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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 1,009 records · Page 56
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Single Primary Heat Extraction and Removal Emulator (SPHERE) Power Transient Testing

For heat pipe cooled microreactor development, it is essential to understand the characteristics of heat pipes and how they function under a wide range of operating conditions. Modeling the startup and shutdown of heat pipe cooled microreactors has been a major challenge for the modeling code. Validating these parameters within the code is key for the development and licensing of heat pipe cooled microreactor designs. Idaho National Laboratory (INL) has completed testing on power transients of a heat pipe with a range of operating conditions. The resulting temperature profiles from this testing can be utilized to aid in the validation of the startup and shutdown portions of the heat pipe modeling code.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Enabling a Physical Twin for Control Methods Evaluation

Advanced nuclear reactors play an important role in the energy future of the United States and the rest of the world. They are designed and operated based on a different model than that of the current operating fleet, thus enabling deployment in remote locations and allowing for safe semi-autonomous or autonomous operations. Such characteristics require the development of a new reactor control paradigm. A significant factor in the development of control technologies and methods is integration of the various technologies and methods with each other and with hardware (both reactor system hardware and control hardware). A recent workshop on control of advanced reactors identified the lack of a flexible, expandable software/hardware infrastructure to enable such integration as a key gap. A previous phase of the current effort involved developing and demonstrating the Control and Optimization Modular Modeling Application for Nuclear Deployment (COMMAND) platform, which is capable of integrating autonomous-control-enabling technologies and methods, without the constraints imposed by existing software solutions. This platform will enable advanced reactor developers to deploy and test any developed technologies and methods by employing a common framework, and to couple them with their own models and hardware. The present phase of this effort entails using the Microreactor Automated Control System (MACS) platform, which was developed by the U.S. Department of Energy (DOE) Microreactor Program, to serve as a control method testbed. MACS can be used by advanced reactor developers to integrate their control related research activities with any reactor system. For the present effort, MACS was customized to mirror Idaho National Laboratory (INL)’s Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor, and COMMAND was leveraged to enable MACS to emulate the physics of MARVEL, thus positioning MACS as a physical twin of MARVEL.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Parametric and Sensitivity Analysis of a Steam Generator Model Using Python and Machine-Learning Tools

For this study, we used Python and machine-learning tools to perform a comprehensive parametric and sensitivity analysis on a steam generator (SG) model. (The Python model was based on a previously completed MATLAB framework for the Holtec SMR-160 SG.) We investigated the influence of various input parameters (e.g., heat transfer coefficient [HTC], Nusselt number, and heat exchanger effectiveness) on the system’s output. With machine-learning tools such as the Risk Analysis Virtual Environment (RAVEN), which was developed at Idaho National Laboratory, we were then able to perform an automated analysis of the SG inputs’ effect on the HTC. The analysis results give valuable insights into the performance and optimization of SG systems. We found the inlet mass flow rate (MFR) to have the greatest impact on the HTC, followed closely by the inlet temperature, and then pressure. Shifting of the input parameters causes the location of the maximum HTC along the SG length to change incrementally. The cold leg (CL) MFR was also found to impact the HTC magnitude as well as the location of the maximum HTC. At between 0.4–0.9 of the total SG length, the input parameters experience maximum impact on the HTC, leading us to suggest that sensors be efficiently placed on the SG so as to closely and effectively monitor thermal-hydraulic properties during reactor operation. We also found that the sensitivity data calculated manually agrees with the RAVEN – based data, confirming the same range of maximum sensitivity. However, the RAVEN-based analysis showed that cold leg pressure and hot leg temperature have a greater impact on the heat transfer coefficient than the mass flow rate, implying that a manual sensitivity study taking only two samples is not accurate.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2022

Code of Record: DOE Standard Canister (DOESC)

The United States Department of Energy (DOE) Spent Nuclear Fuel (SNF) Packaging Demonstration seeks to develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE SNF for road-ready dry storage (RRDS); and establish the processes that will be used in a future production facility. The Packaging Demonstration will utilize the DOE Standard Canister (DOESC) for packaging select DOE-managed SNF types for interim storage, transportation and disposal as part of RRDS. Placing DOE-managed SNF into RRDS is part of the strategic framework for SNF on the Idaho National Laboratory (INL) site. To comply with DOE, INL and Nuclear Regulatory Commission (NRC) requirements, this Code of Record establishes the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 3 as the Code that will govern DOESC and internal support structure constructioni, with certain clarifications. This Code of Record establishes a rationale for proceeding without construction certification of the DOESC (i.e., “N-stamping) and the extent to which a Registered Professional Engineer is required for the DOE Spent Fuel Packaging Demonstration. Given the (i) standard industry practice to pursue independent licensure of commercial storage casks and transportation packages by the NRC in lieu of ASME certification and (ii) guidance provided by 10 CFR 830, DOESC construction activities need not be certified (i.e., “stamped”) to the ASME BPVC. However, additional quality assurance requirements will apply as outlined in this Code of Record.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Scaling Outlooks on Integrated Energy Systems

The report provides research summaries of two developed theories: data projection to enable multi-scale facility digital interactions and introducing the Evaluation Model Development and Assessment Process (EMDAP) approach to Integrated Energy Systems (IES) design. For the data projection, a two-system scaling case including interactions between thermal storage and hydrogen production is provided. The corresponding facilities are the thermal energy distribution system (TEDS) facility and high temperature steam electrolysis (HTSE) facility at the Dynamic Energy Transport and Integration Laboratory (DETAIL) in Idaho National Laboratory (INL). For the EMDAP adaption to system design, a modified process of the EMDAP was developed and applied to indirect two-tank heat storage system design. System phenomena were identified, ranked, and scaled to discover helpful key relations that would determine the ideal component geometry.

08 HYDROGEN↗
DOE OSTI2023

Prioritizing Off-Gas Metrics: A Guide for Comparable Off-Gas Capture Testing

The Material Recovery and Waste Form Development (MRWFD) off-gas team had a workshop, hosted by Idaho National Laboratory (INL), to align goals and expectations for off-gas research. The workshop included team members from four national laboratories. The workshop focused on defining distinct R&D phases with specific metrics, outlining standard test and measurement protocols for Iodine and Krypton/Xenon sorbents, brainstorming approaches to future disruptive technologies, and recognizing parameters with more inherent risk, requiring more rigorous evaluation. This report will serve as a guide for future off-gas work. Its purpose is to foster efficient collaboration across diverse research facilities and invite direct comparison of materials and results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Selection of Components for the Remote, Canister-Monitoring System

This report documents the selection of components as part of the efforts initiated to develop a remote, canister-monitoring system (RCMS) for determining and monitoring environmental conditions within dry fuel storage canisters containing aluminum clad spent nuclear fuel (ASNF) at the Idaho National Laboratory (INL). These efforts are in support of Department of Energy Office of Environmental Management (DOE-EM) investigations into technical issues associated with extended dry storage (50+ years) of ASNF. The capability to establish and monitor the performance of ASNF in situ provides the opportunity to (1) evaluate the appropriate technologies for monitoring, (2) collect canister environment conditions as soon as possible, (3) verify and validate current laboratory-based study results and analytic modeling approaches, and (4) potentially identify additional dry storage options for ASNF at the INL site. The improved understanding of ASNF behavior gained by performing this work would also contribute to the safety basis for extended dry storage in current and future configurations as well as to provide information for future transportation, conditioning, and disposal of ASNF. The parameters to be monitored by the RCMS include temperature, relative humidity, hydrogen gas concentration and radiation environment (dose). Components are selected based on the expected environmental conditions within the canister and fuel storage area, and the desired performance of the RCMS. Several of these components have been purchased and received. Future activities include component testing and calibration, software development for operation of the integrated sensor module and fabrication of a representative test volume. The results of these tests and changes to desired performance characteristics will inform subsequent activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

FY21 Progress Report for Advanced Re fabrication/Re-instrumentation Capability Development

In support of performing follow on irradiation experiments with previously irradiated materials, the Halden Reactor Project developed unique and state of the art capabilities to refabricate and re-instrument previously irradiated materials. Such materials were used in in-pile tests at the Halden reactor, and out-of-pile tests for example using furnaces as a heat source. The decision to close the Halden Reactor Project results in the loss of this refabrication and reinstrumentation capability. As a result, the United States Department of Energy has determined to develop refabrication and re-instrumentation capability at the unique shielded facilities at Idaho National Laboratory. The development of refabrication capability has been completed and demonstrated. This report focuses on the complementary aspects of re-instrumentation and the progress to date. Halden spent nearly 30 years developing both refabrication and reinstrumentation. Collaboration with Halden is allowing INL to develop this capability much more rapidly. The results include development of the capability to drill annular center holes in ceramic UO2 fuel pellets, development of fuel rod end caps with feedthroughs for centerline instrumentation inside the rodlet, The procurement of both fuel drilling and welding demonstration equipment from Halden, evaluation of surface thermocouple attachments to support better understanding of temperature measurement uncertainties, and finally, the conceptual design of a new shielded enclosure where advanced refabrication and re-instrumentation equipment can be housed. The authors would like to thank the numerous colleagues at INL and Halden who provided support in accomplishing this work. Their support both material and intellectual is invaluable in advancing the state of the art and establishing the capabilities for refabrication and re-instrumentation at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Roadmap for the Completion of the Design and Implementation of a Remote Canister Monitoring System

The Office of Environmental Management within the Department of Energy is currently investigating technical issues associated with extended dry storage (50+ years) of aluminum spent nuclear fuel. As part of these efforts, a remote canister-monitoring system (RCMS) has been proposed to collect the real-time data of the environmental conditions in the canisters used for the dry storage of aluminum spent nuclear fuel in the CPP-603 facility at Idaho National Laboratory (INL). The proposed RCMS would provide the opportunity to evaluate the appropriate technologies for monitoring, to collect canister environment conditions as soon as possible, to verify and validate current laboratory-based study results and analytic modeling approaches, and to potentially identify additional dry storage options for aluminum spent nuclear fuel at the INL Site. The document provides a brief update on the progress that has been made toward deploying the RCMS and will present at a high level, the general system design philosophy behind the project and the key future tasks necessary for the completion of the RCMS system design, fabrication of the RCMS prototype, testing, and eventual deployment of the RCMS, and a tentative timeline for these tasks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Architecture and Component Selection for the Remote Canister-Monitoring System

This report documents the selected system architecture and components surveyed for the development of a remote, canister-monitoring system (RCMS) for determining and monitoring environmental conditions within dry fuel storage canisters containing aluminum-clad spent nuclear fuel (ASNF) at Idaho National Laboratory (INL). These efforts are in support of the Department of Energy Office of Environmental Management (DOE-EM) investigations into the technical issues associated with extended dry storage (50+ years) of ASNF. It is important to resolve questions regarding storage because the ATR is expected to continue producing approximately 100 elements of spent fuel per year for at least another twenty years. The parameters to be monitored by the RCMS include temperature, relative humidity, hydrogen gas concentration, and radiation environment (dose). Components are selected based on the expected environmental conditions within the canister and fuel storage area and the desired performance of the RCMS. Several of these components have been purchased and are currently undergoing component testing to confirm their suitability for use in delivering the RCMS. Additional components may be identified through the course of design and may be selected in the future. Furthermore, changing performance, interfacing, and compatibility requirements and the results of component testing may necessitate revisiting component selection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

20A44-121: Modeling and characterization of α-U to accelerate metallic fuels development [Slides]

This presentation contains a high-level description of the Idaho National Laboratory Laboratory Directed Research and Development (LDRD) project, 20A44-121: Modeling and characterization of α-U to accelerate metallic fuels development. It frames the problem, describes a motivation, outlines the research plan involving multiscale modeling and experiments, discusses impact and details research developments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

ECAR-6584 MARVEL Reactor Support Frame Analysis

Idaho National Laboratory (INL) is developing a small microreactor to produce electrical power utilizing a small nuclear core and Stirling engines for power output under the Microreactor Application Research Validation and Evaluation (MARVEL) program. The MARVEL Reactor will be installed in the Transient Reactor Test (TREAT) Facility. Walsh Engineering Services, PC (WES) has contracted with the INL to support the completion of the final design. There are six main systems in the MARVEL, one of which is the MARVEL Reactor System (MRS). The MRS contains a subsystem – the Reactor Support Frame (RSF) – that supports and orients nearly all the MRS and the RSF will be bolted to the TREAT Facility. The subject of this Engineering Calculations and Analysis Report (ECAR) is the frame weldment (see Figure 1), which is the only item that is contained in the RSF subsystem and will be called herein “RSF”.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

ECAR-6585 MARVEL Core Support Structure Analysis: Project #33526

Idaho National Laboratory (INL) is developing a small microreactor to produce electrical power utilizing a small nuclear core and stirling engines for power output under the Microreactor Application Research Validation and Evaluation (MARVEL) program. The MARVEL Reactor (Figure 1) will be installed in the Transient Reactor Test (TREAT) Facility. Walsh Engineering Services, PC (WES) has contracted with the INL to detail the final design. There are six main systems in the MARVEL, one of which is the Fuel and Core System (FCS). The FCS contains a subsystem–the Core Support Structures Subsystem (CSS)–that supports and orients each of the 36 nuclear fuel pins in the MARVEL Reactor System. The FCS (apart from the fixed Beryllium Oxide radial reflectors) is shown in light green and is depicted in more detail in Figure 1 and Figure 2. The main subject of this analysis is the lower grid plate, which is the principal item in the CSS which supports the fuel, see Figure 3. Additionally, the upper grid plate thermal expansion of the upper grid plate is presented in Appendix A.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Fabrication of a Liner Assembly for the MARVEL Microreactor

A liner assembly for an intermediate heat exchanger within the MARVEL microreactor at Idaho National Laboratory has been fabricated using laser-based additive manufacturing. The liner design includes challenging features such as long thin walls, overhangs, and tight geometric tolerances. Two versions of the liner assembly, one full length and the other shortened, were printed from several materials (316L SS, Inconel 625, and Inconel 718) using four powder-based laser printers (AddUp BeAM Modulo 400 DED, EOS 290M LPBF, GE Concept M2, and GE Concept Xline 2000R LPBF) located at the Manufacturing Demonstration Facility of Oak Ridge National Laboratory. Dimensional accuracy, print time, and powder utilization have been evaluated to help assess the feasibility of these AM approaches to liner fabrication. Based on this assessment, a potentially cost-effective pathway for fabrication has been proposed.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Application Guide for the Cyber-Resilient Design Framework for Hybrid Systems

As the energy landscape evolves, hybrid power plants—integrating multiple renewable energy sources (e.g., solar, wind, and battery storage) with the bulk power electric system—play a crucial role in meeting growing energy demands. However, with a hybrid power plant’s increased number of components, complex network connectivity and digitization of controls, these systems face growing cybersecurity risks. To help mitigate these risks, Idaho National Laboratory (INL) presents this application guide specifically designed for operators and systems engineers to evaluate the cybersecurity resilience of their hybrid power plant’s design.

14 SOLAR ENERGY↗
DOE OSTI2024

Coupled SAM/Griffin Model of a Reference Fluoride-Salt-Cooled High-Temperature Reactor for Multi-Physics Simulations

A multi-physics coupled simulation model of a reference pebble bed fluoride-salt-cooled high-temperature reactor (PB-FHR) has been developed with SAM and Griffin computer codes for transient safety analysis. The reference problem of a prototypical reactor design serves as the foundation for the U.S. NRC (Nuclear Regulatory Commission) to verify the adequacy of computer codes and evaluation models for specific reactor types. In this work, the previously developed SAM model for PB-FHR primary loop has been updated for the coupled simulation. The updated SAM PB-FHR model includes a 2-D axial symmetric core region and external core components in 0-D/1-D. In addition to the primary loop, a detailed model of the RCCS (reactor cavity cooling system) is added. The 2-D and 1-D domains are tightly coupled using the single-solve approach developed in SAM. In the pebble bed region, the SAM multiscale explicit pebble model is applied to calculate the pebble and TRISO fuel kernel temperatures. The Griffin model used in this work is based on a model developed at Idaho National Laboratory in collaboration with the U.S. NRC. The Griffin neutronics model and SAM thermal hydraulics model is coupled with the Comprehensive Reactor Analysis Bundle (CRAB or alternately BlueCRAB) application. Both steady-state and transient scenarios are simulated to demonstrate the model's suitability for multi-physics simulations of PB-FHR transients.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Sensitivity Analysis of Irradiated Fueled Experiments using the MOOSE Framework [Slides]

Modeling and simulation (M&S) methods are able to predict uncertainties in experimental parameters (e.g., power and fission density) during irradiation. A shortfall exists in predicting how sensitive some of the parameters will behave during the experimental process. Sensitivity and Uncertainty Quantification (SUQ) is critical in support of qualification and licensing reactor fuels. The application of a method to quantify the uncertainty in these experiments is critical to the prediction of their performance. In this work, we propose the use of a polynomial chaos expansion (PCE) method to quantify the sensitive parameters in these simulations and, in an extension, their experimental surrogates. We propose to perform M&S using PCE uncertainty quantification on a previously irradiated fueled experiment in order to provide a validation case for Griffin and expand its use as a verification and validation (V&V) tool for experiments with a neutronics component. Griffin is an advanced, deterministic neutronics analysis code built using the MOOSE (multiphysics object-oriented simulation environment) framework which can provide state-of-the-art neutronic analysis on M&S of experiments. We will use the stochastic tools module (STM) in MOOSE to provide PCE uncertainty quantification on the proposed experimental setup. Idaho National Laboratory (INL) does not yet have an in-house developed code with V&V approval for experiments performed on-site; this work would provide a necessary addition of support for experiments performed at INL. The Nuclear Regulatory Commission (NRC) has explicitly requested uncertainties in calculated values such as fuel power and burnup, and the development of this capability would benefit the relationship between INL and the NRC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗