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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 577 records · Page 32
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Inspection and Mapping of Savannah River Site (SRS) Waste Tanks via Unmanned Aircraft System (UAS) – 25351

The CSTF at SRS contain 51 waste tanks with 8 closed waste tanks between FTF and HTF. SRMC is the LW contractor. The LW mission includes removing legacy nuclear waste from these tanks and treating it for final disposition. Once the bulk of the waste has been removed from a tank, it will undergo inspection and sampling to characterize the remaining waste in the tank prior to it being operationally closed. There are multiple points in the tank closure process where an inspection is performed, and there are multiple parts of a tank that get inspected. Waste tanks have a primary containment vessel (referred to as the “Primary”) and a secondary containment vessel (referred to as the “Annulus”) that surrounds the primary. Both of these sections of a tank receive multiple inspections throughout the closure process.

Murphy, Lucas D. [Savannah River Mission Completio↗
DOE OSTI2024

Design and Engineering of LUPIN: A Test-Bed Radiofrequency Ion Source for Enhanced Neutral Beam Injection on DIII-D

The Large, Uniform Plasma for Ionizing Neutrals (LUPIN) is an RF inductively coupled plasma (ICP) chamber for demonstrating plasma performance of an RF ICP positive ion source upgrade for the DIII-D Neutral Beam Injection (NBI) system. LUPIN will be used to investigate ion source physics, including neutral gas dynamics, plasma density uniformity, interactions with Faraday shields, and power coupling to novel RF antenna designs. LUPIN has an RF generator capable of delivering 20 kW of power at 2 MHz, which is coupled into a cylindrical quartz vessel measuring 20 cm in length and 10 cm in radius. This configuration matches the power density requirements for a full-scale ion source. Target hydrogen and deuterium plasma densities exceeding 10 18 m -3 would relate to extracted ion current densities of 2100 A/m 2 for 10 s. Vacuum conductance and gas flow calculations predict a maximum achievable neutral gas flow rate of 1675 Pa·L/s at 5 Pa of He, which mimics the gas flow of the DIII-D NBI system. Designs have been developed for an internal Faraday shield to mitigate heat flux and ion sputtering on the dielectric vessel. Thermomechanical finite element simulations demonstrated the Faraday shield design to be capable of withstanding anticipated heat loads from worst-case operation scenarios. Results of upcoming experimental investigations on LUPIN will guide the design of a full-scale prototype for DIII-D integration.

Faraday shield↗
DOE OSTI2024

Fitness-for-Service Analysis of Reactor Components under Flexible Load-Following Operating Conditions

Conventional power-generation plants, including nuclear plants, have been traditionally designed to provide a steady baseload energy capacity, optimizing output efficiency while minimizing variable costs. However, the growing adoption of large-scale renewable energy-generation systems, which rely on intermittent sources such as solar and wind, has introduced more variability into the energy supply in interconnected electricity grids. As a result, the next generation of power plants needs to operate in what is known as the load-following mode, requiring flexible adjustments in electricity production to align with the energy demand on the grid. This transition from the steady baseload operation to load-following operating conditions can significantly increase the number of times various plant components are exposed to transient stresses. This increased thermo-mechanical cycling can lead to accelerated material degradation, thereby elevating the risk of premature failure of a component. It becomes imperative to conduct a comprehensive analysis of fatigue, creep-fatigue, and stress corrosion cracking life, to assess the resilience of the various engineering components under these flexible load-following operating conditions. This study aims to develop a comprehensive numerical model of a light-water reactor pressure vessel (RPV) to investigate its degradation under various operating scenarios. This coupled thermo-mechanical finite element analysis evaluated the stress response of the RPV caused by considering fluctuations in thermal and mechanical loads caused by the varying pressure and temperature occurring during the load-following operation. Critical locations on the RPV are subsequently identified based on the stress response. The stress intensity factors for the postulated flaws at those locations are then calculated, followed by an evaluation of the reactor's life in accordance with the ASME Boiler and Pressure Vessel Code Section XI. This comprehensive life assessment covers a number of transients expected during the flexible load-following operation, providing invaluable insights into the RPV's structural integrity. Moreover, the development methodology can be adapted to other reactor components, as well as components of conventional power stations that are affected by varying operating conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

COMPUTATION FLUID DYNAMICS ANALYSIS FOR GENERIC SMALL MODULAR REACTOR CONTAINMENT SEPARATE EFFECTS TEST

It is desirable for fourth-generation Small Modular Reactors to be passively cooled in standard and accident operations. Passive Containment Cooling Systems can reject heat from the containment structure, without using pumps or blowers. The targeted design containment structure is a large, domed, stainless steel, cylindrical vessel. In a postulated Design Basis Accident, steam will flash inside containment. Steam condensation occurs on the inner containment wall and transfers heat through the steel containment into a large body of water known as the annular reservoir (AR) surrounding the vessel serving as the ultimate heat sink. Natural circulation drives the flow in the AR and heat will be released to the environment by evaporation of water. Unique containment geometry requires a separate effects test (SET) facility for the verification and validation of the computer code and evaluation model development and assessment for reactor licensing efforts. In this study, STAR-CCM+, a computational fluid dynamics (CFD) code was used to inform the decision-making process on the design of the SET. The CFD simulation modeled, a two-phase turbulent flow with fluid film development and heat transfer for different containment geometries. The Reactor Excursion and Leak Analysis Program will also be used in a code-to-code verification against the CFD results.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2026

Modeling Marangoni Convection using a Velocity-Time Relationship

The Waste Treatment and Immobilization Plant (WTP) at the U.S. Department of Energy’s (DOE’s) Handford site is designed to treat 56 million gallons of radioactive waste through vitrification. The melting vessels physical and chemical integrity is essential to maintain, as failure has a high safety and financial penalty. Monofrax® K-3 refractory, the material widely used for the melting vessel, undergoes significant corrosion over time, concentrated at the triple-point junction between the gaseous atmosphere, refractory, and glass melt, producing a characteristic neck-like profile that significantly limits refractory service life. Accurate prediction of the corrosion within the melter can be used to prevent this and extend the lifetime of the melters. A computational fluid dynamics (CFD) model was previously developed, and predicted subsurface refractory corrosion, but failed to predict the neck-line profile observed experimentally. To address this limitation, the glass-melt meniscus geometry at the glass-refractory interface was estimated from experimentally measured neck profiles and wettability experiments. A curve was fitted to the neck profile to identify the triple point location, and the meniscus shape was characterized by a contact angle of 1°. A velocity profile accounting for Marangoni convection-driven corrosion was derived from the estimated meniscus geometry. Surface tension measurements from LORPM14R1 glasses of varying composition were coupled with K-3 coupon dissolution rates to establish a surface tension–time relationship. This analytical velocity-time model was integrated into the existing CFD framework to improve prediction of the neck corrosion profile. Work was done in a computational setting with national collaboration, resulting in professional development through presentational and technical writing growth.

36 - MATERIALS SCIENCE↗
DOE OSTI2023

ECAR-7547 Rev 0 MARVEL Intermediate Heat Exchanger Liner ASME Analysis

This ECAR analyzes the Secondary Coolant System liner, otherwise known as the Intermediate Heat Exchanger (IHX) Liner, as an ASME Section VIII, Division 1 pressure vessel. This analysis is only necessary in the event of a failure of the primary coolant boundary, in which case the IHX liner acts as a secondary confinement boundary to prevent primary coolant leakage. The liner main body or vessel is analyzed, as well as the liner flange ring.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2021

Toxicity of L19-Interleukin 2 Combined with Stereotactic Body Radiation Therapy: A Phase 1 Study

The immunocytokine L19-IL2 delivers interleukin-2 to the tumor by exploiting the selective L19-dependent binding of extradomain B of fibronectin on tumor blood vessels. In preclinical models, L19-IL2 has been shown to enhance the local and abscopal effects of radiation therapy. The clinical safety of L19-IL2 monotherapy has been established previously. In this study, the safety and tolerability of L19-IL2 after stereotactic body radiation therapy (SBRT) was assessed.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2022

Radiosensitizing the Vasculature of Primary Brainstem Gliomas Fails to Improve Tumor Response to Radiation Therapy

Diffuse intrinsic pontine gliomas (DIPGs) arise in the pons and are the leading cause of death from brain tumors in children. DIPGs are routinely treated with radiation therapy, which temporarily improves neurological symptoms but generally fails to achieve local control. Because numerous clinical trials have not improved survival from DIPG over standard radiation therapy alone, there is a pressing need to evaluate new therapeutic strategies for this devastating disease. Vascular damage caused by radiation therapy can increase the permeability of tumor blood vessels and promote tumor cell death.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

Fatigue analysis of the Spallation Neutron Source 2 MW target design

Upgrades to the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory are under way with two major projects. The first project is the Proton Power Upgrade (PPU), which will double the power capacity of the SNS accelerator system to enable the operation of a future second neutron source target station—the second project. PPU also will increase power to the SNS first target station – which currently operates at 1.4 MW with 1.0 GeV protons – to 2 MW with 1.3 GeV protons. Final design of the PPU 2 MW target module is finished, including all necessary design and analysis calculations. The fatigue evaluation described herein is an assessment of the fitness of the PPU 2 MW target mercury vessel to resist failure from cyclic loading from the pulsed 60 Hz proton beam and thermal changes from disrupted operation. In addition to the reliability predictions, the method, detail, and depth of this fatigue evaluation are superior to those of past target design assessments. Furthermore, the thermal fatigue life is predicted to exceed 3125 thermal cycles per 1250 h of operation for all load cases. The design also meets the (relative) fatigue design goal for combined thermal and pulse loading of greater than 0.5 times the minimum fatigue life calculated for a jet-flow target design operating at 1.4 MW—the most robust design to date in terms of resistance to fatigue failure. From the onset of the PPU, it has been known that pulse loading requires effective application of helium gas injection into the target mercury to reduce beam pressure loading. For the PPU 2 MW target, none of the pulse load cases requires a maximum strain reduction from gas injection of more than 50% (for base material) to meet the fatigue design goal. This level of reduction is achievable from gas injection, according to historical strain measurements from 1.4 MW target operation. Overall, the PPU 2 MW target design is predicted to have superior resistance to fatigue failure compared with past target designs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Design and structural assessment of the Spallation Neutron Source 2.0 MW target

The Proton Power Upgrade (PPU) project is underway at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory. The project will double the proton accelerator power from 1.4 MW to 2.8 MW, increase power to the First Target Station (FTS), and enable a future Second Target Station (STS). The power increase partly comes from raising proton energy from 1.0 GeV to 1.3 GeV. The STS will operate at 0.7 MW at 15 Hz when completed. Until then, PPU will provide the capability to operate the FTS at 2.0 MW at 60 Hz. Maximum power at the FTS to date has been 1.4 MW at 60 Hz with 1.0 GeV protons. A new mercury target module design to operate reliably under PPU conditions has been completed after a multi-year effort. The design philosophy and assessment of the structural analysis are described here. This target underwent an unprecedented design and analysis process for SNS using the latest engineering techniques and incorporating years of operating lessons and outcomes from R&D to meet structural design criteria. Finally, it also incorporates high-flow gas injection to further mitigate pulse fatigue stresses as well as cavitation damage to the mercury vessel.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

Endothelial cells control vascular smooth muscle cell cholesterol levels by regulating 24-dehydrocholesterol reductase expression

Highlights: • Suspended endothelial and vascular smooth muscle cells organize to form spherical vascular organoids in hanging drops. • Vascular smooth muscle cells downregulate the expression of metabolism regulating genes in presence of endothelial cells. • Endothelial cells decrease the levels of 24-dehydrocholesterol reductase and cholesterol in vascular smooth muscle cells. • Diminished cholesterol biosynthesis inhibits vascular smooth muscle cell activity. Communication of vascular cells is essential for the control of organotypic functions of blood vessels. In this context, vascular endothelial cells (EC) act as potent regulators of vascular smooth muscle cell (VSMC) functions such as contraction and relaxation. However, the impact of ECs on the gene expression pattern of VSMCs is largely unknown. Here, we investigated changes of the VSMC transcriptome by utilizing 3D human vascular organoids organized as a core of VSMCs enclosed by a monolayer of ECs. Microarray-based analyses indicated that interaction with ECs for 48 h down-regulates expression of genes in VSMCs controlling rate-limiting steps of the cholesterol biosynthesis such as HMGCR, HMGCS1, DHCR24 and DHCR7. Protein analyses revealed a decrease in the abundance of DHCR24 (24-dehydrocholesterol reductase) and lower cholesterol levels in VSMCs co-cultured with ECs. On the functional level, the blockade of the DHCR24 activity impaired adhesion, migration and proliferation of VSMCs. Collectively, these findings indicate that ECs have the capacity to instruct VSMCs to shut down the expression of DHCR24 thereby limiting their cholesterol biosynthesis, which may support their functional steady state.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2020

Area-Preserving Mapping of 3D Carotid Ultrasound Images Using Density-Equalizing Reference Map

Carotid atherosclerosis is a focal disease at the bifurcations of the carotid artery. To quantitatively monitor the local changes in the vessel-wall-plus-plaque thickness (VWT) and compare the VWT distributions for different patients or for the same patients at different ultrasound scanning sessions, a mapping technique is required to adjust for the geometric variability of different carotid artery models. In this work, we propose a novel method called density-equalizing reference map (DERM) for mapping 3D carotid surfaces to a standardized 2D carotid template, with an emphasis on preserving the local geometry of the carotid surface by minimizing the local area distortion. The initial map was generated by a previously described arc-length scaling (ALS) mapping method, which projects a 3D carotid surface onto a 2D non-convex L-shaped domain. A smooth and area-preserving flattened map was subsequently constructed by deforming the ALS map using the proposed algorithm that combines the density-equalizing map and the reference map techniques. This combination allows, for the first time, one-to-one mapping from a 3D surface to a standardized non-convex planar domain in an area-preserving manner. Evaluations using 20 carotid surface models show that the proposed method reduced the area distortion of the flattening maps by over 80% as compared to the ALS mapping method.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Thermal Aging Effects on the Yield and Tensile Strength of 9Cr-1Mo-V (Grade 91)

The long-term exposure of 9Cr-1Mo-V (Grade 91) steel to elevated temperature can have a significant effect on reducing its yield and tensile strength. The yield and tensile strength changes, in turn, have potential implications to the assurance of the integrity of components constructed with this material over their design or intended lifetime. The ASME Boiler & Pressure Vessel Code (BPVC), Section III, Division 5 (III-5, high temperature reactors) provides tabulated reduction factors for Grade 91 yield and tensile strength as a function of exposure temperature and time up to 300,000 h. ASME BPVC III-5 s intent to extend these factors to an exposure duration of 500,000 h, the lack of available historic information to support the existing factors, and the recent development of a physics-based prediction model for ASME BPVC application are prime motivation for this study. This work describes results of the conventional time-temperature Hollomon–Jaffe parameter, strength reduction ratio prediction method using an updated, extensive Grade 91 unaged and related aged material strength database. The method, previously used by Oak Ridge National Laboratory in its evaluation of Grade 91 and likely used in development of the existing BPVC III-5 reduction factors, provides strength reduction ratio predictions useful for general component fitness-for-service assessments and for computing BPVC III-5 reduction factors as defined. Specific reduction factors to 500,000 h at 650 °C applicable to ASME BPVC III-5 are computed from the strength reduction ratios.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Continuum Damage Mechanics Modeling of High-Temperature Flaw Propagation: Application to Creep Crack Growth In 316H Standardized Specimens and Nuclear Reactor Components

Predicting creep crack growth (CCG) of flaws found during operation in high-temperature alloy components is essential for assessing the remaining lifetime of those components. While defect assessment procedures are available for this purpose in design codes, these are limited in their range of applicability. This study assesses the application of a local damage-based finite-element methodology as a more general technique for the prediction of CCG at high temperatures on a variety of structural configurations. Numerical results for stainless steel 316H, which are validated against experimental data, show the promise of this approach. This integration of continuum damage mechanics (CDM) based methodologies, together with adequate inelastic models, into assessment procedures can therefore inform the characterization of CCG under complex operating conditions, while avoiding excessive conservatism. This article shows that such modeling frameworks can be calibrated to experimental data and used to demonstrate that the degree of tri-axiality ahead of a growing creep crack affects its rate of growth. The framework is also successfully employed in characterizing CCG in a realistic reactor pressure vessel geometry under an arbitrary loading condition. These results are particularly relevant to the nuclear power industry for defect assessment and inspections as part of codified practices of structural components with flaws in high-temperature reactors.

42 ENGINEERING↗
DOE OSTI2024

Design by analysis rules for ASME Section III, Division 5, Class B components

The current rules for elevated temperature ASME Section III, Division 5, Class B components, other than piping, were basically adapted from the Design-by-Rule approach for Section VIII, Division 1 vessels. A goal of the proposed new rules is to explicitly account for the design life and cyclic service while recognizing the less-rigorous requirements commensurate with lesser safety consideration. Further goals are to maximize the use of modern computational technology, for example finite element analysis in conjunction with reference stress concepts, and to avoid the need for stress classification. This paper summarizes the work done to address these issues. A summary of the design-by-analysis for primary loads, strain limit evaluation, and creep-fatigue damage assessment is presented. The proposed design-by-analysis creep-fatigue damage calculation approach uses a new elastic follow-up-based Isochronous Stress Strain Curve stress relaxation procedure. A set of sample problems are selected to validate the proposed design-by-analysis rules for creep-fatigue damage assessment. The proposed Class B rules are evaluated against the Class A elastic design rules and the experimental data obtained from a family of a simplified model test-based key-feature test results. The proposed Class B creep-fatigue damage assessment methodology yields conservative design cycles estimates compared to the experimental results.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Benchmarking of ENDF/B-VIII.0 cross sections for LWR ex-core transport applications

ENDF/B-VIII.0 cross section data is benchmarked on a simple Monte-Carlo model representative of the ex-core geometry of a PWR used for reactor vessel neutron fluence calculations. The benchmarking indicates that the ENDF/B-VIII.0 results show large differences as compared to previous evaluations, which have been extensively compared to measured data. This exercise suggests that ENDF/B-VIII.0 nuclear data is not suitable for ex-core LWR radiation transport applications. Evaluators are aware of the problems with the ENDF/B-VIII.0 evaluation of iron and work is underway to improve the evaluations for an upcoming release.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Utah FORGE: Fluid Injection Induced Shearing Experiments on Fractured Granitoid at Elevated Temperatures

This repository contains experimental data from a series of fluid injection-induced shearing tests conducted on Utah FORGE granitoid. The experiments were performed using an aluminum triaxial pressure vessel (TEMCO) apparatus at Pennsylvania State University. The primary aim was to investigate the impact of temperature on fault seismicity and to explore how different pre-stress ratios influence shearing behavior. The data includes results from experiments conducted under varying conditions of temperature, pre-stress ratios, and pore pressure increments. The rock samples used in these experiments were 60-grit granitoid with a single inclined fracture oriented at 60 degrees with respect to the horizontal cross-section. The dataset also includes measurements of normal stiffness at different temperatures and images of the experimental apparatus. The included README file details each experiential setup and outlines which data files represent which experimental conditions.

15 GEOTHERMAL ENERGY↗
DOE OSTI2020

ADCP Acoustic Interference Survey Data, Sequim Bay Inlet, WA

Raw ADCP (Acoustic Doppler Current Profiler) datasets from acoustic interference surveys with a TRDI Workhorse 300, a Nortek Signature500 and two Signature1000 instruments from August 2020. One Signature 1000 ADCP was deployed for 13 days on a bottom lander in Sequim Bay Inlet, WA. Data from the other three instruments were taken from a survey vessel running transects above the deployed lander.

16 TIDAL AND WAVE POWER↗