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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.

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At least 865 records · Page 48
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Condensed Phase Deactivation of Solid Brønsted Acids in the Dehydration of Fructose to Hydroxymethylfurfural

Catalyst deactivation resulting from the hydrothermal leaching of sulfonic acid residues and the deposition of carbonaceous residues was studied using condensed phase flow reactor experiments along with state-of-the-art solid-state NMR. Several commercially available sulfonic acid-containing heterogeneous Brønsted acids were compared by measuring the rates of sulfonic acid breakdown at hydrothermal flow conditions of 160 °C. Amberlyst 45 was found to show higher hydrothermal stability when compared to both Nafion and Amberlyst 15, with <10% loss in acidity after 48 h. The dehydration reaction of fructose to hydroxymethylfurfural (HMF) was used as a model system to compare deactivation rates from carbon deposition (fouling) to those from sulfur leaching, and deactivation from fouling was shown to be dramatically faster than that from sulfonic acid leaching alone. Fouling rates were then investigated in greater detail by comparing the influence of several factors including reactant, solvent, residence time, and feed concentration. The only successful approach to minimize fouling was the use of a polar aprotic solvent [dimethyl sulfoxide (DMSO)] with dilute (50 mM) reactant streams. In aqueous systems, operating the reactor in a regime with low conversion conditions (short residence times) does not significantly improve the longevity of the catalyst. Spent catalysts were characterized using 13C solid-state NMR spectroscopy enhanced by dynamic nuclear polarization. Additionally, in situ 1H and 13C high-resolution magic angle spinning (HR-MAS) solid-state NMR spectroscopies were used to investigate the solvent influence at the catalyst interface. The HR-MAS NMR studies showed that in polar aprotic solvents, the increased acidity leads to greater selectivity toward HMF; more importantly, that the dehydration products do not readily adhere to the surface in DMSO, in contrast to their behavior in water. The results demonstrate that more active and longer-lived acid catalysts could be obtained by tuning the solvent and surface polarity to allow for efficient desorption of products, thereby reducing the catalyst deactivation that occurs due to fouling.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Handbook of Advanced Manufactured Material Properties From TCR Structure Builds at ORNL (FY2019)

There is growing interest in the use of additive manufacturing techniques for the production and deployment of components for nuclear power applications. Specifically, the Transformational Challenge Reactor (TRC) program seeks to demonstrate the application of advanced manufacturing to rapidly develop and deploy a microreactor. As part of the TCR objective, a database on materials performance, such as thermophysical properties and mechanical performance, is required to help inform nuclear reactor designers, operators, and regulators regarding possible material limitations using emerging manufacturing techniques. The following report seeks to collect all available data regarding the material performance of candidate alloys and manufacturing processes pursued within the TCR program at Oak Ridge National Laboratory. At this time, this database focuses only on materials derived within the program and does not extend to additional manufacturing processes, manufacturers, or other advanced materials. Where appropriate, recommendations for applications of the data are provided and current knowledge gaps are identified.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Upgrading Biogas through in situ Conversion of Carbon Dioxide to Biomethane in Anaerobic Digesters

Organic waste streams generated by wastewater treatment plants, agricultural operations, and food processing industries represent an important yet underutilized opportunity for renewable energy production in the United States. Through anaerobic digestion, these waste streams can produce biogas, a mixture primarily composed of methane (CH4) and carbon dioxide (CO2), that can be upgraded to pipeline-quality natural gas. However, most existing upgrading technologies remove CO2 from biogas rather than utilizing it, leaving a significant portion of the potential energy unused. This project investigates a novel biological upgrading approach that converts CO2 into additional CH4 by supplying hydrogen (H2) to specialized microorganisms capable of performing hydrogenotrophic methanation. The main challenges associated with biological biogas upgrading are related to hydrogen supply, gas-liquid mass transfer, and process stability. First, due to the high cost of hydrogen gas, it is preferable that H2 be produced on-site using renewable energy sources such as wind or solar power. Second, hydrogen has low solubility in liquids, which limits its availability to microorganisms and requires strategies to improve gas dissolution and transfer within the reactor. Third, process inhibition may occur as a result of increased pH caused by CO2 consumption or elevated H2 partial pressure, both of which can negatively affect methanogenic activity. Although research in these areas has advanced during the course of this project, these challenges have not yet been fully resolved. To date, the biological systems that have achieved the highest methane concentrations are typically ex-situ reactors, where operational conditions can be more easily controlled. For this reason, the findings of the present project remain highly relevant. The project goal was to develop an innovative system that can accomplish biogas upgrading via biological conversion of CO2 to CH4, in a novel hybrid approach that combines the advantages of both in-situ and ex-situ systems. The proposed system employs a three-phase upflow anaerobic bioreactor with H2 delivery through a gas-permeable membrane, enabling efficient hydrogen transfer and microbial conversion. Under optimized operating conditions, the system achieved 99% H2 consumption and 90% CO2 conversion. A subsequent gas cleaning stage was implemented to further improve gas quality and meet target purity standards. The upgraded gas composition reached 97.7% CH4, 2.2% CO2, and 0.97% O2, while H2S concentrations remained below detection limits. In addition, a flue gas-driven inorganic thermoelectric generator (TEG) system was designed and experimentally validated as a potential source of electricity for H2 production. The system consisted of six TEG modules connected in series and achieved an open-circuit voltage of 4.5 V and a maximum power output of 224 mW at a temperature difference of approximately 53.5 °C, demonstrating effective conversion of waste heat into electrical power under simulated flue gas conditions. Finally, a comprehensive techno-economic analysis was completed to evaluate the capital and operating costs associated with the proposed system. The results provide important insights to guide future scale-up, optimization, and potential deployment of integrated biological biogas upgrading technologies.

09 BIOMASS FUELS↗
DOE OSTI2023

Integral and Separate Effects Test Facilities To Support Water Cooled Small Modular Reactors: A Review

This study reviews previous experimental facilities and test programs relevant to water-cooled reactor system design and analysis to meet regulatory compliances. This study aims to find the best solution for designing the required experiments, obtaining necessary test data, and verifying the developed computer code/models to support the new reactor design and development while minimizing cost and time while leveraging experiences from previous facilities to minimize. Nuclear reactor licensing requires supportive design, analysis, and experimental results to ensure the safety of the full-scale prototype reactor in regular operation, as well as during postulated accident scenarios. These reactor design analyses are generally performed using system codes and other associated simulation tools that require assessment, verification and validation using an appropriate experimental dataset. Experimental facilities used for reactor system safety analysis and system code assessments are categorized into integral effects test (IET) and separate effects test (SET) facilities. Further, these IET and SET experiments and studies use geometrically scaled systems to reproduce the prototype system behavior at a reasonable cost, albeit with some scaling-related distortions. The design challenge of these model facilities is to identify and minimize scaling distortions while reproducing the most important operational phenomena in steady-state operation and in postulated accident scenarios. Lessons learned from previous experimental facilities, models, and correlations can support the development of new multipurpose, scaled, hybrid, integrated, and modular experimental facilities for advanced light water-cooled small modular reactors (SMRs). Successful operation of these facilities can significantly reduce upfront reactor development and demonstration costs and time to deployment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

AGN-201 Digital Twin Concept of Operations

The AGN-201 is a nuclear reactor at Idaho State University (ISU) and is currently being used in the development of a digital twin (DT) with Idaho National Laboratory (INL). The goal is to create a DT (called the AGN-201 DT) which can monitor an operating nuclear reactor to determine when the reactor is being operated normally; normal operations are any operations that is declared by the ISU staff. To accomplish this, researchers at INL developed a DT ecosystem which can ingest data from the AGN-201 reactor. This data is fed into a series of machine learning (ML) and reactor physics models. The ML and reactor physics models assess the data and determine if the reactor is operating normally. Event(s) that are flagged as anomalies are investigated by the INL staff to determine if any undeclared experiments were conducted. The AGN-201 DT was verified in July/August 2023, when the ISU staff performed undeclared experiments and the INL staff were able to assess the anomalous events and determine what likely caused each event. This project was a steppingstone to promote the use of DTs for international safeguards and marks the first time a DT was able to monitor a nuclear reactor for this purpose. This document provides the concept of operations (CONOPS) for the AGN-201 digital twin (DT). The CONOPS describes how the AGN-201 and AGN-201 DT are expected to operate and details how each are established.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Time Scale Effects on the Temperature Coefficient of Reactivity

For a nuclear reactor, the temperature coefficient of reactivity (TCR) is of significance for both power control and safety analysis, but it is not always a constant. For example, the TCR may be dependent upon temperature due to Doppler broadening, and may be also dependent upon time scale due to heat transfer and thermal expansion. A fast burst reactor is a favorable nuclear facility to simulate the time scale effects on the TCR. Fast burst reactors are generally unmoderated research reactors of enriched uranium metal or uranium alloys. The neutrons are fast and the Doppler broadening effects can be neglected, and the TCR is mainly contributed by fuel thermal expansion. Fast burst reactors can operate on both steady state power mode and burst power mode. In steady state power mode, the systems are assumed to be in thermal equilibrium. In burst mode, the full width at half maximum (FWHM) of the power is from tens of microseconds to a few milliseconds, and the heat transfer is generally neglected. If FWHM is short enough, thermal expansion will have a time delay due to inertial effect. This paper is aimed to study the exact time scale effects on the TCR due to thermal expansion. The TCR contributed by the Doppler broadening effects was neglected, because it is relatively small (The k{sub eff} as a function of temperature was also calculated using the Monte Carlo particle transport code JMCT by our group, and no definite difference due to the temperature change was obtained near room temperature), and there is no method to evaluate the time scale dependence of the Doppler broadening effects yet. In the time scale from 1 x 10{sup -6} s to 1 x 10{sup 5} s, the TCR of a Godiva-like spherical fast burst reactor was calculated with perturbation theory. A Godiva-like spherical fast burst reactor was studied, the calculation of the time scale effects on TCR involved thermal conduction, thermoelasticity and reactivity variation analysis. Both the thermal and the thermoelastic problems were solved using ANSYS software. ANSYS is software implementation of finite element simulations, which is widely used in engineering field. The reactivity variation due to thermal expansion was calculated with perturbation theory.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Nuclear power plant instrumentation and control cyber security common vector access leading to relational common cause failures - 340

Nuclear power plants and facilities have been implementing digital system upgrades into their previously analog systems for well over twenty years. New nuclear facilities' control, security, and emergency preparedness systems are almost exclusively built on digital architectures with a high degree of communication between the various systems that are often integrated together into a central control station to aid in operation or security of the facility. As digital systems become more widespread in nuclear facility control system architectures, cyber security related issues have become a significant concern to operators, regulators, governments, and other groups. Among the many concerns related to digital systems and cyber security is the area of common cause and common mode failures. This paper introduces, defines, and discusses some sources of common cause failure from a cyber security perspective: common vector access. This refers to specific access points that an adversary can exploit through a single attack sequence that have the potential to provide relational failures through common cause on multiple components, subsystems, systems, or plants. This paper will further discuss interconnected processes where these access points may exist, the importance of limiting or controlling these pinch points, and some methods of protecting common vector access points. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Equibiaxial Flexure Strength of a Superfine-Grained Nuclear Graphite

The strength of advanced graphite is reported in accordance with ASTM D7846-16 using a two-parameter Weibull distribution for uniaxial strength testing. The rule of thumb for Weibull distributions is to use a minimum of 30 strength measurements to have a high level of confidence in the Weibull characteristic strength and the Weibull modulus. These large sample sets for statistical confidence are easily obtained for as-manufactured graphite, but determination of the Weibull two-factor parameters on graphite that has been exposed to neutron irradiation is nearly impossible due to irradiation testing restraints on specimen size and space limitations in test reactors for the accommodation of specimens. These restrictions have resulted in irradiation programs that use sub-sized specimens to measure uniaxial strength change, but the specimen geometries still hinder the number of replicate specimens for each irradiation condition. In contrast, this work used an advanced ceramic standard, ASTM C1499-05, as the foundation for equibiaxial strength testing of a superfine-grained nuclear graphite to investigate the irradiation-induced strength change and any changes to the two-parameter Weibull distribution. The goal was to investigate the effect of specimen size on the two-parameter Weibull distribution parameters in support of the small specimens and geometry desired for irradiation. A study of the effect of a reduced sample population was also undertaken to investigate the reliability of using the two-parameter Weibull distribution to study changes to the materials properties caused by neutron irradiation damage. The results of this work suggest that these equibiaxial specimens would be ideal for statistically significant studies of the effects of irradiation on the Weibull characteristic strength (but not the modulus) and for providing a method for surveillance specimen campaigns for future operating commercial reactors.

Campbell, Anne↗
DOE OSTI2017

BWR modernization project: installation of a new digital feedwater control system - 295

A European BWR plant carried out a relevant and pioneering modernization project regarding the Feedwater and Feed Pump Turbine Control System, replacing the legacy analogic control with the new General Electric MarkVIe digital system. This digital instrumentation and control modernization, motivated by obsolescence, had a great impact on the main control room and on the plant operation. Tecnatom collaborated with the nuclear power plant in the project performing the following activities: early installation of the system in the full scope and classroom simulators, Human Factors Engineering (HFE) Verification and Validation (V and V), Simulation Assisted Engineering (SAE), design and implementation of cyber secure communication with the Plant Process Computer (PPC), operating procedures modification and plant operators training. Those activities contributed to the success of the project that was completed during the refueling outage as planned. Scheduling HFE verification before the design freeze allowed the plant to modify the Human Machine Interface (HMI) to comply with the standards and increase safety without overcharging the project. SAE resulted in the reduction of costs and time during the system commissioning through the detection of discrepancies and improvements in the control, covering different plant scenarios. All the activities carried out in the simulators before the implementation of the system in the plant contributed to validate it and to build confidence in the new control among the plant personnel. New digital system communication interfaces with the PPC saved cable laying works and allowed future expansions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Health Physics Research Reactor Criticality Accident Alarm System Benchmark Overview [Slides]

This presentation discusses the Health Physics Research Reactor (HPRR) or Fast Burst Reactor (FBR) and how available data from Health Physics Research Reactor (HPRR) operation to can be used to create a benchmark report for inclusion in the ICSBEP, as a Criticality Accident Alarm System (CAAS) shielding benchmark. The evaluation of experimental data is also presented, specifically four experimental candidates are considered of potential value for the benchmark: 1. Neutron source estimation from a HPRR pulse (energy spectrum, fission yield), 2. Threshold Detector Unit (TDU) measurements at different distances from a HPRR pulse, shielded and unshielded, 3. Sulfur pellet activation at different distances from a HPRR pulse, shielded and unshielded, and 4. Total neutron fluence from a HPRR pulse measured by Bonner Sphere Spectrometry, shielded and unshielded. Additionally, a benchmark model overview is discussed as are sample calculation results.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2017

Towards a Remote Handling Toolkit for Fusion: Lessons Learnt and Future Challenges - 17360

It is increasingly recognised within the fusion research community that the next generation of remote maintenance solutions will be a 'device defining driver' for a future fusion power plant. The unique Remote Handling Systems that operates at JET, EUROfusion's research reactor that is operated by the UKAEA at Culham, Oxfordshire, remains the archetype for future designs. At ITER, the international fusion research experiment being constructed in the South of France, many $100's millions will be invested in supplying and then operating remote handling solutions. This paper will provide a brief history of the JET Remote Handling System in recent campaigns and will present initial results of the ITER remote operated cutting and welding trials. Finally, the paper will comment on reasons for establishing RACE, the UKAEA's centre for Remote Applications in Challenging Environments and the task of integrating new robotics including autonomous vehicles, snake-like manipulators and autonomous remote maintenance systems into the design of a future fusion reactor. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Fault propagation and effects analysis for designing an online monitoring system for the secondary loop of a nuclear power plant part of a hybrid energy system - 393

This paper studies the propagation and effects of faults of critical components that pertain to the secondary loop of a nuclear power plant found in Nuclear Hybrid Energy Systems (NHES). This information is used to design an on-line monitoring (OLM) system which is capable of detecting and forecasting faults that are likely to occur during NHES operation. In this research, the causes, features, and effects of possible faults are investigated by simulating the propagation of faults in the secondary loop. The simulation is accomplished by using the Integrated System Failure Analysis (ISFA). ISFA is used for analyzing hardware and software faults during the conceptual design phase. In this paper, the models of system components required by ISFA are initially constructed. Then, the fault propagation analysis is implemented, which is conducted under the bounds set by acceptance criteria derived from the design of an OLM system. The result of the fault simulation is utilized to build a database for fault detection and diagnosis, provide preventive measures, and propose an optimization plan for the OLM system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Light Water Reactor LEU+ Lattice Optimization

Commercial light water reactor (LWR) operators and fuel vendors in the United States are exploring potential changes to nuclear fuel that include low-enriched uranium plus (LEU+) designs to further improve operational economics (e.g., extend cycle length). LEU+ fuel is fuel with a maximum enrichment between 5 wt% and 10 wt% 235 U; it allows for higher assembly burnup but likely requires additional reactivity control, e.g., increased burnable absorbers. This report examines possible LEU+ fuel lattice design changes using the lattice physics code, SCALE/Polaris. An optimization driver called the Metaheuristic Optimization Tool (MOT) is used to automate domain space exploration and optimization of LEU+ lattice designs. Heuristics from previous LWR lattice optimization studies were used to construct the objective function and define the domain space for optimization. This work successfully demonstrated that the optimization algorithms of MOT can generate feasible, nonproprietary LEU+ lattice designs (GE14 10 × 10 and Westinghouse 17 × 17) that meet the constraints of traditional LWR lattices while extending cycle length.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Isotopic and Fuel Lattice Parameter Trends in Extended Enrichment and Higher Burnup LWR Fuel Vol I: PWR fuel

Commercial light water reactor (LWR) operators and fuel vendors in the United States are pursuing changes to nuclear fuel that include extended enrichment (EE) and accident-tolerant fuel (ATF) designs. The term EE (8% > 235 U > 5%) is used in this report to refer to a subset of high assay low-enriched uranium (HALEU) that is considered usable in commercial US LWRs in the near term. ATF features are designed to improve fuel system performance under accident conditions. One goal of EE is to improve fuel cycle economy by enabling fuel to be depleted to higher burnup than the typical current maximum pin burnup limits (62 gigawatt-days per metric ton of uranium [GWd/MTU]). Adoption of EE, ATF, and high burnup (HBU) fuels in the US commercial fleet requires a clear understanding of the effects on core physics parameters and used fuel isotopic content, as well as confidence in the accuracy of computer code predictions over an expanded range of materials, enrichment, and burnup. A thorough understanding of the applicability and adequacy of benchmark data (e.g., criticality, decay heat, isotopic content) for computer code validation is necessary to ensure that appropriate safety margins are maintained. To prepare for and support these potential changes, the effects of EE, ATF, and HBU are being assessed for selected representative LWR fuel designs. The project is divided into phases: this report summarizes the findings of Phase 1, which focuses on the lattice physics parameter and used fuel isotopic changes for a conventional Westinghouse 17×17 pressurized water reactor (PWR) design. The primary investigation tool is the SCALE Polaris code using the SCALE 56-group Evaluated Nuclear Data File (ENDF)/B-VII.1 cross sections. The goal of the current work is to (1) identify and explain important effects of EE and HBU (reactivity, lattice physics, and isotopic effects) assuming that PWR fuel design and usage remain similar to those for current enrichment fuel, (2) provide limited code-to-code comparisons with higher order cross section libraries and/or codes, and (3) identify any apparent anomalous trends in the results for further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Approach to Improve the Retention of the Fission Product Ag in TRISO Fuel Particles - 18596

The silicon carbide (SiC) coating included in TRISO particles acts as the main barrier against the radioactive fission products release into the coolant circuit. However, radioactive silver (Ag{sup 110m}) was identified outside the intact particles with a very high release fraction, which could be problematic in the licensing process for modern HTR fuel. Our previous work proposed that dissolution/recrystallization of SiC by Ag could help Ag migrate at 1800 deg. C. The aim of this work is to study whether dissolution of SiC by Ag could take place and contribute to the Ag migration at operation temperature (from 1000 to 1300 deg. C). Both, experiment and thermodynamic analysis, showed that SiC dissolution by Ag was possible above 890 deg. C. But the dissolution from 1000 to 1300 deg. C is far less effective in moving Ag due to the insufficient driving force for activating self-penetration into the grain boundaries, compared with the previous work conducted at 1800 deg. C. Compared with silver alone, the silver-palladium mixture was more permeable in SiC and showed a diffusive behavior. The diffusion front was Ag rich while most palladium remained in reaction zone rather than in the diffusion front. Furthermore, thermodynamic analysis suggested that non-stoichiometric SiC with additional free silicon could impede such a reaction. Hence it is conceivable that a sacrificial layer of SiC+Si on the inner side of the main SiC coating could improve fission product retention overall without a significant increase in cost. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Formalization of the functional analysis methodology to improve NPP I and C system design process - 244

Functional analysis is a universal methodology, which is able to support NPP and I and C design process. The paper proposes an approach to combining the functional analysis process and I and C design process into a single I and C functional design process. Operation of a system is considered as a set of functions providing not only the achievement of goal, but also keeping material and energy balances. A notation for graphical representation of semantic network (hierarchy of functions) describing a system is suggested. This network includes four levels, namely functional goals, abstract functions level, process functions level, and equipment level. The elements of the hierarchy are connected by three kinds of relationships, namely: parent-child, main-support, and cause-consequence. Then the semantic network (hierarchy) of functions is converted into a functional network which looks as a directed graph. This graph is used for development of control algorithm and for building of instrumentation and control functions hierarchy. Farther analysis of process functions reveals a set of criteria indicating real-time status and effectiveness of functions. The criteria are used for allocation of instruments and actuators. The process of functional analysis and design is illustrated in application to a simple heat generation system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

DEMONSTRATION OF MULTI-PHYSICS EVALUATION OF FUEL ROD BURST PROBABILITY WITH BURNUP EXTENSION UNDER LBLOCA CONDITIONS

Extending the fuel discharge burnup level, e.g., from the current limit of rod averaged discharge burnup limit of 62 GWD/MT to a proposed new limit of 75 GWD/MT, can provide significant economic benefits to the current fleet of operating light water reactors (LWRs). It allows for longer operating cycles and improved resource utilization. The major economic gain of longer operating cycles is attributable to the increased capacity factor resulting from decreased refueling time as a fraction of total operating time, as well as fewer assemblies to be discharged for a given amount of energy produced. The main licensing challenges for higher burnup fuel are to ensure fuel rod safety under design basis accident conditions, especially under large-break loss-of-coolant accident (LBLOCA) and reactivity insertion accident (RIA). In this work, two-year cycle core design for a typical 4-loop pressurized water reactor (PWR) is performed with enrichment increased up to 6% and burnup extended to 75 GWD/MT. The fuel rod burst potential evaluations under large-break loss-of-coolant accident (LBLOCA) conditions are subsequently performed using the multi-physics best estimate plus uncertainty analysis framework LOTUS (LOCA Toolkit for the U.S. LWRs) and the preliminary results are presented.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

A data-driven framework for error estimation and mesh-model optimization in system-level thermal-hydraulic simulation

We report over the past decades, several computer codes have been developed for simulation and analysis of thermal-hydraulics and system response in nuclear reactors under operating, abnormal transient, and accident conditions. However, simulation errors and uncertainties still inevitably exist even while these codes have been extensively assessed and used. In this work, a data-driven framework (Optimal Mesh/Model Information System, OMIS) is formulated and demonstrated to estimate simulation error and suggest optimal selection of computational mesh size (i.e., nodalization) and constitutive correlations (e.g., wall functions and turbulence models) for low-fidelity, coarse-mesh thermal-hydraulic simulation, in order to achieve accuracy comparable to that of high-fidelity simulation. Using results from high-fidelity simulations and experimental data with many fast-running low-fidelity simulations, an error database is built and used to train a machine learning model that can determine the relationship between local simulation error and local physical features. This machine learning model is then used to generate insight and help correct low-fidelity simulations for similar physical conditions. The OMIS framework is designed as a modularized six-step procedure and accomplished with state-of-the-art methods and algorithms. A mixed-convection case study was performed to illustrate the entire framework.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗