuclear power plants, further exploration on increasing the value of the existing fleet is critical. An additional uprate of 50% to existing nuclear generation will mostly meet the carbon-free electricity needs of the next 10 years. In 2008, MIT and Westinghouse studied the feasibility of a 50% power uprate, and the major barriers to its adoption were the lack of availability for greater than 5% enriched fuel and uncertainty regarding feasible plant lifetimes to motivate the needed refurbishment investments. The advent of LEU+ (5-10% enrichment) and Accident Tolerant Fuels (ATFs) in a carbon-constrained economy as well as approval of an 80-year life by U.S. NRC for selected plants, provides an opportune time for revisiting the extend for an achievable power-uprate. This paper will outline and quantitatively support additional extended power-uprate for a generic PWR and BWR by leveraging LEU+ and ATF based on neutronics, thermal-hydraulics, fuel performance and economic assessments. The maximum allowable uprate is achieved by fuel geometry optimization while constraining the fuel materials to near term options (e.g., UO2/Zr). With adequate refurbishment of plant equipment, an economical 50% power uprate can be possible while shortening the fuel cycle to 12-month. A fleet-wide power uprate of such magnitude will not only enable meeting the U.S. emission-free electricity generation targets by 2030s, but it will also provide a much-needed bridge for advanced reactor deployment timelines and revitalizes U.S. utilities interest to invest in nuclear energy at a scale needed to achieve future net-zero targets
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