Conceptual design of nuclear-geothermal energy storage systems for variable electricity production

Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2011.Cataloged from PDF version of thesis.Includes bibliographical references (p. 170-171).by Youho Lee.S.M

Predicting crack patterns in SiC-based cladding for LWR applications using peridynamics

SiC continuous fibre reinforced SiC matrix (SiC-SiC) composites are a proposed material for accident tolerant fuel cladding. Thermomechanical models of SiC-based cladding under light water conditions indicate that microcracking in the radial direction of the tubing may lead to a loss of hermicity. SiC-based tubing is known to have anisotropic elastic properties but the effect of this anisotropy have not been incorporated into existing thermomechanical models of clad cracking. This work augments an existing isotropic 2D peridynamic model of cracking and damage in the r-θ plane of a SiC-based cladding to account for the orthotropic elastic properties of SiC-SiC composite tubing. Three SiC-based architectures are modelled under normal operating conditions of a UO2-fuelled pressurised water reactor (PWR). The results of the anisotropic SiC-cladding model are compared with the results of the isotropic model, and the sensitivity of results to material anisotropy, thermal conductivity, and applied linear power rating are analysed. The results of this analysis show that anisotropy has a significant effect on the damage and crack patterns observed in the r-θ plane of SiC-based cladding, if either an inner or outer monolith is present. The anisotropic model predicts more cracks in two layer clad with an inner monolith and higher levels of damage in a two layer clad with an outer monolith than the isotropic model. Under normal reactor conditions the outer monolith clad architecture was found to remain hermetic

Safety of light water reactor fuel with silicon carbide cladding

Thesis: Ph. D., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2013.Cataloged from PDF version of thesis.Includes bibliographical references (pages 303-314).Structural aspects of the performance of light water reactor (LWR) fuel rod with triplex silicon carbide (SiC) cladding - an emerging option to replace the zirconium alloy cladding - are assessed. Its behavior under accident conditions is examined with an integrated approach of experiments, modeling, and simulation. High temperature (1100°C~1500°C) steam oxidation experiments demonstrated that the oxidation of monolithic SiC is about three orders of magnitude slower than that of zirconium alloys, and with a weaker impact on mechanical strength. This, along with the presence of the environmental barrier coating around the load carrying intermediate layer of SiC fiber composite, diminishes the importance of oxidation for cladding failure mechanisms. Thermal shock experiments showed strength retention for both [alpha]-SiC and [beta]-SiC, as well as A1₂O₃ samples quenched from temperatures up to 1260°C in saturated water. The initial heat transfer upon the solid - fluid contact in the quenching transient is found to be a controlling factor in the potential for brittle fracture. This implies that SiC would not fail by thermal shock induced fracture during the reflood phase of a loss of coolant accident, which includes fuel-cladding quenching by emergency coolant at saturation conditions. A thermo-mechanical model for stress distribution and Weibull statistical fracture of laminated SiC cladding during normal and accident conditions is developed. It is coupled to fuel rod performance code FRAPCON-3.4 (modified here for SiC) and RELAP-5 (to determine coolant conditions). It is concluded that a PWR fuel rod with SiC cladding can extend the fuel residence time in the core, while keeping the internal pressure level within the safety assurance limit during steady-state and loss of coolant accidents. Peak burnup of 93 MWD/kgU (10% central void in fuel pellets) at 74 months of in-core residence time is found achievable with conventional PWR fuel rod design, but with an extended plenum length (70 cm). An easier to manufacture, 30% larger SiC cladding thickness requires an improved thermal conductivity of the composite layer to reduce thermal stress levels under steady-state operation to avoid failure at the same burnup. A larger Weibull modulus of the SiC cladding improves chances of avoiding brittle failure.by Youho Lee.Ph. D

Application of Deep Belief Network for Critical Heat Flux Prediction on Microstructure Surfaces

Considering the highly nonlinear behavior and phenomenological complexity of critical heat flux (CHF), this study proposes a novel method to predict CHF on microstructure surface using machine learning technologies. An extensive literature survey was conducted to collect experimental data on microstructure surfaces. Data on horizontal silicon specimens of cylindrical pillars with square arrangements were selected for both training and testing various machine learning methods, including nu-support vector machine, back-propagation neural network, radial basis function neural network, general regression neural network, and deep belief network (DBN). Among the tested machine learning methods, DBN is shown to provide the best accuracy for CHF prediction. The obtained parametric CHF behavior of DBN with respect to pillar diameter, spacing, and height agrees with the physical understanding of CHF on microstructure surfaces. The presented approach is expected to support the design optimization of microstructure for CHF maximization.N

Spent fuel simulation during dry storage via enhancement of FRAPCON-4.0: Comparison between PWR and SMR and discharge burnup effect

Spent fuel behavior of dry storage was simulated in a continuous state from steady-state operation by modifying FRAPCON-4.0 to incorporate spent fuel–specific fuel behavior models. Spent fuel behavior of a typical PWR was compared with that of NuScale Power Module (NPM™). Current PWR discharge burnup (60 MWd/kgU) gives a sufficient margin to the hoop stress limit of 90 MPa. Most hydrogen precipitation occurs in the first 50 years of dry storage, thereby no extra phenomenological safety factor is identified for extended dry storage up to 100 years. Regulation for spent fuel management can be significantly alleviated for LWR-based SMRs. Hydride embrittlement safety criterion is irrelevant to NuScale spent fuels; they have sufficiently lower plenum pressure and hydrogen contents compared to those of PWRs. Cladding creep out during dry storage reduces the subchannel area with burnup. The most deformed cladding outer diameter after 100 years of dry storage is found to be 9.64 mm for discharge burnup of 70 MWd/kgU. It may deteriorate heat transfer of dry storage by increasing flow resistance and decreasing the view factor of radiative heat transfer. Self-regulated by decreasing rod internal pressure with opening gap, cladding creep out closely reaches the saturated point after ∼50 years of dry storage

Post-LOCA ductility assessment of Zr-Nb Alloy from 1100 degrees C to 1300 degrees C to explore variable peak cladding temperature and equivalent cladding reacted safety criteria

This study experimentally examines the variable PCT-ECR embrittlement criteria by conducting post-LOCA ductility assessments in the temperature range of 1100~ 130 0 degrees C. Post-LOCA ductility assessments demonstrate the viability of variable PCT-ECR criteria, allowing either higher PCT for less oxidized specimens, or higher ECR limit for lower PCT specimens. The variable PCT-ECR criteria can be useful for high burnup fuel regulation. For potential burnup uprates that may lead to the violation of the current CP-ECR limit (18%) due to excessive steady-state oxidation, it may be possible to circumvent the regulation limit by allowing PCT less than the current limit (1204 degrees C). This represents a new, yet more rational, safety envelop that may liberate nuclear reactor from sweeping conservatism associated with LOCA safety assessments. For the specimens oxidized at temperatures greater than 1200 degrees C, ductile-brittle transitions take place even with substantially thicker prior-beta layer (375 mu m) and lower oxygen concentration compared to 1200 degrees C specimens. As such, premature ductile to brittle transition occurs from the view point of oxygen uptake for steam annealing temperature above 1200 degrees C. Premature embrittlement of 1250 degrees C and 1300 degrees C that cannot be explained in terms of the thickness of prior-beta phase is relevant to promoted grain growth at elevated temperatures. Unlike lower temperature (<= 1200 degrees C), steam annealing temperature of 1250 degrees C and 1300 degrees C made grains continuously grow beyond the average grain size of 50 mu m. The enhanced grain growth is considered to degrade cladding ductility as commonly observed in other HCP materials, in addition to solid-solution hardening of oxygen. (C) 2022 Elsevier B.V. All rights reserved.N

Conceptual Design of Nuclear-Geothermal Energy Storage Systems for Variable Electricity Production

Nuclear plants have high capital costs and low operating costs that favor base-load operation. This characteristic of nuclear power has been a critical constraint that limits the portion of nuclear power plants in a grid to stay below the base-load demand. A novel gigawatt-year thermal-energy storage technology is proposed to enable base load nuclear plants to produce variable electricity to meet seasonal variations in electricity demand. A large volume of underground rock is heated with hot water (or steam or carbon dioxide) from a nuclear power plant during periods of low electricity demand, and the heat is extracted during times of high demand and converted to electricity using a standard geothermal plant (Figure 1). Among various technical options, technically mature ones were selected for the reference design; a Pressurized Water Reactor (PWR) injects hot fluid into an underground reservoir through an intermediate heat exchanger and bypass flow lines on either the primary or secondary side. The reservoir size of 500 m in each dimension at 1.5 km underneath the surface is engineered to have permeability of 2 Darcy using commercial hydraulic fracture methods, and is cyclically heated up and cooled down between the temperatures of 50°C and 250°C. Peak power electricity is produced by exploiting the stored thermal energy via an Enhanced Geothermal System (EGS) that employs a binary flash cycle. Models of a nuclear-EGS system performance, taking into account heat transfer in the reservoir, thermal front velocity in the reservoir, conductive heat & water losses, geothermal power plant electricity production performance, operating conditions and system interfaces were developed and independently compared with Computational Fluid Dynamics (CFD) simulations using FLUENT 6.3 to confirm the validity of the models. The design study with the validated models reveals that the reference nuclear-EGS system based on 2.8~6.0 GW(th) of nuclear power would have a thermal storage size of 0.7~1.5 GW(th)-year, which corresponds to 0.08~0.2 GW(e)-year with electricity round trip efficiency of 0.34~0.46. Reservoir permeability and geofluid temperature are found to be the most important design parameters that affect performance of nuclear-EGS storage systems. A grid that deploys a nuclear-EGS system will have three distinct electricity sectors: nuclear base load, EGS intermediate load, and gas turbine peak power. The nuclear-EGS storage system introduces economic benefits to a grid by leveraging economic gains arising from replacing expensive intermediate and peak electricity with cheap base-load electricity. A nuclear-EGS system has a higher capital cost than natural gas turbines; consequently, it replaces intermediate-load power plants but not all the gas turbines that operate for a small number of hours per year. It was found that the deployment of a operate for a small number of hours per year. It was found that the deployment of a Nuclear-EGS could cut the electricity production cost of the New England Independent Systems Operator (NE-ISO) by as much as 14% of the storage-free cost (Fig. 2). Economic competitiveness of nuclear power plants is the most decisive factor for the deployment of the system in a grid. Because this was the first analysis of a nuclear EGS system, we used off-the-shelf technology wherever possible to reduce uncertainties and have confidence that the system will work. Significant improvements in roundtrip efficiency and economics may be possible by development of more advanced systems. For example, existing geothermal power plants are small (megawatts) versus several hundred megawatts for a nuclear EGS system. They use double flash power systems. The larger scale may enable the use of triple-flash and other more efficient power cycles. Reservoir development methods designed explicitly for nuclear EGS systems may significantly lower the costs of reservoir development. Like any other system dependent upon geology, costs and performance will depend upon the local geology.Idaho National Laboratory (Hybrid Systems for Process Integration and Dynamic Studies)Korea Institute of Energy Technology Evaluation and Planning (fellowship

Heat transfer foot print on ceramics after thermal shock with droplet impingement: Development of thermal shock tolerant material with hydrophobic surface

We perform a systematic study of the thermal shock experienced by the alumina during quenching by cold water droplet impingement with heated surface temperature ranging from 125°C to 475°C for Weber number ≈32. We explore the effect of surface heat transfer mode on the thermal shock experienced by the material. It is found that the variation of residual strength translates into the mode of boiling heat transfer, hence surface heat flux. The material remembers the degree of thermal shock; the heat transfer foot print is embedded in the residual strength. This finding speaks to a possibility of developing a ceramic detector for heat transfer modes in extreme environments. This study finds that superior thermal shock tolerance can be achieved by removing the heat transfer footprint with reduced heat flux. By promoting the film boiling with nano-fractal hydrophobic surface, we achieved superior thermal shock tolerance for alumina substrates. This is a novel approach to reduce thermal shock by controlling the heat transfer with surface modification, different from conventional, yet expensive, method of improving the bulk material properties

Mechanical analysis of surface-coated zircaloy cladding

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness

Flow Accelerated Corrosion of Stainless Steel 316L by a Rotating Disk in Lead-Bismuth Eutectic Melt

Stainless steel 316L specimens were snug fitted into a rotating disk submerged in molten LBE and subjected to spatially varying local flow velocity along the radial position. The specimens experienced LBE flow velocity from 0.50 m/s to 3.14 m/s depending on their radial location. The test was conducted at 600 degrees C with an oxygen concentration of 2.87 x 10(-8) wt% for 150 h. Resulting microstructural characteristics of the corroded zone were found to be sensitively affected by local flow velocity and were categorized into four regimes. For linear disk velocity > 2.0 m/s, the affected zone thickness became increasingly less sensitive to flow velocity as the overall reaction became reaction rate controlled. At the velocity of similar to 3.0 m/s, erosion-corrosion started to take place. The flow effect on the affected zone thickness (l) agreed with the model based on the disk velocity (v) effect on the mass transfer of a rotating disk as 1/l similar to v(-0.792)