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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 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

SPC-2918 Rev 0 MARVEL Test Loop Flowmeters

This specification contains the requirements for the design, manufacture, testing, and delivery of magnetic flowmeters to be installed on the MARVEL Primary Coolant Apparatus Test (PCAT) system. The PCAT system is a heated test loop which will be used to simulate the coolant system of a proposed sodium or NaK cooled microreactor. The system has four identical cooling loops. Installation of flowmeters on all four loops is anticipated at this time, however, the number of flowmeters may be reduced by the Project. Note that the number of flowmeters described herein is not a commitment to a particular number of procured units. The number of flowmeter units procured shall be identified on the purchase order. The MARVEL reactor is still under design, so the PCAT flowmeter design may not be directly transferrable to the reactor. It is intended that the design and development effort for the PCAT flowmeters be leveraged to the extent practical for the reactor flowmeter design.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

SPC-70734 Rev 1 MARVEL Primary Coolant System (PCS) Fill and Eutectic NaK Procurement

This specification contains the requirements for the filling the Microreactor Applications Research Validation and Evaluation (MARVEL) reactor Primary Coolant System (PCS) with eutectic sodium-potassium (NaK). The subcontractor will design and fabricate a NaK filling system (referred to herein as the PCS Fill System), the requisite volume of high purity NaK, and perform filling operations in conjunction with the Contractor Operations. The MARVEL reactor system will be operated by the Contractor, and the PCS fill system will be operated by Subcontractor personnel.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

ECAR-7813 Rev 0 Secondary Coolant Cover Gas System Stirling Engine Tube Rupture Analysis, Insulation Block Analysis, and Labyrinth Seal Design

The purpose of this ECAR is to document compliance with best engineering practices for the Secondary Coolant Cover Gas System of the MARVEL microreactor which includes a Stirling Engine helium tube rupture analysis, an analysis to verify the strength of the insulation blocks, and the design of the Labyrinth Seal.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-6076 Rev 0 MARVEL Reactor End of Life Enveloping Radiological Source Term

As part of the Microreactor Applications Research Validation and Evaluation Project (MARVEL) design effort, an enveloping end of life radiological source term for the reactor fuel and sodiumpotassium (NaK) eutectic coolant is calculated in support of the MARVEL preliminary hazard evaluation. The source term calculation assumes slightly more than two-years of uninterrupted operations at a power of about 120 kW for a total energy of 7,600,000 MJ which exceeds the planned two-year operation at 100 kW. The Reactor modeling was performed with MCNP6 Version 1.0 and the depletion was performed using the COUPLE and ORIGEN modules from the SCALE Version 6.2 code package.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2022

ECAR-6415 Rev 0 CFD Analysis of the Primary Coolant Apparatus Test Core with Wire Spacers for Heterogeneous Power Distributions

Primary Coolant Apparatus Test (PCAT) is a thermal-hydraulic test loop for Microreactor Applications Research Validation and Evaluation (MARVEL) reactor. The heater elements were manufactured within a straightness tolerance for the designated gap thickness of 1.0 mm, but several elements have been distorted out of tolerance during the welding process. The distortion of heater rods can lead to flow maldistribution and excessive hotspot where the gap between two heater rods is too tight. In order to secure the heater-to-heater gap in an acceptable range, wire spacers are adopted for the PCAT. The wire spacers affect the thermal-hydraulic performance of the PCAT because they enhance the flow mixing among the flow subchannels in the core region. The objectives of this ECAR is to evaluate the effects of the wire spacers on the thermal-hydraulic performance of the PCAT core region by assuming several heterogeneous power distributions.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

ECAR-6580 Rev 0 ASME Section III, Division 5 Analysis of the MARVEL PCS and GVS Top Corner

The purpose of this Engineering Calculations and Analysis Report (ECAR) is to document the structural evaluation for part of the Microreactor Applications Research Validation and Evaluation (MARVEL) Primary Coolant System (PCS) and Guard Vessel System (GVS). For the PCS this relates specifically to the Distribution Plenum (DP), The Intermediate Heat Exchanger (IHX), and the Upper Downcomer (UD). For the GVS, this will include only the top corner that is machined into the DP Top Plate. These components will be evaluated using the 2021 version of ASME Section III, Division 5 [1] which is a design code that governs the construction of vessels, storage tanks, piping, pumps, valves, supports, core support structures and nonmetallic core components for use in high temperature reactor systems and their supporting systems. Materials at high temperature are subject to creep and fatigue mechanisms that require additional analyses that aren’t covered in ASME Section III, Division 1 rules. Division 5 contains two approaches: elastic or inelastic, however, additional Code Cases specific to Division 5, allow for an Elastic-Perfectly Plastic (EPP) approach. The components analyzed in this ECAR will use a combination of the elastic and EPP approach. Only Design and Service Levels A and B are evaluated in this ECAR. Service Level D evaluations for the entire PCS are documented in ECAR-6564, “MARVEL Project Primary Coolant System Pressure Vessel Stress Documentation” [2] and for the entire GVS are documented in ECAR-6574, “MARVEL Guard Vessel System FEA and ASME Analysis” [3]. The Design and Service Level A and B analyses for the PCS Downcomer piping and Core Barrel are documented in ANL-24/36, "Engineering Calculations and Analysis of the Core Barrel and Downcomer Piping in the MARVEL PCS” [4], and the GVS (except for the top corner) is in ECAR-6574, “MARVEL Guard Vessel System FEA and ASME Analysis.” See Section 2.0 for more detail on the analysis boundaries.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-6589 Rev 0 Reflector Support Structure Analysis

The purpose of this ECAR is to document compliance with best engineering practices, ANSI/AISC N690 and the seismic requirements of IBC-2015 for the Reflector Support Structure of the MARVEL microreactor.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-7365 Rev 0 MARVEL Over Pressure Protection Report

This ECAR encompasses and documents compliance to the pressure safety requirements of the Microreactor Applications Research Validation and Evaluation (MARVEL) Reactor Primary Coolant Boundary (referred to herein as the Primary Coolant System, or PCS, which includes the outer boundary of the Central Insurance Absorber (CIA)) as a Class A system (equivalent to a Section III Division 1 Class 1 system) for high temperature reactors subjected to elevated temperature service.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-7428 Rev 0 MARVEL Upper Confinement Structure Airflow Analysis

This ECAR documents airflow values calculated for the Upper Confinement Structure (UCS) design, which is a subsystem of the Microreactor Applications Research Validation and Evaluation (MARVEL) reactor. These airflow values are calculated based on current design parameters and heat removal requirements and are needed to verify the UCS can adequately maintain a target design temperature.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-7448 Rev 0 MARVEL Post-Shutdown Dose Analysis in the Upper Confinement

This report documents a bounding post-shutdown dose rates over time in the upper confinement area and above the top of the TREXc Pit Shield at the end of two years of operation of the MARVEL microreactor. The dose rates take into consideration the doses from the core fission products, the sodium-potassium coolant (NaK), the steel components, the Stirling engines, and the gallium eutectic (EGaInSn) secondary coolant. The MCNP model in this analysis is based on the 90% design of the MARVEL reactor from ECAR- 7447. Neutron fluence used as inputs to this analysis were calculated in ECAR-7447.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

ECAR-7932 Rev 0 Large Eddy Simulation of MARVEL Reactor Core Subchannel to Evaluate Model Uncertainty of Reynolds-Averaged Navier-Stokes Equation Based Computational Fluid Dynamics Analysis

In the previous work (ECAR-7210), the peak cladding temperature of the MARVEL microreactor has been evaluated by steady-state Reynolds-Averaged Navier-Stokes (RANS) based computational fluid dynamics (CFD) simulations. Although numerical uncertainties of RANS-based CFD simulations has been assessed in ECAR-7210, the model uncertainty of RANS turbulence models must be investigated to resolve the issues related to inaccurate prediction of turbulent heat flux and flow pulsation in a tight lattice rod bundle using the steady-state RANS simulations. Consequently, this ECAR conducted a high-fidelity CFD analysis utilizing Large Eddy Simulation (LES) to generate reference data and investigated the model uncertainty of RANS-based CFD simulations.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

Web-based Preprocessing and Visualization of 3D FIB Tomography Data for Nuclear Fuel Characterization

Three-dimensional (3D) focused ion beam (FIB) tomography enables reconstruction of internal nuclear fuel features that can't be fully evaluated through surface imaging alone. This capability supports characterization of fuel constituents and defects under thermal and irradiation conditions relevant to microreactor development. However, large tomography datasets can create data-handling, loading, and visualization challenges, especially when image-stack preparation and file conversion must be completed with separate tools. The Computational Ultraspatial Tomography Toolkit for High-Resolution Object Analysis Tools (CUTTRHOAT) is an open-source web application being developed to display FIB tomography datasets available through the Nuclear Research Data System (NRDS). The current alpha version requires prepared HDF5 datasets and has limited integrated data-preparation capabilities. This project improves CUTTHROAT by adding dataset-folder selection, automatic input detection, dataset scanning, missing-slice identification, blank-slice insertion, and image-stack-to-HDF5 conversion. Two applications will be compared: the baseline CUTTHROAT alpha workflow and the updated application containing the integrated data-handling and preprocessing functions. Evaluation will consider dataset detection accuracy, conversion success, loading time, rendering responsiveness, application stability, and user interaction. Preliminary results demonstrate successful loading of existing HDF5 files and converted image stacks, while testing also identified performance reductions caused by excessive blank-slice generation. The updated workflow reduces reliance on external preparation tools and supports more direct movement from image stacks to color-code 3D visualization. Future work includes refining missing-slice handling, integrating additional preprocessing functions, like a denoising feature, parsing TIFF metadata for automatic voxel scaling, and adding manual X, Y, and Z voxel-spacing inputs for PNG and JPEG.

36 - MATERIALS SCIENCE↗
DOE OSTI2020

High Temperature Irradiation Resistant Thermocouples

The High Temperature Irradiation Resistant Thermocouple (HTIR-TC) is the world’s leading temperature sensor for reactor fuel experiments and material test reactors (MTR), in general (e.g. Advanced Test Reactor, Massachusetts Institute of Technology Nuclear Reactor, etc.). Recently commercialized, the HTIR-TC provides real-time centerline and fuel cladding temperature measurements of reactor fuel experiments during irradiations. This provides direct information about fuel temperatures during Gen IV reactor, small modular reactor (SMR), and microreactor vetting processes. Further, the HTIR-TC recently won the 2019 R&D100 award under category: Analytical/Test.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2026

TRISO Spent Nuclear Fuel Recycling or Waste Reduction Using SRNL Vapor Digestion Technology – 25635

There is a renewed interest in advanced reactors, including high-temperature gas cooled reactors (HTGRs). Tri-structural isotropic (TRISO) fuel is being used in many HTGR designs, whether as SMRs or microreactors. However, TRISO-based HTGRs discharge the largest volume of used fuel per megawatt-hour of energy produced compared to other reactors. An order of magnitude reduction or more in the volume of SNF could be realized if the TRISO particles were separated from the graphite moderator and the carbon dispositioned as LLW. The Savannah River National Laboratory (SRNL) has a patented technology readiness level (TRL) 4/5 vapor digestion process for separating nuclear-grade graphite from HTGR SNF. The SRNL process is based on the reaction of NOx species with carbon to form CO2. Because NOx species are several orders of magnitude more reactive with graphite than oxygen, the process can operate at lower temperatures with uncrushed HTGR pebbles or prismatic blocks. Because the fuel elements do not need to be crushed and the graphite is digested using a vapor-based process, the potential for damaging the TRISO particles is much reduced. The DOE Office of Technology Transitions (OTT) is funding SRNL and the University of South Carolina at Columbia to close certain gaps that exist within the technology which impede its direct application to the processing of commercial TRISO-based SNF coming from HTGR advanced reactors.

Pierce, Robert [Savannah River National Laboratory↗
DOE OSTI2020

Powering DNA strand-displacement reactions with a continuous flow reactor

Living systems require a sustained supply of energy and nutrients to survive. These nutrients are ingested, transformed into low-energy waste products, and excreted. In contrast, synthetic DNA strand-displacement reactions typically run within closed systems provided with a finite initial supply of reactants. Once the reactants are consumed, all net reactions halt and the system ceases to function. In the study, we ran DNA strand-displacement reactions in a continuous flow reactor, infusing fresh reactants and withdrawing waste, enabling the system to dynamically update its outputs in response to changing inputs. Running DNA strand-displacement reactions inside of continuous flow reactors allows the system to be re-used for multiple rounds of computation, which could enable the execution of more elaborate information processing tasks, including single-rail negation and sequential logic circuits.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2021

Process Heat for Chemical Industry

Process heating with nuclear energy can reduce greenhouse gas emissions by reducing combustion of fossil fuels in fired heaters as steam boilers. Light water reactors can replace the majority of steam duties used by industry; however, high temperature processes such as steam methane reforming require advanced high temperature reactors. Here, guidance on matching the scale of nuclear reactors with specific industries is provided. Principles of heat transport, temperature boosting, and substitution for hot combustion gases are also discussed in this section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Reproducible benchmark for the SNAP 8 experimental reactor at dry conditions

Here, this work provides fully reproducible benchmark models of the Systems for Nuclear Auxiliary Power (SNAP) 8 Experimental Reactor. Validation benchmarks of the criticality configuration experiments under dry conditions with no coolant are presented. In addition, relevant reactivity worth experiments, and measurement of power distributions are validated and presented here. Discrepancies between modeled and experimental results are at most 110 pcm for fuel and poison worths, and critical configurations that do not manipulate control elements are within 50 pcm. Larger discrepancies found in reflector element worth experiments are due to experimental methodology which was noted as being simplified due to limited calculational capabilities at the time, while discrepancy in the control element manipulated critical configuration is due the absence of structural components for modeling simplification. Power distributions closely follow what was observed in experiment with expected peaking in reflecting elements. All models are well documented with cited references describing and justifying assumptions, simplifications, and adjustments to reproduce the models with any nuclear safety codes; model inputs and outputs are stored in the snapReactors GitHub repository.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Reproducible benchmark for the SNAP 8 experimental reactor at operating conditions

This work presents fully reproducible multiphysics benchmark models of the Systems for Nuclear Auxiliary Power (SNAP) 8 Experimental Reactor at operating conditions with coolant flow. Wet experiment (with coolant, at power) validation benchmarks are presented using both deterministic (Serpent-Griffin) and Monte-Carlo (OpenMC-Cardinal) multiphysics frameworks coupled with thermal-hydraulic solvers in MOOSE. Reactivity coefficient measurements including fuel temperature, isothermal temperature, and power coefficients show good agreement with experiments, with discrepancies within experimental uncertainty. Reactivity worth experiments for coolant, samarium, and xenon poisoning are reproduced with differences under 200 pcm. Comparison between Serpent-Griffin and OpenMC-Cardinal frameworks reveal multiphysics coupling introduces positive reactivity effects (100-200 pcm) compared to uniform temperature and density fields at nominal operating conditions. Comparison between Serpent-Griffin and reference Serpent solution shows that power distributions maintain consistent radial and axial peaking behavior. All models, assumptions, thermophysical and thermomechanical properties, and material definitions are thoroughly documented with cited references; model inputs and model generating scripts are stored in the snapReactors GitHub repository.

SNAP↗