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Small and micro-reactor advanced reactor development in the USA. Advanced Reactor programs, nuclear renaissance, and INL as a test bed facility. These slides are for presentations in general on the past, INL and nuclear engineering worldwide, GEN IV reactor designs, SMRs and micro Rx designs, multiple commercial designs, and what INL as a test bed means and who uses this to their advantage.
To license new and advanced reactor designs, regulators must be convinced that their unique safety cases—relative to existing large scale reactors—have been adequately addressed by the designed reactor protection systems. In water cooled small modular reactors (SMRs), droplet entrainment in steam flow has significant implications on the progression of accident scenarios due to its compact design features, which requires representative test data applicable to SMR designs. Computer code, modeling and simulation (M&S) tools and models require adequate verification, assessment, and qualification. This includes M&S results validation against scaled empirical data within allowable uncertainty bands to gain regulatory approvals during the various stages of reactor system design, demonstration, and commercialization. However, measurement uncertainty within the empirical datasets and test data applicability ranges requires careful consideration of M&S inputs (i.e., boundary conditions, and initial conditions), and verification and validation efforts. This study focuses on uncertainty quantification in designing scaled test facilities for SMR applications with appropriate measurements and a standard data-reduction method to estimate thermal hydraulics characteristics parameters that incorporate physics phenomena of interest. In addition, this study supports the evaluation model development and assessment process using M&S that interfaces with advanced computing tools and digital twin capabilities. This will allow synchronization between experiment and modeling approaches for droplet entrainment testing and analysis, improving diagnostics, prognostics, and decision-making to accelerate regulatory approval.
The AGN-201M reactor is a small, thermal spectrum reactor located at the University of New Mexico (UNM). It is moderated by polyethylene, reflected by graphite, with fuel comprised of uranium microspheres embedded in polyethylene plates that are separated by an aluminum baffle. The dominance ratio of a multiplying system is the ratio of the first higher-mode eigenvalue to the fundamental, k{sub 1}/k{sub 0}. It is a key parameter in characterizing the rate of convergence, analyzing system stability, and determining how tightly coupled the system's neutron fission is. The purpose of this paper is to investigate the system's unusually low dominance ratio of 0.632. This investigation varied the reactor's physical parameters, such as the fuel geometry, fuel density, and reflector thickness, and examined their impact on the reactor's dominance ratio. Additionally, this study examined neutronically similar systems in hopes of identifying common causes for systems with low dominance ratios. The reason why the dominance ratio of the reactor is small compared to larger thermal reactors was determined to be because of its significantly smaller geometry and the composition of the fuel plates. The reflector's effect on the dominance ratio is small in comparison to the other factors, but was found to have a non-zero effect. Furthermore, the AGN-201M was found to have a significantly lower dominance ratio than systems in which it shares a very high (c{sub k} >95%) degree of neutronic similarity.
Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.
Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.
The United States Department of Energy (US DOE) is currently supporting the development of various advanced and small modular reactor (SMR) designs. Several of these designs have commenced license application with the US Nuclear Regulatory Commission (US NRC). These reactors have improved safety features that may significantly reduce radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a smaller containment volume relative to the available fission product depositional surface area, which supports increased fission product retention in the containment vessel. Pressurized water reactors (PWR) and boiling water reactors (BWR) with this feature include the integral pressurized water reactor (iPWR) and the BWRX-300 design by General Electric. A prior research program supported by the US DOE quantified the source term reduction associated with light water iPWRs and developed iPWR-specific theoretical models for fission product deposition rates. This program included a sequence of research projects that started with a feasibility study, development of theoretical models that predict higher deposition rates, and finally, development of empirical data for verification and validation of the theoretical models. The current project, which is a feasibility study, is the first step in a similar program to quantify the source term reduction associated with small and advanced light water BWRs.
The Advanced Gas Reactor Fuel Development and Qualification (AGR) program has established a substantial technical foundation to support private entry into the U.S. high-temperature gas-cooled reactor market. However, emerging tristructural isotropic (TRISO)-fueled reactor applications include small modular reactors and microreactors with longer fuel residence times, which may expose fuels to higher time-at-temperature (TAT) values than were explored by the AGR program. Increased TAT could affect diffusive and thermomechanical behaviors such as Pd penetration, fission gas release, creep, and fission product transport. In this work, we applied multiscale best-estimate BISON fuel performance modeling to assess these effects within a representative design space based on the AGR-5/6/7 experiment and analyzed trends in predicted particle and compact fuel performance metrics with possible implications for near-term fuel design and qualification. BISON unambiguously predicted that TRISO fuel performance is sensitive to TAT. Increasing TAT was not predicted to increase the magnitude of failure-inducing tangential stresses in particle coating layers. Predictions obtained using a mechanistic model for Pd penetration indicated that penetration depth does not depend strongly on TAT. While these observations suggest that AGR testing provides a conservative upper bound for the steady-state operation of TRISO particles at lower powers and higher residence times, BISON also predicted that the release of poorly retained Ag would increase with TAT. Because these analyses applied models to extrapolate beyond the available experimental data, the authors recommend performing targeted experiments to confirm these predictions. Nevertheless, these predictions may provide reactor developers with enough confidence to make near-term design decisions associated with the potential fuel performance trade-offs of increasing TAT.
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To enable the broad deployment of microreactors in fundamentally new application regimes (i.e., mobile and autonomous operations), their safety must be indisputable in terms of possessing inherent resistance to severe offsite dose consequences. Therefore, mechanistic beyond-design-basis event source term calculations that demonstrate a sufficiently large margin of safety will be required to accommodate these new application regimes, which have no history of commercial regulation. Even for traditional reactor operation configurations, safety analysis expertise and familiarity with accident phenomena and conditions in microreactors—specifically those with heat pipe primary cooling arrangements—are lacking compared with other advanced reactor concepts and small modular reactors. Recently, modeling and simulation tools to account for unique heat pipe design aspects have been developed by Sandia National Laboratories with MELCOR and by the US Department of Energy’s (DOE’s) Office of Nuclear Energy Advanced Modeling and Simulation Program with BlueCRAB. However, further demonstration and assessment of potential knowledge gaps are needed to support these codes’ broad usage by the microreactor community. Through the DOE Microreactor Program, an initial assessment of these two tools and guidance on how an evaluation model could be constructed was performed and is reported herein. Moreover, a proposed approach for demonstrating an evaluation model using these two tools is outlined.
The purpose of this initial siting evaluation is to consider the suitability of the Salt River Project (SRP) owned land at/near Coronado Generating Station (CGS) for a nuclear generating station (including advanced light water reactors, light water small modular reactors, and non-light-water designs1) and to assess if there are any exclusionary or avoidance factors associated with coal to nuclear repowering at CGS. This initial siting evaluation considers several characteristics (e.g., environmental conditions, seismic concerns, site footprint, water use, etc.) and highlights favorable/preferred characteristics as well as potential risks to feasibility. The results are intended to inform SRP on the strengths and weaknesses associated with the CGS site and inform the selection of candidate nuclear technologies for the CGS site that align with SRP’s mission and business objectives. This report relies on industry-recognized siting guidance, including Electric Power Research Institute’s (EPRI’s) Siting Guide (Reference 1) 2 and Nuclear Regulatory Commission’s (NRC’s) Regulatory Guide (RG) 4.7 (Reference 3), as well as nuclear domain expertise within the Gateway for Accelerated Innovation in Nuclear (GAIN), MPR Associates, Inc. (MPR), and the Idaho National Laboratory (INL).
An investigation was conducted to determine minimum dimensions and minimum weight obtainable in a design for a reactor using uranium-233 nitride or plutonium-239 nitride as fuel. Such a reactor had been considered by Krasner et al. (1971). Present space power status is discussed, together with questions of reactor design and power distribution in the reactor. The characteristics of various reactor types are compared, giving attention also to a zirconium hydride reactor.
An inadvertent actuation block valve includes inlet and outlet orifices being in selective fluid communication via a chamber. A disc is disposed within the chamber and a bellows is configured to contract at a predetermined pressure differential between reactor fluid entering a reference pressure orifice and control fluid entering the inlet orifice. When the bellows contracts, the disc engages the outlet orifice and isolates fluid communication between the inlet and outlet orifices. The inadvertent actuation block valve prevents inadvertent opening of an emergency core cooling valve when a reactor is at operating pressure that is above the predetermined set pressure range. The inadvertent actuation block valve permits the emergency cooling valves to open and to remain open when reactor pressure is below the predetermined set pressure range. The inadvertent actuation block valve does not impede long term emergency cooling that occurs when the reactor is at low pressure.
This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.
This study focuses on design parameter sensitivity studies pertaining to a steam generator (SG) model, using both Python and machine-learning tools. The SG model is a mathematical representation (including fluid flow and heat transfer equations/models/correlations) of a steam-generating unit in a pressurized water reactor (PWR)-type small modular reactor (SMR) system. Design studies involve changing the model’s input design parameters (e.g., temperature, pressure, mass flow rate) to observe the resulting effects on the output of the system (e.g., heat transfer coefficient [HTC], Nusselt number, heat transfer performance). Sensitivity studies analyze the degree to which system output and/or desired parameters (e.g., HTC or heat transfer performance) are sensitive to changes in input parameters. By using machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN) developed at Idaho National Laboratory (INL), detailed design parametric sensitivity studies and model optimization were performed. Six input parameters—namely, the pressure, temperature, and mass flow rate for the inlet of the primary-side (hot fluid) and secondary-side (cold fluid) of the SG—were randomly perturbed via RAVEN’s Monte Carlo Sampler module, using uniform distributions (±1% relative changes). The analysis results give valuable insights into SG system performance and optimization, and provide justification for researching optimized sensor placement to effectively monitor and obtain experimental data.
Small modular reactors (SMRs) are gaining significant attention as a promising solution for clean and sustainable nuclear-power generation. However, the operation of SMRs is subject to various challenges, including two-phase flow instabilities. Flow instability has the potential to trigger flow-induced vibration and cyclic fluctuations in local thermal stress. These instabilities frequently manifest because of the complex interplay among a multitude of factors, encompassing thermal-hydraulic conditions, the geometric configuration of the steam generator, and operational parameters. These conditions could subsequently lead to premature critical heat flux, equipment malfunctions, and other safety concerns. Here, the endeavor to address steam-generator flow instabilities is of utmost importance in augmenting the sustainability and efficiency of contemporary energy production. This study offers a comprehensive review of instabilities in two-phase flow, with a particular focus on the influential factors impacting the stability of flow boiling. Furthermore, it delves into the processes of identifying, characterizing, and ameliorating these instabilities, emphasizing pivotal findings, methodologies employed, and avenues for prospective research. The primary parameters of concern encompass the efficient transfer of thermal energy, the optimization of mass-flow rates, and the establishment of favorable boundary conditions, all in the context of steam generator design to alleviate instability for water-cooled SMRs. These discernments bear substantial ramifications for enhancing reactor performance and ensuring operational safety.
The power distribution estimation method based on the ex-core detectors, or PHOEBE, was demonstrated at Kyoto University Critical Assembly, KUCA. Generally, core monitoring systems use in-core neutron detectors. Since inside the core is a harsh environment, the maintainability and reliability of the detectors are deteriorated. On the other hand, the environment outside the core is milder: core monitoring by the ex-core detectors improves the maintainability and reliability especially for small modular reactors and micro-reactors. However, neutron information from the inner region of the core is lost at the ex-core detectors. To recover the information, the authors proposed to utilize the power correlation between the fuel regions. PHOEBE concept was demonstrated at KUCA with distorted power distributions simulated by control rod patterns. The relative power distribution estimated by PHOEBE agreed with that calculated by Monte Carlo simulation code MVP. PHOEBE approximately reproduced the trend of the distorted power distributions calculated by MVP code. By contrast, the case of without power correlation between fuel regions produced significant different power distributions. Therefore, the advantage of considering the power correlation between fuel regions was also demonstrated. (authors)
As part of a cooperative research and development agreement between the US department of Energy's Oak Ridge National Laboratory and Holtec International subsidiary SMR LLC, core simulator models of the SMR-160, a light water-cooled and -moderated, natural circulation, small modular reactor, were developed and compared. The model results indicate pin-by-pin power histories, critical boron concentration, and other key performance indicators useful for developing multi-cycle core loading patterns. One model uses industry standard nodal methods with a simple thermal-hydraulics solution, and the other uses an advanced deterministic neutronics solver with a coupled subchannel calculation. The goals of the comparison are to provide the reactor designer with a detailed and coupled core physics reference solution, inform the reactor designer of nodal model best practices based on the code-to-code comparison, and set up a workflow for performing subsequent detailed core simulator calculations. The results of this work will be used in the development of coupled neutronic, thermal hydraulic, and fuel performance models for advanced analysis of the SMR-160. (authors)
On 25–26 March 2023, the U.S. National Council on Radiation Protection and Measurements (NCRP) held its 2024 annual meeting in Bethesda, Maryland, USA. The NCRP dates from 1929, and this meeting celebrated the 60th anniversary of receiving a U.S. Congressional Charter. For this annual meeting the NCRP felt it was essential to provide a briefing about advanced and small modular nuclear reactors (SMRs). The Journal of Radiological Protection is delighted to publish the following synopsis of material presented at the U.S. NCRP meeting. This synopsis is divided into five sections. The first section provides an overview of the whole meeting together with summaries of two context setting overview papers. The following four sessions of this synopsis are specific to advanced and small modular nuclear power reactors. The meeting also included keynote presentations by three of NCRP annual award recipients. The meeting topical areas were Technology Overview and Critical Issues. The individual papers laid the groundwork to understanding reactor technologies, terminology, and the fundamental concepts and processes for electrical generation. The perspectives of the U.S. Environmental Protection Agency and states, through the Conference of Radiation Control Program Directors were provided. The papers included a discussion of diverse topics including potential emergency preparedness considerations, radiological survey requirements, an evaluation of the future of nuclear power, the economics of reactors (both large and small), and the critical issues identified by the recent National Academies of Sciences’ study on advanced reactors. Here, the summary papers were developed to briefly document the major points and concepts presented during the oral papers presented at the 2024 NCRP Annual Meeting. The meeting heralded the dawn of a new era for commercial nuclear power.