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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 1,315 records · Page 73
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Design and Optimization of a Gas-Cooled, Airfoil Fin Microchannel Heat Exchanger

High-performance microchannel heat exchangers are needed to supply heat for power conversion for nuclear microreactors. An airfoil fin microchannel design, constructed of Alloy 617 with helium as the working fluid, was analyzed and optimized using a design of experiments with artificial intelligence and machine learning techniques. The use of airfoil fins offers the potential to reduce pressure drop across the heat exchanger, as compared to other types of channel configurations. A framework for topology optimization of airfoil fin PCHEs has been developed that can be readily extended to different fin sizes and shapes, as well as different inlet and operating conditions, materials of construction, and working fluids. An optimization procedure was developed that employs computational fluid dynamics for a set of design points identified using Latin hypercube sampling. STAR-CCM+ was used to analyze a simplified two-channel configuration where five parameters were varied – inlet angle, fin scale, extent of staggering, transverse and longitudinal pitches. Two methods were compared for generating surrogate models – a 5D polynomial and a regression neural network. A response surface approximation was created from the surrogate models and input to a genetic algorithm. The genetic algorithm identified a set of optimal points on the Pareto front. The optimal geometry was found across six channel Reynolds numbers ranging from 1000 to 5000 to analyze how varying inlet conditions affects the optimal design. A set of optimal designs that maximizes heat transfer and minimizes pressure drop was identified, and a thermal stress analysis was performed on the optimal design. This work has developed a digital framework for the expedient topology design and evaluation of PCHE designs for gas-cooled microreactor applications. Correlations for the Nusselt number and Darcy friction factor were developed that can be useful for thermal hydraulic analyses using system codes. A thermal stress analysis was conducted and a brief discussion of the status of code cases of PCHEs for nuclear applications is given. Testing and thermomechanical modeling is needed to facilitate future code compliance of PCHEs for high pressure and high temperature applications.

42 ENGINEERING↗
DOE OSTI2021

2021 Source Term Template.xlsx

This is a screening spreadsheet and is used as a tool to calculate the source term for annual reporting, a new proposed project (APAD) or to calculate a periodic confirmatory measurement (PCM). NASA is doing a study of possible locations for a nuclear thermal propulsion ground test site and would like to use the spreadsheet to refine their study and provide guidance for future planning.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2021

Fuel Refabrication Prototype - Presentation for 2021 ASI Annual Webinar

This power point presentation is a summary of the work produced in FY2021 under the LWR Fuel Reinstrumentation project funded by DOE's Advanced Sensor Initiative, and the outlook for FY22. The objective of this portion of the ASI program is to "Capture critical technology created by the Institute For Energy (formerly Halden Reactor Project) to reinstrument irradiated fuel rodlets, and further this technology to enable incorporation of advanced instrumentation: fiberoptics, LVDTs, ultrasonic based sensors". The presentation explains that two of the three equipment modules were received and INL and a number of practice runs were conducted with them. In FY22 the third equipment module will arrive and using the three equipment modules measurement sciences personnel will create a complete prototypical instrumented fuel rodlet (using ceramic surrogates in the place of UO2).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Neutron Coincidence Measurements of Uranium-233 Oxide

Renewed international interest in thorium-fueled advanced reactors has challenged the safeguards community to address future proliferation concerns. Thorium-based technology presents many benefits but does not eliminate the proliferation risks associated with producing and processing fissile material. A byproduct of thorium-fueled reactors is uranium-233, which is classified as a direct-use material. As a result, the development of new or improved methods to characterize and measure materials containing 233U must mirror the pace of development of reactors and facilities that produce such material. Research is underway to assess, develop, and test approaches for safeguarding nuclear materials within the thorium fuel cycle. Neutron signatures from the nondestructive assay (NDA) of materials containing 233U are being quantified to inform the potential characterization of these materials. Using a traditional neutron coincidence counter and a series of well-documented 233U oxide samples, initial measurements have been made to assess the feasibility of 233U characterization and discrimination from other uranium isotopes, primarily 235U, using a combination of measurement techniques and analysis methods. Data acquisition is performed in list mode, allowing for a variety of analyses to be performed on the raw data that is not available using traditional shift register technology. Measurements were performed in passive and active configurations to quantify the strength of signal and to validate simulations in support of this work. This paper presents and discusses the results of the initial measurements of 233U oxide performed at Oak Ridge National Laboratory.

Lockhart, Madeline↗
DOE OSTI2024

Fusion Ignition and the Path to Inertial Fusion Energy

The achievement of fusion ignition at Lawrence Livermore National Laboratory's (LLNL) National Ignition Facility (NIF) in December 2022 was the culmination of more than 60 years of research and development in laser-driven inertial confinement fusion at LLNL. That historic scientific, engineering, and technological accomplishment, a prime example of the value of ingenuity and commitment in the face of a grand scientific challenge, marked a significant advance in LLNL's support of the National Nuclear Security Administration's science-based Stockpile Stewardship Program to maintain the reliability and security of the nation's nuclear deterrent without underground testing. It also furthered Livermore's research in high energy density science and established the fundamental scientific basis for inertial fusion energy (IFE), emboldening further public and private research into the development of IFE as a potential source of abundance clean, safe, and reliable energy. Finally, the U.S. government has funded a multi-disciplinary, multi-institutional program that LLNL is now leading to make inertial fusion energy a reality.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Post-irradiation 1600°C Heating Test of AGR-1 UCO Fuel Kernels

The Advanced Gas Reactor Fuel Development and Qualification (AGR) Program was established to perform research and development on tristructural isotropic (TRISO) coated-particle fuel to support deployment of high temperature gas-cooled reactors (HTGRs) . This talk will discuss the results from a post-irradiation heating test of five individual fuel kernels extracted from four TRISO fuel particles that were deconsolidated from an irradiated AGR-1 fuel compact.

heating test↗
DOE OSTI2025

NRIC Annual Report FY 2025

The National Reactor Innovation Center (NRIC), established in August 2019, is a national United States (U.S.) Department of Energy (DOE) program. NRIC’s mission is to partner with industry and national laboratories to bridge the gap between concept, demonstration, and commercialization of advanced nuclear technology. NRIC accomplishes this through building or enhancing existing DOE infrastructure to support testing of components and systems that are key to successfully deploying advanced nuclear technology. NRIC’s vision is that by 2028, NRIC will be partnered with industry and accelerating the demonstration and deployment of advanced nuclear technology using DOE national laboratory infrastructure and expertise. NRIC will establish four new experimental facilities and two large reactor test beds for integrated technology demonstrations and experimentation by 2028 and complete two advanced nuclear technology tests by 2030. Achieving this vision will enable urgently needed abundant and affordable clean energy both domestically and internationally. NRIC’s success will inspire our nation and the global community to embrace the promising contribution of innovative nuclear reactor technologies to the clean energy economy and re-establish the U.S. as the global leader in advanced nuclear energy. NRIC is tasked with expediting the development of advanced nuclear energy technologies by bringing together private-sector technology developers and the world-class capabilities of the DOE national laboratory system. Through this program, the U.S. private sector is given access to the physical infrastructure available at DOE national laboratories to test and demonstrate their reactor concepts. NRIC works closely with the DOE-Nuclear Energy (NE) program that grants access to technical, regulatory, and financial support for commercializing nuclear energy. NRIC builds upon these new reactor concepts and technology successes to effectively strengthen U.S. nuclear leadership.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Conceptual Design of Engineering-Scale Molten Salt Spill Tests to Support Molten Salt Reactor Licensing

Nuclear accident progression analysis codes for reactor licensing must be validated using experimental data that was generated at a large enough scale so that the results are representative of full-scale accidents. This report describes a conceptual design of engineering-scale molten salt spill tests that will provide the necessary datasets to support molten salt reactor (MSR) licensing. Design considerations, salt compositions, test variables, processes to be quantified, and suitable measurement techniques are presented and discussed. Aspects of the conceptual engineering-scale molten salt spill test design presented in this report were made using insights from recent laboratory-scale molten salt spill tests conducted at Argonne. Additional input from model developers, MSR vendors, and technical experts will benefit the design of the proposed engineering-scale molten salt spill tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

An Architectural Survey of the Area 25 Nuclear Rocket Development Station, Nevada National Security Site, Nye County, Nevada, Volume 3 Of 3

The purpose of the current architectural survey is to record the Nuclear Rocket Development Station (NRDS) district, which is recommended eligible for listing to the NRHP under all four criteria as the testing center for the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) program from the initial boundary survey in 1956 to the end of nuclear rocket development on January 5, 1973. Resources in the NRDS district were evaluated to determine if they are eligible for listing to the National Register as district contributors. The current survey identified the Reactor Control Point subdistrict to provide a temporal, spatial, and functional framework for organizing the resources involved in the NRDS operations. However, the subdistrict was not evaluated as a potential historic district against the Secretary of Interior’s Significance Criteria.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

An Architectural Survey of the Area 25 Nuclear Rocket Development Station, Nevada National Security Site, Nye County, Nevada, Volume 2 Of 3

The purpose of the current architectural survey is to record the Nuclear Rocket Development Station (NRDS) district, which is recommended eligible for listing to the NRHP under all four criteria as the testing center for the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) program from the initial boundary survey in 1956 to the end of nuclear rocket development on January 5, 1973. Resources within the boundary of the NRDS district were evaluated to determine if they are eligible for listing to the National Register as district contributors. Subdistricts within the larger NRDS district were defined to provide a temporal, spatial, and functional framework for organizing the resources involved in the NRDS operations. However, developing contexts for the individual subdistricts was beyond the scope of the NRDS survey. Therefore, the potentially eligible subdistricts were not individually evaluated under any of the Secretary of Interior’s Significance Criteria for eligibility to the National Register.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS2010

Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

Bragg-Sitton, Shannon M.↗
DOE OSTI2022

Safeguards by Design Projects (Final Report-FY-22)

This University Engagement project challenged engineering students at universities, that do not have Bachelor degree programs in nuclear engineering but do have research reactors and some nuclear engineering coursework, to incorporate Safeguards by Design concepts into their Senior Capstone Design Project. This University Engagement project was part of the U. S. Department of Energy’s (DOE) National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation, Office of International Nuclear Safeguards, Next Generation Safeguards Initiative (NGSI), Human Capital Development (HCD): University Engagement Program. This program exposed university students with Mechanical Engineering majors and Nuclear Engineering minors to the concepts of international nuclear safeguards. In FY22, three teams at the University of Rhode Island and two teams at the University of Texas - Austin participated in researching, designing, building, and testing projects to support international nuclear safeguards measurements or verification. The projects involved engaging in activities at the university’s research reactors. All the projects engaged students with prototyping a design and/or tool for application at the Universities’ reactor. At the end of the course, most of the students expressed the experience was positive and they learned more about international nuclear safeguards and applying requirements than they had previously encountered. This school year the projects were further complicated by the COVID-19 pandemic. Both universities had limited in classes on campus, still relying on Zoom classes, and limited direct student/professor interactions. Furthermore, Los Alamos National Laboratory (LANL) greatly restricted travel, therefore making it impossible to visit the students at the end of the semester for the review of their design projects. The final design and review meeting for the projects happened via meetings over the internet. Additionally, the teams did build and test some prototypes but could only do so in a limited capacity.

42 ENGINEERING↗
DOE OSTI2025

Evaluation of an Accident Tolerant Fuel Leak in the Advanced Test Reactor

Accident Tolerant Fuels (ATF), which are nuclear fuel sources designed to withstand operational irregularities and incidents, have been a topic of interest in the nuclear industry for several decades. Interest in ATF technology surged following the 2011 accident at Fukushima Daiichi in Japan. At the Advanced Test Reactor (ATR), one of Idaho National Laboratory’s (INL) four operating nuclear reactors, the ATF program is a collaborative effort between the national laboratory and various stakeholders within the nuclear industry. This program focuses on the research and development of novel fuel compositions, cladding, and component materials with enhanced accident-resistant properties. During one of ATR’s 60-day operating cycles in 2024, the reactor experienced five unplanned shutdowns. Following the fifth shutdown, radiation monitors detected an increase in radiation levels coming from the loop piping. Subsequent water samples confirmed the cause was a leak of fission products from the ATF experiment, designated as ATF-2C. The source of the leak was identified as the instrumented section of the test train. The primary discussions in this paper are 1) the design of the ATF test train, 2) the operating parameters leading up to and following the detection of the leak, and 3) the quantification and characterization of the released fission products.

Accident Tolerant Fuels↗
DOE OSTI2019

Fracture Toughness Evaluation for Sandia Mock-up Stainless Steel Canister Weldment Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials that have no current outlet than storage. These materials are often held in long-term storage as an interim stage within their lifecycle. Lifecycles can include reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations are very important, due to a great concentration of fission products, actinides and activation products. Stress corrosion cracking of interim storage containers has been identified as a high priority data gap by the Department of Energy. However, little has been done with regards to the canister weld material properties, including the fracture toughness, and their impact on stress corrosion, until recently from Sandia National Laboratory canister mockup program. Furthermore, because no post-weld heat treatment was required, the associated high tensile residual stress within these canister welds can drive the initiation and growth of stress corrosion cracking (SCC) cracks. Thus, the fracture mechanical property, such as fracture toughness, is an essential information to understand the canister weldment fracture resistance behavior and to support mitigation protocol development to counter SCC. In this work, we carried out fracture toughness evaluation on the received mock-up canister weldments from Sandia, using spiral notch torsion fracture test technology. The details of the weldment fracture toughness evaluation program development at ORNL are presented in this progress report. The objective of this research is to collect fracture toughness data on the mock-up canister weldment from Sandia National Laboratory. Data will be collected under quasi-static fracture loading using the Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support the ongoing SNF dry storage canister modeling activities and to address regulatory issues associated with canister container integrity and the follow-on SNF system transport reliability. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre-cracks are needed. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient crack length to increase the sample fracture tip localized constraint. Fatigue pre-crack growth techniques are employed to introduce sharp crack front into the sample. Previously, only rough calculations were applied to estimate the compliance evolution in the SNTT crack growth process; while accurate quantitative descriptions have never been attempted. This generates an urgent need to understand the crack evolution during the SNTT fracture testing process of ductile materials. The newly developed governing equations for SNTT crack growth estimate are used to control the fatigue crack growth to effectively reaching the target crack length. The detailed SNTT approach and its estimated fracture toughness for the as-received Sandia canister weldment and the associated baseline 304 stainless steel are presented in this report. In the past SNTT technology has also been successfully applied to investigate the fracture behavior of X52 and X80 steels and the X52 welded materials used for hydrogen infrastructures. The current study in supporting Sandia SNF Dry Storage Mock-up Canister Program is the first attempt of applying SNTT to the highly ductile stainless steel and the associated weld materials. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to SS304 steel and SS304/308 weld material. The estimate JQ upon fracture for the baseline SS304 steel is at 1618 lb/in. The estimate JQ upon fracture for the SS304/308 weld from HAZ and weld specimens are 720 lb/in and 850 lb/in, respectively.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Experiment design for the neutron irradiation of $\mathrm{PM-HIP}$ alloys for nuclear reactors

Here, this article describes the design of an Advanced Test Reactor (ATR) drop-in neutron irradiation experiment aiming to directly compare the performance of nuclear structural alloys fabricated by powder metallurgy with hot isostatic pressing (PM-HIP) against conventional casting or forging. There is growing interest in PM-HIP alloys for nuclear applications because of their microstructural uniformity, superior mechanical properties, and reduced dependence on welding and machining, compared to cast/forged alloys. Nuclear code-qualification of PM-HIP alloys requires neutron irradiation testing to demonstrate performance under relevant conditions. In this experiment, six nuclear structural alloys were irradiated: Ni-based alloys 625 and 690, Grade 91 ferritic steel, SA508 pressure vessel steel, and 304L and 316L austenitic stainless steels. The experiment is assembled into seven capsules in four test trains and irradiated in three ATR inboard A positions. Both the PM-HIP and cast/forged versions of each alloy were irradiated under nearly identical conditions for comparative purposes, to target doses of 1 ± 0.2 and 3 ± 0.2 dpa at temperatures of 300 ± 50 °C and 400 ± 50 °C. A thorough description of the experiment design and thermal, structural and neutronic analyses performed to ensure the targeted irradiation conditions are met is provided. Specimens were configured as small disks, compact tension specimens and tensile bars to facilitate post-irradiation examination (PIE) that will include mechanical testing, microstructure characterization, and fracture toughness testing. Given the considerations for ASTM standardized mechanical testing, comparative fluence and temperature across specimen pairs, and comprehensive PIE planning herein, this work serves as a template for future nuclear materials qualification experiment designs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗