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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Qualification of Digitized Legacy Fast Reactor Data

The Integral Fast Reactor (IFR) fuel compatibility test program (1984-1994) included a variety of fuel pin examinations conducted at the Hot Fuel Examination Facility (HFEF) and the Alpha-Gamma Hot Cell Facility (AGHCF). Hard copy data records of these examinations have been recovered, scanned, and preserved in PDF format. Many hard copy records are now qualified in accordance with an NRC-approved Quality Assurance Program Plan (QAPP), and there is an ongoing effort to qualify additional legacy records. This legacy fuel performance data is vital to support design and licensing of fast reactors with validation of state-of-the-art codes and advanced methods for design and analysis. Stakeholders can most easily utilize this data when the PDF scans have been converted into digital data tables. However, qualification of the scanned hard copy data does not qualify the digital data file resulting from the digitization of the data contained in the record; the subject matter expert (SME) must make a review of the digitized data table as well before it can be designated as qualified. This report outlines a peer review process to qualify the digital data file(s), typically in CSV format, corresponding to hard copy records in accordance with the existing QAPP.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Remembering Robert S. Fitzhugh, 1932-2007

Robert S. Fitzhugh, a Laboratory pioneer and mainstay of the Laboratory’s nuclear testing program, died January 7, 2007, just two days after celebrating his 85 th birthday. An engineer dedicated to craft, Fitz was one of the longest serving Laboratory employees and one of the most respected. His pension, because of his long tenure, was higher than his salary. Born January 5, 1922, in Philadelphia, Fitz graduated high school in 1939 and from Michigan State University with a BS in Electrical Engineering in June 1943. He enlisted in the United States Army in May 1943 and, after completing basic training, attended the University of Iowa as part of the Army Specialized Training Program. When the ASTP program was disbanded in early 1944, Fitz was sent to Columbia University as a laboratory technician and then on to Oak Ridge, where he worked on the thermal diffusion program. Fitz did not like Oak Ridge, describing the Zeppelin-like hanger he worked in as “a horrible place.”

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2025

Test Report for the Regulatory Testing of the Welded Isotek-SFC-W-3 Special Form Capsule

Two prototypes the welded ISOTEK-SFC-W-3 special form capsules (SFCs) of the same design were evaluated to determine if the requirements of Title 49, Code of Federal Regulations (CFR), Part 173.469, Tests for Special Form Class 7 (Radioactive) Materials, were met. The results of the special form tests are documented in this test report. This report describes the special form testing activities performed on the two welded ISOTEK-SFC-W-3 capsules. The test units were designated as TU-1 and TU-2. TU-1 was subjected to leak testing, percussion testing, and impact testing. TU-1 was leak tested using the leak rate test specified in 49 CFR 173.469(a)(4)(i). The impact test on TU-1 was conducted according to the requirements in 49 CFR 173.469 (b)(1). The percussion test that was carried out on TU-1 was conducted in accordance with 49 CFR 173.469 (b)(2). TU-2 was subjected to a leak rate test as specified in 173.469(a)(4)(i) and a heat test as specified in 49 CFR 173.469(b)(4). Each test unit was leak tested before and after these respective tests. The leak rate tests performed were helium back-pressure tests and bubble tests, as specified in American National Standards Institute (ANSI) N14.5-2014. The measured leak rates were converted to standard condition leak rates as specified in American Society for Testing and Measurement (ASTM) E 493. The determined standardized leak rates obtained from the test and the calculations for all the test units met the requirements for special form certification. The testing was performed under the direction of the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Storage Field Development Plan: One Earth Energy

This Storage Field Development plan presents the Storage Complex characterization results, construction, monitoring, and operational plans, and costs associated with the proposed One Earth Sequestration Carbon Capture and Storage (OES-CCS) site in McLean County, Illinois, near Gibson City. The proposed storage complex, known as the Mt. Simon Storage Complex, comprises the Cambrian Mt. Simon Sandstone reservoir and the primary seal, the Cambrian Eau Claire Formation. The lowermost Underground Source of Drinking Water (USDW) identified for the site is the Ordovician St. Peter Sandstone. Geologic characterization of the Mt. Simon Storage Complex at the OES-CCS site was performed by the Illinois Storage Corridor CarbonSAFE Phase III project, which also prepared and submitted three UIC Class VI applications to construct three injection wells; the permit applications were submitted and are in the federal EPA review process. A characterization well, OEE #1, was drilled to collect site-specific data. These data were analyzed and used to develop the UIC Class VI applications. The OEE #1 well will be converted to an in-zone monitoring (IZM) well for the injection phase. The proposed buildout for the OES-CCS site includes (1) three injection wells (OES #1, OES #2, and OES #3), (2) two IZM wells, (3) two above confining zone (ACZ) monitoring wells, one of which will be used to monitor the lowermost USDW, (4) capture and compression facilities, and (5) transportation facilities, i.e., pipelines. A pre-operational testing program was proposed in the Class VI permit application and will be employed at the site pending approval. Additional pre-injection (baseline), syn-injection, and post-injection monitoring and site care procedures will be followed by OES to ensure that injection activities are protective of human health and the environment. Injection is scheduled to begin in 2025, distributed across the three injection wells in accordance with the permit operating conditions. One Earth Sequestration intends to inject up to 90 million tonnes of CO 2 over a period of approximately 20 years. Injection will begin at approximately 0.5 million tonnes of CO 2 annually and ramp up to a maximum of 4.5 million tonnes annually. Daily injection rates are expected to range from 1,400 to 1,500 tonnes per day initially and reach a maximum of approximately 4,225 tonnes per day, depending on site geology and injectivity at each injection well location, and CO 2 availability. The costs associated with the OES-CCS project include pre-operational costs (e. g. additional seismic data acquisition and well drilling), capture and transportation facility and equipment costs, predicted field operating expenditures (OpEx), and decommissioning and post-injection site care (PISC) costs. The risks associated with project activities, such as site construction, injection operations, and verification of secure storage were evaluated, and mitigation strategies proposed to alleviate those risks.

09 BIOMASS FUELS↗
DOE OSTI2026

Laboratory Tests of Laser Control of Electron Beams for Future Colliders

Laser-driven Compton backscattering (CBS) has been proposed as method for controlling the intensity of colliding bunches in the FCC-ee so as to avoid the flip-flop instability caused by intensity asymmetry in colliding bunches. Laser-based collimation has also been proposed as an indestructible collimator for high-intensity electron beams. We have initiated a laboratory-based test program of these concepts with the E344 experiment at FACET-II. In this paper, we describe simulations of laser-beam interactions at FACET-II and the relevant scaling for FCC-ee. We also describe the experimental setup and diagnostics that will be used to make the measurements at FACET-II.

Accelerator Physics (physics.acc-ph)↗
DOE OSTI2021

Global Power System Transformation Consortium

The Global Power System Transformation (G-PST) Consortium, an expert- and practitioner-driven initiative, engages key power system operators, research and educational institutes, governments, businesses, and stakeholders in all regions of the world to accelerate clean energy transitions at the ambitious scope and scale that is required. The G-PST Consortium provides coordinated, holistic support and knowledge infusion for system operators pursuing clean energy transitions, including performing cutting-edge research; providing implementation support for world-class engineering and operational solutions; supporting workforce development; building and disseminating open-access data and tools; and accelerating localized technology adoption, standards development, and testing programs.

consortium↗
DOE OSTI2022

A709 Qualification Plan Update and Mechanical Properties Data Assessment

This report provides a summary of the development effort for the qualification of Alloy 709, an advanced austenitic stainless steel, in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. It provides an assessment of the mechanical properties data generated to date from the tensile, creep, fatigue and creep-fatigue tests of the precipitation-treated Alloy 709 from two commercial heats in plate product form. It was concluded that the mechanical properties of Alloy 709 with the precipitation treatment continued to outperform those of Type 316 stainless steel. This affirms the recommendation to Code qualification this alloy as a replacement for Type 316 stainless steel to support the objective of reducing the construction and operating costs to incentivize advanced reactor deployment. The data also demonstrated that the precipitation treatment is effective in enhancing the creep-fatigue resistance of Alloy 709 while maintaining a significant creep strength advantage over Type 316 stainless steel. This report also provides an update to the test conditions for the creep, fatigue and creep-fatigue test matrices in order to cover the Code Case data package more effectively. Finally, it is recommended to continue the Alloy 709 Code Case Testing Program to develop the data package needed for the determination of the material-specific design parameters for inclusion in the Alloy 709 Code Case.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Drawings

These are drawings that display dimensions and orientation of specimens for the graphite baseline program test specimens.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2020

Water Capsule RIA Testing

Presentation on the status of the water-based RIA testing program in TREAT for the 2020 AFC Annual Review Meeting.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

High Temperature Ternary Chloride Molten Salt Pump

This work investigates the design criteria considering thermal, structural, and material selection of a ternary chloride (MgCl2-NaCl-KCl) molten salt pump that is capable of running at temperatures above 720 °C. In this work, preliminary testing of salt-wetted hydrodynamic bearings was done to investigate different material pairings including Colmonoy grades, Stellite grades, and a novel Nickel Tungsten Carbide alloy (HybrimetTM NiWC3b) used as a High Velocity Oxygen Fuel (HVOF) coating. The coating has shown to withstand the corrosive nature of the molten salt. The pump wetted components are made of Inconel 625 and features a carbon ring seal (CIRPAC Seal) flooded with a N2 gas flow on the overhead to keep salt vapors from interacting with the external ambient air. A comprehensive pump test program has been completed. Over 90 hours run time pumping hot chloride salts at temperatures up to 720 °C in which enough data was captured to produce performance curves at 3 different operating speeds. This research showcases a promising design for use in the Concentrated Solar Power (CSP) Gen 3 liquid-pathway and sets the ground for further development and use of the HybrimetTM NiWC3b as a competitive alternative to available superalloys.

14 SOLAR ENERGY↗
DOE OSTI2020

Application of FREMES to Characterize and Sort Soil During Site Remediation - 20182

During decommissioning of buildings and the related release of nuclear sites, large amounts of bulk materials such as soil must be measured for clearance. These measurements are required to demonstrate that for example potentially contaminated soil is below the respective radiological limits. Doing so also helps to minimize the amount of bulk material which has to be disposed of as nuclear waste. NUKEM Technologies Engineering Services GmbH (NUKEM) applies their FREMES technology in a project at the former fuel element factory of FBFC International at Dessel, Belgium. The purpose is to measure and check for potential contamination within the excavated soil. The project started in 2017 with design and procurement of the necessary equipment. In October 2017 the FREMES system was installed at the site. After installation and an extensive test program the FREMES facility went fully into operation in July 2018. Original foreseen to measure between 4.000 and 8.000 tons of material, it has currently (mid 2019) measured 23.600 tons. It is expected that another 5.000 tons still have to be measured before the project successfully ends. (authors)

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

GCR: Development of Improved Alloy 800H Weldment

In Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Alloy 800H is qualified for elevated-temperature nuclear construction for temperatures up to 760°C (1400°F) and a maximum service life of 300,000 hours. There are two permissible filler metals for Alloy 800H weldments specified in Division 5: ENiCrFe-2 (Alloy A) and ERNiCr-3 (Alloy 82). Low creep-rupture strengths of these weldments at the upper limits of the qualified temperatures and service lives may restrict the design envelope for elevated-temperature nuclear construction with Alloy 800H. As a result, an alternative filler metal is desired to improve the creep-rupture strengths of Alloy 800H weldments for the qualified temperatures and service lives. This work investigates an overmatched filler metal. Specifically, a weldment with Alloy 800H base metal and Alloy 617 filler metal fabricated by semiautomated gas tungsten arc welding is investigated. A scoping creep-rupture test program was conducted of cross-weld specimens at temperatures ranging from 750 to 1000°C (1292 to 1832°F). Preliminary results on the creep-rupture strengths of the Alloy 800H weldment with an Alloy 617 filler metal do not show significant improvement compared to the filler metals currently qualified in Division 5 for Alloy 800H weldments. Consequently, work is in progress to investigate a matching filler metal, UTP A 2133 Mn.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Poster for 12th World Conference for Neutron Radiography (WCNR-12)

Complimentary Relationships Between Neutron Computed Radiography and Digital X-ray Radiography of Radioactive Specimens The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory which is used for transient testing of materials and experiments. A neutron radiography facility is installed which allows for low resolution neutron imaging with a capability of items up to 20 cm wide and 400 cm long. Along with the neutron radiography facility, a 370 kV digital X-ray system is located onsite for when X-ray imaging is more practical or when rapid imaging is requested. The X-ray facility can accommodate specimens up to 61 cm wide and 1000 cm long. This poster discusses the complimentary relationship between transfer-method Neutron Computed Radiography (nCR) and Digital X-ray Radiography (DR) imaging of radioactive specimens along with practical applications for nuclear fuels and materials testing programs. Both nCR and DR images were acquired of inert gas and water filled apparatus containing differing materials at varying radiation levels. Highly radioactive items have no effect on nCR image quality whereas radioactivity has varying degrees of effect in DR. Radioactive items as high as 20 mSv/h have been successfully imaged using X-ray DR methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MOOSE Optimization Module Overview: Application to Residual Stress Inversion in Nuclear Fuel Plates

Manufacturing methods, material selections and unique plate geometries used in low enriched uranium plate fuels for the High Performance Research Reactor (HPRR) have led to concerns about the residual stresses effect on bond strength of the fuel/cladding interfaces and overall fuel performance under irradiation. The current HPRR plate fuel test program performed residual stress calculations using the contour method with experimental displacement data obtained by incrementally cutting the plate fuel. In this work, we use a gradient based inverse optimization module developed within the MOOSE framework to estimate normal and shear residual stresses within the plate fuels using displacement data taken from a cut placed in the HPRR fuel plate. These results are verified against residual stress calculations obtained from the contour method. The inversion protocol used in this work utilizes two meshes to separately discretize the physics and parameter space. This allows a fine mesh to be used to fully resolve the physics and geometry of the forward problem. A coarser mesh is used to resolve the parameter space with spatially varying mesh density determined from a sensitivity analysis. The adaptive resolution of the parameter space essentially has the effect of regularization as it is designed to achieve the bias-variance tradeoff.

42 ENGINEERING↗
DOE OSTI2022

The Ground Truth Program: Simulations as Test Beds for Social Science Research Methods.

Social systems are uniquely complex and difficult to study, but understanding them is vital to solving the world’s problems. The Ground Truth program developed a new way of testing the research methods that attempt to understand and leverage the Human Domain and its associated complexities. The program developed simulations of social systems as virtual world test beds. Not only were these simulations able to produce data on future states of the system under various circumstances and scenarios, but their causal ground truth was also explicitly known. Research teams studied these virtual worlds, facilitating deep validation of causal inference, prediction, and prescription methods. The Ground Truth program model provides a way to test and validate research methods to an extent previously impossible, and to study the intricacies and interactions of different components of research.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Pellet Cladding Interaction In-Reactor Ramp Testing in a World without the Halden Boiling Water Reactor

One of the most crucial performance areas for fuel rods in water cooled nuclear power plants is interaction between cladding tubes and fuel pellets. Experimental programs in test reactors have provided key data to help fuel developers and plant operators understand Pellet Cladding Interaction (PCI) phenomena and optimize their strategies for reliable fuel performance. Approximately 50 years of “ramp” testing programs, where the fission heating rate is deliberately manipulated in test rods, have been performed in a handful of test reactors to reveal and understand PCI behaviors such as iodine-assisted stress corrosion cracking. Unfortunately, the test reactors most engaged in this type of work have all been retired over the years up to the recent and unexpected closure of the Halden Boiling Water Reactor which effectively caused a hiatus in PCI ramp testing programs. The need for ramp testing is crucial at this time to enable refined PCI understanding as more plants consider implementing flexible operations, increased fuel rod burnup limits, and new fuel technologies with enhanced accident tolerance. This paper reviews some of the key PCI phenomena that must be addressed with in-pile testing and surveys past test reactor’s methods for achieving the needed conditions. A strategic approach is then presented using test reactors and complimentary facilities which are still available today. Near term data opportunities are put forth along with capability development strategies that will ensure future longevity in this field of research.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Benchmark Modeling and Simulation of the FFTF LOFWOS Test #13 Using SAM

The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor, built to assist development and testing of advanced fuels and materials for fast breeder reactors. In July 1986, a series of unprotected Loss of Flow Without Scram (LOFWOS) transients were performed in FFTF as part of the Passive Safety Testing (PST) program. The LOFWOS Test #13, which was initiated at 50% power and 100% flow with the pump pony motors left off, has been chosen as a benchmark case by IAEA to support collaborative efforts within international partnerships on the validation of simulation tools and models in the area of sodium fast reactor passive safety in an IAEA Coordinated Research Project (CRP), launched in October 2018. The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR safety analysis. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. To participate the IAEA CRP and enhance the SAM validation base for advanced reactor transient safety analysis, benchmark simulations of the FFTF LOFWOS Test #13 are performed using the SAM code. In this first phase of the validation effort, the thermal-hydraulic behavior of the reactor system is the focus and the reactor kinetics is not considered in the SAM FFTF model. Instead, the results of Argonne’s neutronics calculations are directly used, including the power shape of the active core region and the power history during the transient. The simulation results of FFTF at steady state agreed well with the measured data from the test. During the transient, reasonably good agreement were also obtained. Future work to improve the model will focus on introducing the reactivity predictions into the model, as well as better understanding or resolving the current discrepancies with the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗