Search NASA⌕ Search

SEARCH · Search NASA

Results for “Small Modular”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios

The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios and to be significantly impacted by the varying parameters. Furthermore, analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

A compact 10 kW, 476 MHz solid state radio frequency amplifier for pre-buncher cavity of free electron laser injector linear accelerator

A 10 kW, 476 MHz, 0.1% duty cycle solid state RF amplifier system for driving sub-harmonic, pre-buncher cavity of IR-FEL injector LINAC, has been developed at RRCAT. The 10 kW power is achieved by combining output of eight 1400 W amplifier modules using 8-way planar corporate combiner. The solid state amplifier modules have been developed using 50 V RF LDMOS transistors which although meant for push-pull operation are being used in single ended configuration with matching circuit developed on a thin (25 mils), high dielectric constant (9.7), low loss microwave laminate with an aim to have a compact structure. Ease of fabrication, modularity, small size, and low cost are the important features of this design which could be used as a template for low duty cycle medium to high pulsed power UHF amplifier system.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

Stand-alone polarization-modulation infrared reflection absorption spectroscopy instrument optimized for the study of catalytic processes at elevated pressures

This article describes the design and construction of a compact, “user-friendly” polarization-modulation infrared reflection absorption spectroscopy (PM-IRRAS) instrument at the Center for Functional Nanomaterials (CFN) of Brookhaven National Laboratory, which allows studying surfaces at pressures ranging from ultra-high vacuum to 100 Torr. Surface infrared spectroscopy is ideally suited for studying these processes as the vibrational frequencies of the IR chromophores are sensitive to the nature of the bonding environment on the surface. Relying on the surface selection rules, by modulating the polarization of incident light, it is possible to separate the contributions from the isotropic gas or solution phase, from the surface bound species. A spectral frequency range between 1000 cm -1 and 4000 cm -1 can be acquired. While typical spectra with a good signal to noise ratio can be obtained at elevated pressures of gases in ~2 min at 4 cm -1 resolution, we have also acquired higher resolution spectra at 0.25 cm -1 with longer acquisition times. By way of verification, CO uptake on a heavily oxidized Ru(0001) sample was studied. As part of this test study, the presence of CO adsorbed on Ru bridge sites was confirmed, in agreement with previous ambient pressure X ray photoelectron spectroscopy studies. In terms of instrument performance, it was also determined that the gas phase contribution from CO could be completely removed even up to pressures close to 100 Torr. A second test study demonstrated the use of the technique for studying morphological properties of a spin coated polymer on a conductive surface. Note that this is a novel application of this technique. In this experiment, the polarization of incident light was modulated manually (vs. through a photoelastic modulator). It was demonstrated, in good agreement with the literature, that the polymer chains preferentially lie parallel with the surface. This PM-IRRAS system is small, modular, and easily reconfigurable. It also features a “vacuum suitcase” that allows for the integration of the PM-IRRAS system with the rest of the suite of instrumentation at our laboratory available to external users through the CFN user proposal system.

36 MATERIALS SCIENCE↗
DOE OSTI2018

A roadmap for nuclear energy technology

The prospects for the future use of nuclear energy worldwide can best be understood within the context of global population growth, urbanization, rising energy need and associated pollution concerns. As the world continues to urbanize, sustainable development challenges are expected to be concentrated in cities of the lower-middle-income countries where the pace of urbanization is fastest. As these countries continue their trajectory of economic development, their energy need will also outpace their population growth adding to the increased demand for electricity. OECD IEA’s energy system deployment pathway foresees doubling of the current global nuclear capacity by 2050 to reduce the impact of rapid urbanization. The pending “retirement cliff” of the existing U.S. nuclear fleet, representing over 60 percent of the nation’s emission-free electricity, also poses a large economic and environmental challenge. To meet the challenge, the U.S. DOE has developed the vision and strategy for development and deployment of advanced reactors. As part of that vision, the U.S. government pursues programs that aim to expand the use of nuclear power by supporting sustainability of the existing nuclear fleet, deploying new water-cooled large and small modular reactors to enable nuclear energy to help meet the energy security and climate change goals, conducting R&D for advanced reactor technologies with alternative coolants, and developing sustainable nuclear fuel cycle strategies. Since the current path relying heavily on water-cooled reactors and “once-through” fuel cycle is not sustainable, next generation nuclear energy systems under consideration aim for significant advances over existing and evolutionary water-cooled reactors. Among the spectrum of advanced reactor options, closed-fuel-cycle systems using reactors with fast-neutron spectrum to meet the sustainability goals offer the most attractive alternatives. However, given these circumstances, unless the new public-private partnership models emerge to tackle the licensing and demonstration challenges for these advanced reactor concepts, realization of their enormous potential is not likely, at least in the U.S.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2023

Modeling and control of nuclear–renewable integrated energy systems: Dynamic system model for green electricity and hydrogen production

The need for decarbonization and diversification of energy resources has led to the development of integrated energy systems (IESs), where multiple resources supply more than one energy sector. Here, one such IES with small modular nuclear reactors and renewables (wind and solar) as generating resources, catering to the demand of the electric grid while producing hydrogen for industries, is modeled in this paper. The physics-based component models are represented using the Modelica language and interconnected to form the IES. The control and coordination of the overall system are ensured by designing a suitable control architecture composed of individual subsystem-level controls and supervisory control. The dynamic performance and the load-following capability of the IES are evaluated, while satisfying the safe operational limits of the components. Different configurations and modes of IES operation are considered, where the adaptability of the control system in the presence of varying demands and renewable generations is validated. The simulation results indicate that hydrogen as a flexible load facilitates the supply of varying grid demand. Additionally, the renewables are also accommodated into the IES owing to the flexibility of the balance of plant associated with the nuclear reactors.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2019

Numerical Simulation of Isothermal Flow Across Slant Five-Tube Bundle with Spectral Element Method Code Nek5000

The helical coil steam generator (HCSG) is a specific type of shell-and-tube heat exchanger known for having a higher heat transfer coefficient than many other designs. For this reason, they are considered in small modular reactor and high temperature reactor designs. Investigation of flow behavior in HCSGs is important for better design. In this paper we present our study for modeling the coolant flow in the primary side of the HCSG. We used Nek5000, an open source, high-order spectral element computational fluid dynamics code developed in Argonne National Laboratory. Nek5000 accepts only hexahedral mesh, which makes the meshing process for the complicated HCSG geometry very challenging. A tetrahedral-to-hexahedral meshing strategy was applied to bypass the geometric complexities. In this study large eddy simulation (LES) was performed at the Reynolds number of 9000 based on the inlet velocity and the tube diameter. The employed subgrid-scale model for LES relies on explicit filtering. First- and second-order statistics were compared to available experimental data. We report overall velocity and turbulent kinetic energy showed good agreement with particle image velocimetry data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Simulation of Boiling Two-Phase Flow in a Helical Coil Steam Generator Using the Spectral Element Code Nek-2P

To enable the design of a light water small modular reactor, the boiling flow inside a helical coil steam generator has been simulated with the two-fluid model in Nek-2P. Nek-2P is the multiphase branch of the spectral element code Nek5000. Details of the implementation of the two-fluid model and the included closure models are discussed. The presented closure models include interactions for momentum, heat, and mass transfer between phases. Models for the drag, lift, and turbulent dispersion forces are included. The complete model is fully consistent in the limits of both phasic volume fractions approaching zero and is able to simulate flows of dispersed vapor, continuous liquid, dispersed liquid, continuous vapor, or any combination thereof. The closure models and their implementation in Nek-2P have been validated by comparing to experimental data for a boiling flow, demonstrating excellent agreement. Results from the simulation of the helical coil indicate strong phasic separation driven by the effects of buoyancy and inertia. Finally, significant differences were observed in the results compared to simulations performed using Star-CCM+, although these differences were somewhat expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Severe Accident Phenomena: A Comparison Among the NuScale SMR, Other Advanced LWR Designs, and Operating LWRs

During a severe accident in a nuclear reactor, there are a number of phenomenological events which can present a challenge to containment integrity. These include the generation and combustion of hydrogen, energetic fuel-coolant interactions, thermal attack of fission product barriers, core-concrete interactions, direct containment heating, and gradual overpressurization. The advanced design of the NuScale small modular reactor has resulted in the reduced likelihood and severity of severe accident challenges to containment. This paper discusses the features of the NuScale design that reduce the likelihood of occurrence of these severe accident phenomena, and discusses the ability of containment to survive in the unlikely event that they do occur. Here, the impact of severe accident phenomena for the NuScale design is compared and contrasted against other advanced light water reactors, such as AP1000 and ESBWR, as well as the existing fleet, using information from publicly available documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Impact of Radial Reflector Fidelity on Neutronics and Vessel Fluence Simulations

The Consortium for Advanced Simulation of Light Water Reactors is developing the Virtual Environment for Reactor Applications (VERA), and the MPACT code, which is the primary deterministic neutron transport solver in VERA, provides sub-pin level flux and power distributions as part of full-scale cycle depletion and analysis. In such calculations, an important aspect is the radial reflector treatment. To improve the fidelity of the radial reflector treatment, MPACT was extended to approximate the modeling of the reactor’s structural components such as the core shroud, barrel, neutron pads, and vessel. This work explores several modeling configurations with varying levels of fidelity and computational burden and assesses the importance of modeling fidelity on the eigenvalue and pin power distribution. Two two-dimensional (2-D) problems were analyzed to assess the impact on eigenvalue and pin power distributions with low-fidelity, coarse square cell reflector representations: (1) a Watts Bar Nuclear Plant Unit 1 (WBN1) quarter-core slice with depletion and (2) an AP1000 quarter-core slice. In this work, the analyses showed that the effect on eigenvalue is fairly small, but the effect on pin power is more pronounced, especially locally in the assemblies closest to the periphery, where the maximum pin power difference is nearly 3.5% in the AP1000 case. Two additional 2-D problems were used to assess the comparison between the low-fidelity coarse square cell treatment and a high-fidelity geometric representation that uses subpin material specification: (1) the same WBN1 quarter-core slice and (2) a representative model of the NuScale small modular reactor (SMR), which features a solid reflector design with moderator holes. These results demonstrate that even a coarse, low-fidelity representation adequately captures the necessary simulation characteristics. Last, these capabilities were applied to the 2-D WBN1 quarter-core depletion to assess the impact on vessel fluence using VeraShift. From adjoint calculations, pins along the periphery were observed to be of highest importance for fluence calculation, so the impact of the reflector representation in MPACT could theoretically substantially affect the predicted result. However, it was observed that the change in pin powers along the periphery minimally impacts the maximum vessel fluence with a difference within the statistical uncertainty but provides terrific insight on the sensitivity of the peripheral pins.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors

Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.

TRISO↗
DOE OSTI2012

Calibration Monitoring for Sensor Calibration Interval Extension: Gaps in the Current Science Base

Currently in the United States, periodic sensor recalibration is required for all safety-related sensors, typically occurring at every refueling outage, and it has emerged as a critical path item for shortening outage duration in some plants. International application of calibration monitoring has shown that sensors may operate for longer periods within calibration tolerances. This issue is expected to also be important as the United States looks to the next generation of reactor designs (such as small modular reactors and advanced concepts), given the anticipated longer refueling cycles, proposed advanced sensors, and digital instrumentation and control systems. Online monitoring (OLM) can be employed to identify those sensors that require calibration, allowing for calibration of only those sensors that need it. The U.S. Nuclear Regulatory Commission (NRC) accepted the general concept of OLM for sensor calibration monitoring in 2000, but no U.S. plants have been granted the necessary license amendment to apply it. This paper summarizes a recent state-of-the-art assessment of online calibration monitoring in the nuclear power industry, including sensors, calibration practice, and OLM algorithms. This assessment identifies key research needs and gaps that prohibit integration of the NRC-approved online calibration monitoring system in the U.S. nuclear industry. Several technical needs were identified, including an understanding of the impacts of sensor degradation on measurements for both conventional and emerging sensors; the quantification of uncertainty in online calibration assessment; determination of calibration acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and assessment of the feasibility of using virtual sensor estimates to replace identified faulty sensors in order to extend operation to the next convenient maintenance opportunity.

Calibration Interval Extension↗
DOE OSTI2013

Advanced Instrumentation for Extreme Environments

The Department of Energy (DOE) Office of Nuclear Energy (NE) is pursuing embedded instrumentation and controls (I&C) technology for next generation nuclear power generation applications. Embedded systems encompass a wide range of configurations and technologies; we define embedding in this instance as the integration of the sensors and the control system design into the component design using a systems engineering process. Embedded I&C systems are often an essential part of developing new capabilities, improving reliability, enhancing performance, and reducing operational costs. The new intrinsically safe, more efficient, and cost effective reactor technologies (Next Generation Nuclear Plant and Small Modular Reactors) require the development and application of new I&C technologies. These new designs raise extreme environmental challenges such as high temperatures (over 700 C) and material compatibility (e.g., molten salts). The desired reliability and functionality requires measurements in these extreme conditions including high radiation environments which were not previously monitored in real time. The DOE/NE Nuclear Energy Enabling Technologies (NEET) program currently has several projects investigating I&C technologies necessary to make these reactor designs realizable. The project described in this paper has the specific goal of investigating embedded I&C with the following objectives: 1.Explore and quantify the potential gains from embedded I&C improved reliability, increased performance, and reduced cost 2.Identify practical control, sensing, and measurement techniques for the extreme environments found in high-temperature reactors 3.Design and fabricate a functional prototype high-temperature cooling pump for molten salts represents target demonstration of improved performance, reliability, and widespread usage There are many engineering challenges in the design of a high-temperature liquid salt cooling pump. The pump and motor are in direct contact with molten fluoride salt at 700 C (1,292 F) as part of a reactor cooling loop. The motor-pump combination during normal operation would be red-hot (Figure 1). This environment challenges every facet of the design including seals, wiring, magnetic materials, and sensors. In this paper, we discuss the challenges of sensor design in extreme environments and specifically the sensor design for a high-temperature fluoride salt coolant pump. This pump will be used as a test-bed for embedded I&C development and validation in extreme environments.

Extreme Environments↗
DOE OSTI2024

Data-driven Quasi-static Surrogate Model for Nuclear-powered Integrated Energy Systems

The integration of nuclear power into energy systems presents a promising avenue to address the growing global energy demands while minimizing greenhouse gas emissions. In this paper, we introduce a data-driven quasi-static surrogate model for nuclear-powered Integrated Energy Systems (IES) that comprises various components, including a small modular reactor (SMR), energy manifold (EM), balance of plant (BOP), high-temperature steam electrolysis (HTSE), and district heating (DH) system. Traditional physics-based models for these components often entail significant computational overhead and time consumption, necessitating the development of efficient surrogate models. The development of a complete surrogate model for the IES involves the creation of individual surrogate models for each component, leveraging machine learning techniques and simulated data. These isolated surrogate models are subsequently integrated, enabling a holistic view of the IES and reducing the computational burden associated with detailed physics-based simulations. This paper outlines the development process, validation, and the performance evaluation of the surrogate models. The findings shed light on the accuracy and applicability of the surrogate models in practical scenarios, demonstrating their potential to expedite the analysis of nuclear-powered IES and inform future research and development efforts.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2016

Methodology for Risk-informed O&M decision making in power plants

Incorporation of real-time component information using equipment condition assessment (ECA) through the development of enhanced risk monitors (ERM) for active components in advanced reactor (AR) and advanced small modular reactor (SMR) designs. We incorporate time-dependent failure probabilities from prognostic health management (PHM) systems to dynamically update the risk metric of interest. This information is used to augment data used for supervisory control and plant-wide coordination of multiple modules by providing the incremental risk incurred due to aging and demands placed on components that support mission requirements.

Source record↗
DOE OSTI2018

Rancor Microworld Simulation Environment for Nuclear Process Control

The Rancor Microworld is a simplified nuclear power plant simulator. It represents the process of an integrated pressurized light water reactor core with a secondary Rankine cycle turbine/generator. The reactor core is governed by simplified core physics. Operators have the ability to raise and lower control rods to change the reactivity and consequently generate more or less electricity. The microworld provides a digital human machine interface for operating and monitoring the plant efficiently to maximize profits within the simulation. The simulation supports multi-unit operations analogous to small modular reactor plants and supports faults on various components within each unit to simulate some common training scenarios such as reactor and turbine trip events. The microworld includes experimental modules to log process parameters and both automated and user control inputs and events, administer situation awareness probes, and interface with eye tracking software.

Ulrich, ThomasA.↗
DOE OSTI2022

Initial study on cross-section generation requirements for a PBR equilibrium core

A Serpent model of the equilibrium core HTR-PM small modular nuclear reactor in China, was developed for use in cross-section preparation studies in order to guide methods development for the Griffin reactor multiphysics application. The model includes detailed isotopics for 10 distinct pebble burnup groups in 126 core zones with unique fuel and moderator temperatures obtained from a coupled neutronics-thermal-fluids equilibrium core calculation using Griffin-Pronghorn. A sensitivity study of the fuel and moderator temperatures for various core regions was performed with the MOOSE stochastic tools. The results show that the uncertainties are, not unexpectedly, dominated by the value of the fluid temperature and that the power level, heat transfer coefficient and effective conduction to neighboring pebbles and fluid constitute, at best, second order effects. The temperature uncertainty range varies from 28 K to 57 K between the core entry and exit planes, respectively, but these values are probably higher. We still have to quantify the significance of these uncertainties in the preparation of cross-sections in future work. In addition, we verified that the effective pebble approximation used in the PEBBED and V.S.O.P. computer codes works well for the preparation of region averaged cross-sections. Nevertheless, there are some discrepancies in the cross-sections when compared to the multi-pebble model, which could affect the prediction of peak values and the depletion calculation. We conclude that is highly desirable for future studies with Griffin to be able to handle both the 'effective' pebble approximation and the multi-pebble approach for various pebble burnup groups. This enables Griffin users with the flexibility to perform higher-fidelity studies. Finally, we initiated the preparation of cross-sections for various core regions from the full core Serpent reference model. We quantified the differences in 26 group cross-sections from infinite domain models versus the full core approach. These reference cross-sections will serve to verify the double heterogeneity, self-shielding, and spectrum-correction methods in Griffin. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Simulation of CEFR neutronic start-up tests with FENNECS

This paper presents simulation results of selected Neutronic Start-up Tests of the China Experimental Fast Reactor (CEFR) obtained by the neutronics code FENNECS that have been performed within the frame of the IAEA Coordinated Research Program I31032. The Finite Element Neutronics code FENNECS is developed at GRS and solves the few-group steady-state and transient diffusion equation using a Galerkin-based finite element approach. Its main purpose is the safety assessment of Small Modular Reactors and Micro Reactors with complex geometry (e.g., rotating control drums) which gain increased interest internationally. Serpent has been applied to create reference models of the CEFR and cross-section libraries in 10 energy groups for FENNECS. Using these libraries, the following experiments have been simulated with FENNECS: net criticality, control rod integral and differential worth, void reactivity effect, subassembly exchange reactivity effects and reaction rate distribution. The obtained satisfactory agreements with the measurements represent a valuable contribution to the validation of FENNECS. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗