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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Development of 17 N as a time-tagged neutron source for calibration of large antineutrino detectors

Large antineutrino detectors hold the potential for remote nuclear reactor monitoring and discovery, which is of great interest for nonproliferation and treaty verification applications. Correlated-particle calibration sources are especially useful for large-volume designs based on detection of inverse beta-decay events, as the timing information of the coincident particles improves background rejection and event reconstruction. 17 N is a promising calibration source since it emits beta-correlated delayed neutrons and can be produced relatively easily using high-energy neutrons. In this work, we examine the feasibility of 17 N as a time-tagged calibration source for large antineutrino detectors. 17 N was produced using a DT neutron generator from a water sample enriched in 17 O, and the observed production rates were found to be consistent with previous measurements of the 17 O(n,p) 17 N reaction cross-section. As a result, time-tagging of 17 N delayed neutrons is demonstrated via beta-neutron coincidence measurements conducted using a specialized beta detector.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

NucMesh: nuclear reactor geometry creation and mesh generation module in NEMoSys

NucMesh is a parameterized geometry and mesh generator for nuclear reactors developed within the Nuclear Energy Modeling System NEMoSys at Illinois Rocstar. NEMoSys is a platform developed for mesh generation, adaptive refinement, and solution verification. NucMesh is implemented to be generalized and extensible with a robust computer aided design engine and multiple mesh generation algorithms for unstructured triangular, quad-dominant, and structured quadrilateral meshing. In this paper, we present the geometric and meshing features of NucMesh. Geometrically objects are constructed bottom-to-top and overlaps are addressed automatically. A sophisticated object tracking algorithm prevents data from being lost for segmented objects. We discuss the primitive objects of circle and polygons that constitute the module and show how they are used with example inputs. Arrays of primitives and arrays of arrays are utilized to build large assemblies of objects. The concept of saved objects is discussed to demonstrate how repetitive objects can be reused easily and augmented in place. Three dimensional meshes can be obtained through mesh extrusion where all materials and side sets are extended to three dimensions. We show that side sets can be defined nearly anywhere within the geometry and can then be applied to the mesh. Finally, example reactor meshes are demonstrated for the Idaho National Laboratory Advanced Test Reactor and Los Alamos National Lab Empire reactor, both of which use control drums that NucMesh handles easily. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

CTF Validation and Verification: Version 4.3

Coolant-Boiling in Rod Arrays- Two Fluids (COBRA-TF) is a thermal/hydraulic (T/H) simulation code designed for light water reactor (LWR) analysis. It uses a two-fluid, three-field (i.e., fluid film, fluid drops, and vapor) modeling approach. Both subchannel and 3D Cartesian forms of nine conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and has been used and modified by several institutions over the last several decades. COBRA-TF is also used at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Modeling Group (RDFMG) and has been improved, updated, and subsequently became the CTF. One part of the improvement process includes validating the methods in CTF. This document seeks to provide a certain level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model—rod bundle geometries with operating conditions that are representative of prototypical pressurized water reactor (PWR)s and boiling water reactor (BWR)s in both normal and accident conditions. This is done by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

CTF Validation and Verification: Version 4.4

Coolant-Boiling in Rod Arrays- Two Fluids (COBRA-TF) is a thermal/hydraulic (T/H) simulation code designed for light water reactor (LWR) analysis. It uses a two-fluid, three-field (i.e., fluid film, fluid drops, and vapor) modeling approach. Both subchannel and 3D Cartesian forms of nine conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and has been used and modified by several institutions over the last several decades. COBRA-TF is also used at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Modeling Group (RDFMG) and has been improved, updated, and subsequently became the CTF. One part of the improvement process includes validating the methods in CTF. This document seeks to provide a certain level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model—rod bundle geometries with operating conditions that are representative of prototypical pressurized water reactor (PWR)s and boiling water reactor (BWR)s in both normal and accident conditions. This is done by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

The Release of the Simplified Radionuclide Transport (SRT) Code (V.2.1)

Reactor licensing centers on the protection of the public and environment from the inadvertent release of radioactive material. Therefore, mechanistic source term analysis, or the realistic evaluation of radionuclide transport from the source to the environment for specific transient scenarios, is vital to reactor licensing efforts. Developed to resolve a gap in mechanistic source term modeling capabilities for sodium fast reactors (SFRs), the Simplified Radionuclide Transport (SRT) code created by Argonne National Laboratory (Argonne) is now utilized by advanced reactor vendors, universities, and research institutions to support a multitude of design, licensing, and research efforts. Recently, SRT version 2.1 was released, which includes improvements to code models and verification and validation (V&V) suite to support the SRT user community. The following work provides an overview of the improvements made as part of the release of SRT version 2.1. This effort is supported by the U.S. Department of Energy Office of Nuclear Energy (DOE:NE) Advanced Reactor Technologies (ART) Fast Reactor Program (FRP), as part of the program’s support of national laboratory design and safety analysis computer codes utilized by the fast reactor industry. The expansion of SRT code capabilities and improvements to code V&V associated with version 2.1 are in response to user requests and lessons learned from recent source term analyses performed by Argonne and advanced reactor vendors. They also align with the evolving role of SRT, from research and development tool to software utilized for reactor licensing calculations. The report is structured in alignment with the code improvements, as outlined in Figure 1-1. Section 2 provides background information on SRT, including its history, capabilities, and utilization. Section 3 details new code capabilities as part of version 2.1, while Section 4 focuses on the expansion of the code’s V&V suite. Lastly, Section 5 provides a summary and discussion of next steps.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

A custom high-stability pulse generator for the test of a novel in-situ magnetic sensor developed to detect radiation damage in undulators

High-performance analog-to-digital converters (ADC) have been widely applied in many areas of science. For instance, magnetic field measurements based on the Faraday’s induction law require high-precision voltmeters to measure induced voltages. In this study in the context of free-electron lasers technology, the Magnetic Measurements Team at SLAC National Accelerator Laboratory proposed a novel in-situ radiation damage detection system (RDDS) for detecting small field variations in undulators. The system measures the flux change in a flexible printed-circuit coil attached to the magnet array during the undulator gap movement. The gap movement changes the magnetic field, which induces a voltage signal in the coil that is measured and integrated with an ADC. Although the system is capable of detecting relative flux changes better than 100 ppm, drift in the ADC’s gain or offset can cause apparent changes in the relative flux. This paper describes the first attempt to developed a high-precision verification circuit to perform ADC testing in the framework of the novel RDDS. The circuit generates a reference voltage pulse with a voltage–time integral relative precision better than 50 ppm for a few hundreds of mVs — the typical order of magnitude measured with the RDDS. The circuit’s design combines a fast and precise switch with a low-noise voltage reference. Long-term measurements allowed statistical analysis and showed that averaging the voltage–time integral of ten pulses gives the required 50 ppm stability. Moreover, reproducibility tests confirmed that the circuit’s output is invariable under small power supply instabilities and equipment shutdown. Instruments and applications designed to quantify the magnetic field by integrating voltage signals may use the pulse generator proposed in this paper for verification purposes.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Application: Acceptance Testing Report

AS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over sixty years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. To support the dedication effort, this report has been generated to provide a detailed description of the available verification testing. The verification testing presented in this report captures functionality testing, focusing mainly on the testing of specific functions and algorithms for accuracy and precision of output, and interface testing, focusing mainly on the testing of critical input parameters and their valid ranges. Although SAS was developed to support the analysis of any liquid-metal-cooled nuclear reactor, the testing described in this document primarily focuses on the verification of SAS capabilities as they relate to a generic pool-type Sodium Fast Reactor (SFR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Applications (Acceptance Testing Report)

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal- cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over fifty years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. To support demonstration of software pedigree and confirm key functional requirements, this report has been generated to provide detailed verification of the software. Although SAS was developed to support the analysis of any liquid-metal-cooled nuclear reactor, the acceptance testing described in this document focuses on the verification of SAS capabilities as they relate to pool-type Sodium Fast Reactors (SFRs). This report includes documentation of the test problem definition, analytical solution(s), computational solution(s), acceptance criteria, comparisons of analytical/computational solutions, and determination of acceptance of the computational solution(s). Deviations from acceptance criteria are noted.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

SAS4A/SASSYS-1 Verification Testing for Sodium Fast Reactor Applications: Acceptance Testing Report

SAS4A/SASSYS-1 (SAS) is a simulation tool used to perform deterministic analyses of anticipated events as well as design basis and beyond design basis accidents for advanced liquid-metal-cooled nuclear reactors. With its origin as SAS1A in the late 1960s, the SAS series of codes has been under continuous use and development for over forty-five years and represents a critical investment in safety analysis capabilities for the U.S. Department of Energy. To support demonstration of software pedigree and confirm key functional requirements, this report has been generated to provide detailed verification of the software. Although SAS was developed to support the analysis of any liquid-metal-cooled nuclear reactor, the acceptance testing described in this document focuses on the verification of SAS capabilities as they relate to pool-type Sodium Fast Reactors (SFRs). This report includes documentation of the test problem definition, analytical solution(s), computational solution(s), acceptance criteria, comparisons of analytical/computational solutions, and determination of acceptance of the computational solution(s). Deviations from acceptance criteria are noted.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Flow instabilities in helical-coil steam generators for small modular reactors: A review

Here, this study covers the research and discoveries in two-phase flow-boiling instabilities available in the literature—specifically for a helical-coil steam generator (HCSG), including experimental findings, theoretical research, computational models, and system code analyses—supporting research and development of representative small modular reactors (SMRs). Like other new and advanced reactor systems, water-cooled SMRs require experimental data from both integral and separate thermal-hydraulics test facilities for the verification and validation (V&V) of the computational models and computer codes in order to design and obtain regulatory approval. The complex dynamics of two-phase flow-boiling instabilities includes flow regimes physics phenomena, flow-channel geometries, heat-transfer behavior, and interactions among the solid–liquid-gas within the system boundary, all of which are pivotal for understanding the design and operational challenges of SMRs. This study focuses on identifying the relevant knowledge gaps on boiling instabilities—specifically for a HCSG—and provides insights about future research direction optimizing the transport of thermal energy, mass-flow rates, and boundary conditions that ensure the adequate heat-transfer performance, operational stability, and safety associated with SMR systems.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2021

Alignment tolerance analysis for divided-pulse nonlinear compression

We present an analytic model that describes the output pulse after Kerr-based spectral broadening with divided-pulse nonlinear compression that includes errors in unequal pulse division, birefringent plate retardance, and thermal drift. The model shows that alignment tolerances become impractically tight at high levels of nonlinearity and that the angle of incidence on the birefringent plates can be utilized as a compensator to loosen those tolerances. We present experimental verification of the model, which is expected to be a fast and flexible tool to design future divided-pulse nonlinear compression systems.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

FY23 Status of Quality Assurance Plan for Out-of-Pile Test Data

The DOE Advanced Reactor Technology program has supported recovery and preservation of legacy metallic fuel data collected as part of the US fast reactor program, recognizing it as essential to development and licensing activities for advanced fast reactors. Databases were established as organized collections of experimental records and data generated from in-pile experiments at EBRII, FFTF, and TREAT as well as related out-of-pile examinations of irradiated fuels. The Out-of-Pile Transient Database (OPTD), includes records of over 150 out-of-pile furnace transient tests on metallic fuels conducted at Argonne’s Alpha-Gamma Hot Cell Facility to evaluate their transient performance and characterize fuel/cladding interaction. The database is accessible to registered users from US universities, laboratories, and nuclear industry. Because the data in OPTD has not been formally qualified, its applicability and ability to support licensing activities is limited. This report outlines progress and plans to quality assure data in OPTD, maximizing its impact for model validation and verification as well as qualification of fuels for safe and effective use in advanced reactor designs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

HydroChrono: An Open-Source Hydrodynamics Package for Project Chrono

In this paper we present the development and verification of HydroChrono, a hydrodynamics package for the Project Chrono physics engine. This package includes the implementation of hydrodynamics equations, the added mass for multibody systems, the development of I/O functions as well as a Python API, and comparison against standard reference cases and other existing tools. HydroChrono provides a flexible, fully open-source solution for simulating wave energy converters (WECs), floating offshore wind turbines (FOWTs) platforms, and other hydrodynamic systems. Here we show, via comparisons with existing tools for benchmark verification cases, that HydroChrono accurately models hydrodynamic forces - making it a useful tool for the design and optimization of these systems. Additionally, the integration of HydroChrono with Project Chrono offers access to finite element modeling capabilities and high-fidelity modelling - with Chrono's existing coupling to CFD and SPH codes. This provides numerical modelers with a multifidelity simulation framework for designing and validating these systems. The development of HydroChrono provides a new, open-source solution for simulating hydrodynamic systems. Its compatibility with other simulation tools enables a more streamlined and efficient design process, advancing the field and providing new opportunities for innovation in this area.

BEM↗
DOE OSTI2020

Validation of an Advanced High-Pressure PEM Electrolyzer and Composite Hydrogen Storage

One of the challenges associated with transitioning our nation’s transportation economy to any alternative fuel is bringing fuel production, storage, transportation, and end-customer delivery infrastructure online. There are challenges associated with planning and implementing hydrogen fueling stations at locations relevant to the parallel effort of bringing the related transportation vehicles into the market. There are additional challenges associated with the operation of hydrogen fueling stations and reducing operating costs through energy reduction. Through calculation and design efforts it was concluded that elevating the pressure of a PEM cell stack and further developing hydrogen processing equipment that support elevated pressures while maintaining similar characteristics to their current production counterparts will yield in energy savings and by extension cost reductions. In this report, the design of a 57-bar (827-psig) electrolysis system and electrolysis cell stack is described. Furthermore, the fabrication, component verification, and integrated testing are summarized. Prior to this development effort, Proton had demonstrated thousand of hours of durability on 30-bar (435-psig) differential pressure PEM electrolysis hardware. Proton has sold and installed many of these 30-bar devices at customer sites. Through this project, Proton has developed and tested 57-bar PEM electrolysis, thereby extending the proven capability of PEM electrolysis as a hydrogen generation and electrochemical compression device. A design and validation effort to arrange hydrogen processioning equipment for refueling OEM vehicles was also undertaken within this project. By arranging equipment within two 6-meter ISO shipping containers and further developing NFPA 2 Hydrogen Technologies Code to support containerized hydrogen equipment, efforts to site, install, and commission, and approve hydrogen fueling stations were reduced and now provide a groundwork for future installations.

08 HYDROGEN↗
DOE OSTI2021

Design Considerations for the Field Use of International Nuclear Safeguards Technology

The process of developing and deploying safeguards technology should be carried out in such a way that it supports the International Atomic Energy Agency’s (IAEA’s) international safeguards verification mission, while also being efficient in terms of both cost and time. This process, however, may be hindered by a general lack of nuclear facility operating experience among the scientists or engineers designing safeguards technology, or the inability to interact with the IAEA inspector end user. As a result, equipment designers may have difficulty understanding inspector needs and the limitations of using safeguards technology in the field, often leading to inefficiencies (in the form of increased cost and time spent) in the safeguards technology design process. Here, to mitigate this knowledge gap, the Y-12 National Security Complex in Oak Ridge, Tennessee, has developed design considerations that should be incorporated into safeguards technologies, focusing on non-destructive assay equipment. Timelines for integration of these considerations into the design process are also discussed.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Integrated Turbine Component Cooling Designs Facilitated by Additive Manufacturing and Optimization

This research program involved the development of improved gas turbine cooling by optimizing the combined internal and film cooling configurations used for turbine blades and vanes. Improved cooling of turbine airfoils allows the gas turbines to be operated at higher temperatures which improves efficiency for gas turbine-based power generation. This also facilitates operation with hydrogen-based fuels which inherently generate higher operating temperatures. Designs of the new cooling configurations were based on the added geometric flexibility that is available when using metal additive manufacturing techniques. When making these designs, adjustments were made to account for imperfections occurring in the metal additive manufacturing process. New designs for film cooling holes were developed using computational adjoint gradient optimization techniques which generated a unique film cooling hole with 70% greater film cooling effectiveness compared to conventional shaped film cooling holes. A major factor in improving the film cooling effectiveness were small external protrusions on either side of the new hole which generated vortices that pushed coolant towards the surface and increased the lateral spreading of the coolant. Our studies also included optimization of film cooling holes that can be built with conventional manufacturing techniques and optimizing the internal coolant channels that supply coolant to the film cooling holes. Various internal coolant channel shapes were investigated and designs that maximized the cooling of the airfoil while minimizing pressure losses in the channel flows were identified. Testing and verification of the performance of the final configurations was done with a transonic wind tunnel facility which allowed testing at engine realistic high velocities. Final testing of the combined internal and film cooling configurations was done by incorporating the configurations in Penn State’s National Experimental Turbine (NExT) vane. These tests confirmed the enhanced overall cooling effectiveness provided by the new cooling configuration designs.

03 NATURAL GAS↗
DOE OSTI2021

Integrable optics design principles for beam halo suppression in accelerator rings at the intensity frontier

RadiaSoft LLC is collaborating with Fermilab to fundamentally advance the field of particle accelerator design, in order to reduce cost and technical risk for next-generation hadron accelerators, and to more rapidly advance the intensity frontier of high energy physics. The Fermilab PI is one of the originators of the theory of integrable nonlinear magnetic lattices. Fermilab is the only laboratory in the world pursuing the experimental verification of the theory. Transferring Fermilab expertise to RadiaSoft LLC will enhance US competitiveness by enabling US industry to engage in the design of future high intensity hadron rings for spallation sources, neutrino sources, muon sources, subcritical nuclear reactors and waste transmutation facilities.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Implications of End Points on Remediation - 20476

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI compliantly, on time and within budget is CNL's number one priority. The PHAI is being implemented as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste. The Port Hope Project includes a unique and important component - the remediation of approximately 1,200 private residences and businesses to include both interior and exterior verification of soil and surface-contaminated objects. Planning and consultation for the PHAI clean-up criteria and remediation verification standard operating procedures (RVSOP) were initiated in the early 2000's. The criteria were developed through a cooperative effort involving the PHAI, scientific specialists, federal and provincial government agencies, peer reviewers, the municipalities and members of the public. Subsequent verification procedures which detail the unique systematic methods to verify compliance with the clean-up criteria were developed by CNL in consultation with federal regulators and relevant stakeholders. Now, specifically with the Port Hope Small-Scale Sites project having been underway for a year and approximately 38 properties in some stage of remediation, it has become increasingly evident that the practical application of the clean-up criteria and RVSOP methodology is posing unique challenges that are having direct impact on project execution. The current RVSOP approach is based on discrete sample verification not allowing failure of any one sample. When implemented in the field, particularly at the Small-Scale Sites, it leads to a much more extensive excavation than what was accounted for in the original design, the implications of which are substantial delays in project progress and much more significant impact on the individual properties than originally anticipated. This paper will examine the evolution of PHAI clean-up criteria, subsequent RVSOP methodology and its impact on current project execution and the justification for exploring potential revisions to methods of remediation verification. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗