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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Evaluation of thermal energy storage for central receiver concentrating solar power plants under charging cycles through structural assessment

Electricity production by concentrated solar power (CSP) systems has stood out among energy transition alternatives due to their large generation capacity (between 30 MW and 400 MW) and high capacity factor (80%) when incorporating thermal storage systems (TES). Although TES tanks in operation worldwide are made of austenitic steels with remarkable mechanical and anticorrosive properties, operational failures have been reported due to low-cycle fatigue, creep, and the abrupt release of residual stresses generated during tank manufacturing. A critical operation for salt tanks is the initial daily charging operation, given the low fluid level and the thermal gradients between the inventory and the salt entering the tank. This study presents a structural simulation of a 39.6 m-diameter molten salt tank for central receiver CSP plants, accounting for temperature and pressure variations during charging and different sparger ring inlet configurations. The initial pre-stressed condition of the floor, resulting from the manufacture of welded plates, is used as a boundary condition. The fatigue and creep life analyses were performed in accordance with the ASME BPVC and API standards. It was found that the tank's floor is in a less favourable mechanical condition than the wall, due to its initial manufacturing condition and the maximum temperature differences during charging. A critical zone was identified in an area affected by residual stress on the floor and under the sparger ring. The baseline sparger ring configuration, with 52 2-inch-diameter orifices and a flow direction of 90 degrees , results in an average floor stress of approximately 25 MPa and creep damage after 1.2 years of tank operation.

14 SOLAR ENERGY↗
DOE OSTI2019

Evaluating and modeling rate sensitivity in advanced reactor structural materials: 316H, Gr. 91, and A617

Advanced reactor systems operate at high temperatures to achieve better thermal efficiency. During transients from lower temperatures to high operating temperatures structural metals transition from rate insensitive behavior to rate dependent behavior. In the rate dependent regime creep and plastic deformation are not separable hence requiring a unified treatment with a viscoplastic model. Several bounding design methods in the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B assume a decoupled treatment of creep and plasticity so establishing the boundary of the rate insensitive regimes for advanced reactor structural materials is necessary to set limits for the use of these bounding design methods. In the rate sensitive regime new bounding methods or fully inelastic analysis are required for the design of Division 5 Class A components. During temperature transients these components may operate in both the rate insensitive and rate sensitive regions, requiring computational models that can represent both types of behavior. This work establishes criteria for material rate sensitivity and develops a new composite method for modeling constitutive response in both the rate sensitive and rate insensitive regimes for three Division 5, Class A materials: 316H austenitic stainless steel, Grade 91 ferritic-martensitic steel, and Alloy 617, a nickel-based alloy. Both the criteria and new modeling strategy apply the concept of a normalized activation energy for thermally activated mechanisms developed by Kocks, Mecking, and co-workers. Specifically, this work quantifies the average rate sensitivity of the three materials from an experimental database collected from the literature. Additionally, this work develops rate sensitivity criteria and a complete model for each material that can transition between both deformation regimes and provides several example simulations demonstrating the accuracy of the models and the numerical performance of the composite modeling method.

36 MATERIALS SCIENCE↗
DOE OSTI2021

A method for predicting failure statistics for steady state elevated temperature structural components

This paper presents the initial development of a high temperature life prediction method that accounts for the variability in the material properties of Grade 91 steel. The method accounts for material variability by fitting a variable 3-parameter Weibull distribution to experimental rupture data and accounts for the variability of creep deformation on the steady-state stresses via a Monte Carlo approach. To ensure reasonable computational times, the model represents the material as an extremely viscous Stokes fluid with a non-Newtonian viscosity, therefore solving the stress relaxation problem with a steady, static, instead of transient, analysis. Furthermore, the complete statistical analysis combines this model for creep deformation with a probabilistic model for creep rupture to evaluate the probability of premature failure for a set of sample problems, comparing the predicted failure statistics to the design life predicted by the ASME Boiler and Pressure Vessel Code rules.

42 ENGINEERING↗
DOE OSTI2022

Development and performance of INCONEL® alloy 740H® seam-welded piping

INCONEL® alloy 740H® is an age-hardenable nickel-based alloy approved for pressure vessels and piping within ASME Section I, VIII, and B31.1. Currently, the code applies a weld strength reduction factor (WSRF) of 0.7 to the allowable stresses for longitudinal seam welds in the time dependent creep regime. In this work, a full-scale seam weld was successfully produced using typical industrial practices. The component was solution heat-treated and aged after fabrication to improve the performance of the weldment. Tensile, bend, impact, and low-cycle fatigue tests showed the component met all the specification requirements and elevated temperature properties were within base metal expectations with failures predominately in the weld metal. Long-term creep tests, including large samples more representative of the entire weldment, were fabricated and tested to times in excess of 10,000 h at multiple temperatures. Analysis of the creep data supports a WSRF of 0.9 for the solution annealed + aged weldments in contrast to the WSRF of 0.7 currently applied to welded + aged weldments. Metallurgical analysis shows that the solution annealing causes recrystallization of the weld metal and reduces the chemical compositional and microstructural gradients in the weldments resulting in acceptable short-term performance. In long-term creep, evidence for coarsened zones in the weldments due to discontinuous coarsening reactions were identified as the mechanism leading to accelerated creep damage formation in the weld metal and sample failure.

14 SOLAR ENERGY↗
DOE OSTI2016

Global Pattern of Nasopharyngeal Cancer: Correlation of Outcome With Access to Radiation Therapy

Purpose: This study aimed to estimate the treatment outcome of nasopharyngeal cancer (NPC) across the world and its correlation with access to radiation therapy (RT). Methods and Materials: The age-standardized mortality (ASM) and age-standardized incidence (ASI) rates of NPC from GLOBOCAN (2012) were summarized, and [1−(ASM/ASI)] was computed to give the proxy relative survival (RS). Data from the International Atomic Energy Agency (IAEA) and the World Bank were used to assess the availability of RT in surrogate terms: the number of RT equipment units and radiation oncologists per million population. Results: A total of 112 countries with complete valid data were analyzed, and the proxy RS varied widely from 0% to 83% (median, 50%). Countries were categorized into Good, Median, and Poor outcome groups on the basis of their proxy RS (<45%, 45%-55%, and >55%). Eighty percent of new cases occurred in the Poor outcome group. Univariable linear regression showed a significant correlation between outcome and the availability of RT: proxy RS increased at 3.4% (P<.001) and 1.5% (P=.001) per unit increase in RT equipment and oncologist per million population, respectively. The median number of RT equipment units per million population increased significantly from 0.5 in the Poor, to 1.5 in the Median, to 4.6 in the Good outcome groups, and the corresponding number of oncologists increased from 1.1 to 3.3 to 7.1 (P<.001). Conclusions: Nasopharyngeal cancer is a highly treatable disease, but the outcome varies widely across the world. The current study shows a significant correlation between survival and access to RT based on available surrogate indicators. However, the possible reasons for poor outcome are likely to be multifactorial and complex. Concerted international efforts are needed not only to address the fundamental requirement for adequate RT access but also to obtain more comprehensive and accurate data for research to improve cancer outcome.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2021

High-throughput ion irradiation of additively manufactured compositionally complex alloys

Several advanced nuclear reactor designs promise efficiency and safety improvements over the current reactor fleet but are limited by the current set of ASME code-qualified materials. Novel alloys including high-entropy alloys (HEAs), and more broadly compositionally complex alloys (CCAs), have shown promising irradiation-tolerance. However, the vast range of alloy compositions adds to an already time-consuming alloy development process. In this study, to accelerate the development of novel alloys for nuclear applications, a high-throughput (HTP) methodology has been employed. Additive manufacturing has been used to produce a compositional array of unary, binary, ternary, and quaternary alloys, including several CCAs, which span the Cr-Fe-Mn-Ni composition space. Additionally, the compositional array was homogenized at 1000°C for 24 hours and each sample was irradiated using 4-MeV Ni2+ ions at room temperature to a peak damage of 50 dpa, as estimated using SRIM, at the University of Wisconsin Ion Beam Laboratory. A custom XY stage was built to accommodate the large compositional array and half of each sample was masked during irradiation enabling both the irradiated and unirradiated properties of each alloy to be characterized side-by-side. Each alloy was characterized using X-ray fluorescence (XRF), X-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and nanoindentation. CALPHAD simulations spanning the entire Cr-Fe-Mn-Ni composition space at 1000°C were performed to compare predicted equilibrium phases with phases identified experimentally from the unirradiated regions of each alloy. Nanoindentation measurements indicate radiation-induced hardening ranging from ~1-1.5 GPa in each of Cr-Fe-Mn-Ni CCAs, which is relatively insensitive to modest changes in alloy composition and comparable to hardening observed in neutron irradiated Cr-Fe-Mn-Ni CCAs in the literature. Overall, a substantial time savings was realized by employing HTP synthesis, irradiation, and characterization in this study compared to conventional techniques, the implications of which are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Physical and thermal property changes under uniform oxidation in nuclear graphite

Material property changes of fine- and medium-grain nuclear graphite grades were measured after subjection to uniform oxidation. The cores of high temperature reactors are composed of large nuclear graphite block components. Here, these large core components are designed to provide neutron moderation and reflection, create a large thermal sink to assist in operational control, and form the solid core structure containing the nuclear fuel, coolant channels, and the safety critical channels for control rod insertion. Oxidation is a principle degradation mechanism affecting all aspects of the nuclear graphite component functions. This study addresses the underlying physical property changes of nuclear-graphite components for oxidized mass loss ranges beyond the current recommended ASME code rule limits to ensure structural integrity within the graphite components (a maximum mass loss = 10%).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Evaluation of corrosion coupons exposed to molten-chloride electrorefiner salts long term

Corrosion mitigation has long been considered a challenge in long-term service of both pyrochemical reprocessing and molten salt reactor applications. Corrosion coupons consisting of 2.25Cr 1Mo ASME SA387 Grade 22 Class 2 steel, were installed in a uranium electrorefiner containing molten chloride salt at 500°C. This electrorefiner went into service at Idaho National Laboratory in 1994. Seven of these coupons were removed in 2019 (i.e., submerged in molten salt for 25 years) for analysis and characterization including neutron imaging, optical microscopy, scanning electron microscopy, and wave/energy dispersive x-ray spectroscopy. The results showed that all seven coupons experienced between 30 to 175 µm of surface degradation without evidence of brittle fracture. Meaning there was no evidence of accelerated corrosion or fractures, and the average steel surface loss was of less than 5 µm per year. All indications were that the environmental conditions inside the electrorefiner were favorable for the steel alloy.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Burst pressure models and validations for thick-walled pipelines containing corrosion defects

Corrosion is one major threat to pipeline integrity. Over the past decades, many corrosion models have been developed for determining the remaining strength of corroded pipelines, including ASME B31.G, Modified B31.G, LPC, PCORRC and their modified models. All these corrosion models are applicable only to large diameter, thin-walled pipelines with a diameter to wall thickness ratio D/t ≥ 20. In practice, many pipelines have a small diameter and thick wall with a D/t ratio < 20, and thus an adequate corrosion model is needed for assessing remaining strength for corroded thick-walled pipelines. This paper briefly reviews the theoretical burst pressure models for defect-free thin and thick-walled pipelines and four representative corrosion assessment models for thin-walled corroded pipelines. On this basis, two modified corrosion models are proposed to thick-walled pipelines in terms of the average shear stress yield theory. To verify the proposed corrosion models, comprehensive validations are performed. Numerical validations include the elastic-plastic finite element analysis to determine burst pressure for pipelines without and with corrosion defects and the model evaluation using a large dataset of available FEA results of burst pressure for machined defects. Experimental validations include a set of burst pressure tests for defect-free thick-walled pipes with different thicknesses and the model evaluation using one large burst dataset for machined defects with flat bottoms and another large dataset for real corrosion defects with curved river bottom profiles. Both numerical and experimental validations show that the proposed corrosion models can more accurately predict the remaining strength for corroded thin and thick-walled pipelines.

Pipeline↗
DOE OSTI2018

Precipitation behavior during aging and creep in 18Cr–9Ni–3Cu–Nb–N steel

Highlights: • The creep strength of 18Cr–9Ni–3Cu–Nb–N steel degraded in the long term. • Nb rich MX particles precipitated in grain interior after aging for just one hour at 650 to 800 °C. • The time required for precipitation of modified Z phase was about 2500 h at 650 °C and 1050 h at 700 °C. • Fine Cu particles were detected in grain interior after aging for one hour. • Large Cu particles were located at the interface between the σ phase and the matrix after long-term aging at 650 °C and 700 °C. - Abstract: Microstructural changes during aging and creep were investigated in 18Cr–9Ni–3Cu–Nb–N steel (KA-SUS304J1HTB, ASME Code Case 2328), focusing on copper phase, MX carbonitride and Z phase. The creep strength of 18Cr–9Ni–3Cu–Nb–N steel was much higher than that of conventional SUS304 steel in the short term. However, the difference of these two steels was small in the long term, because the creep strength of 18Cr–9Ni–3Cu–Nb–N steel degraded in the long term. Niobium rich MX particles precipitated in grain interior after aging for just one hour at 650 to 800 °C. Modified Z phase [Cr(Nb, V)N] particles were observed after aging accompanied by the disappearance of MX particles. The time required for precipitation of modified Z phase was about 2500 h at 650 °C and 1050 h at 700 °C. Fine copper particles were detected in grain interior after aging for one hour. Large copper particles were located at the interface between the σ phase and the matrix after long-term aging at 650 °C and 700 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Effect of hold time on high temperature creep-fatigue behavior of Fe–25Ni–20Cr (wt.%) austenitic stainless steel (Alloy 709)

To understand high temperature creep-fatigue interaction of the Alloy 709, strain-controlled low-cycle fatigue (LCF) tests were performed at strain ranges varying from 0.3% to 1.2% with fully reversible cycle of triangular waveform at 750 °C. In addition, different hold times of 60, 600, 1800 and 3600 s were introduced at the maximum tensile strain to investigate the effect of creep damage on the fatigue-life at strain range of 1% at 750 °C. The creep-fatigue life and the number of cycles to macro-crack initiation and failure are found to decrease with increasing hold time indicating higher crack initiation and growth rates. Creep-fatigue life is evaluated by a linear summation of fractions of cyclic and creep damages according to ASME code. The fractographs of the samples deformed at 1% strain range indicated that fatigue might have been the dominant mode of deformation whereas, for the samples deformed at the same strain range with different hold times, both fatigue and creep have contributed to the overall deformation and fracture of the alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Statistical analysis of S—N type environmental fatigue data of Ni-base alloy welds using weibull distribution

In this study, the probabilistic fatigue life model for Ni-base alloys was developed based on the Weibull distribution using statistical analysis of fatigue data reported in NUREG/CR-6909 and the new fatigue data of Alloy 52M/152 and 82/182. The developed Weibull model can consider right-censored data (i.e., non-failed data) and quantify the improved safety (or reliability) based on the level of failure probability. The overall margin in the current fatigue design limit model (ASME design curve + NUREG/CR-6909 F en model) is similar to that of the Weibull model with a cumulative failure probability of approximately 2.5%. The margin in the current fatigue design limit model demonstrated inconsistencies for the Ni-base alloy weld data, whereas the Weibull model showed a consistent margin. Therefore, the Weibull model can systematically mitigate the excessive safety margin.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Advanced heat exchanger development for molten salts

This study addresses present work concerned with advanced heat exchanger development for molten salt in nuclear and non nuclear thermal systems. The molten salt systems discussed herein use alloys, such as Hastelloy N and 242, which show corrosion resistance to molten salt at nominal operating temperatures up to 700°C. These alloys were diffusion welded, and the corresponding information is presented. Test specimens were prepared for exposing diffusion welds to molten salt environments. Hastelloy N and 242 were found to be weldable by diffusion welding, with ultimate tensile strengths about 90% of base metal values. Both diffusion welds and sheet material in Hastelloy N were corrosion tested in?58 mol% KF and 42 mol% ZrF4 at 650, 700, and 850°C for 200, 500, and 1,000 hours. Corrosion rates found were similar between welded and nonwelded materials, typically <10 mils per year. For materials of construction, nickel and alloys with dense nickel coatings are effectively inert to corrosion in fluorides, but not so in chlorides. Hence, additional testing of selected alloys for resistance to intergranular corrosion is needed, as is a determination of corrosion rate as a function of contaminant type and alloy composition with respect to chromium and carbon to better define the optimal chromium and carbon composition, independent of galvanic or differential solubility effects. Also presented is the division of the nuclear reactor and high temperature components per ASME standards, along with design requirements for a subcritical Rankine power cycle heat exchanger that has to overcome pressure difference of about 17 MPa.

42 ENGINEERING↗
DOE OSTI2015

Full-scale 3-D finite element modeling of a two-loop pressurized water reactor for heat transfer, thermal–mechanical cyclic stress analysis, and environmental fatigue life estimation

This paper discusses a system-level finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent sequentially coupled thermal-mechanical stress analysis were performed for typical thermal-mechanical fatigue cycles. The in-air fatigue lives of example components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in US-NRC report: NUREG-6909.

Full-scale 3-D finite element model, Pressurized w↗
DOE OSTI2018

On the flow induced vibration of an externally excited nuclear reactor experiment

An in-pile, drop-in experiment design is presently being designed and studied for the near-term deployment within the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL); this experiment is termed the Miniplate-1 Large-B (MP-1 LB) Experiment. A number of explicit studies are performed during the design- and safety-related stage. Traditionally, a clear and logical methodology has been developed and utilized for analyses such as hydraulics, thermal-loads, mechanical loads, and others for these experiments. Recently a small component from a different experiment assembly mechanically separated while in the reactor’s core. While this experiment didn’t compromise the safety of the reactor, it led to a higher-level question which centered on whether the appropriate level of consideration was being made toward the fluid–structure-interactions of these experiments. The outcome yielded separate flow test experiments of like-for-like geometry in an experimental loop located at Oregon State University which produces experimental data compliant with applicable parts and requirements to ASME’s NQA-1 2008, 2009a standard – suitable for benchmark evaluation. Here, the objectives of this study are to (1) develop a process for handling and interpreting the mechanical response of the test elements during hydraulic testing, to (2) characterize the motion of a specific test element during a flow test which imposes a wide range of hydraulic conditions, and (3) provide objective observations toward the potential safety related implications that are tied to the synthesized data. The outcome of this study has led to a confident process in inspecting the experimental data, synthesized it for interpretation, identified several unique hydraulic characteristics of the experiment design which were previously unknown, and demonstrated that the likelihood for mechanical failure resulting from fluid-structure-interactions in the reactor is far below any criterion for concern of the element’s safety.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Seismic isolation: A pathway to standardized advanced nuclear reactors

Standardizing advanced nuclear reactors is a pathway to substantially reducing their overnight capital cost and achieving parity with other power sources, including renewables and fossil fuels. The seismic load case has thwarted standardization of nuclear power plants because site-specific seismic hazard and local near-surface geology has triggered soil-structure-interaction analysis, design, equipment qualification, regulatory review, and licensing, ensuring that each build is different. To achieve standardized or site-independent certified advanced reactor designs, the impact of the seismic load case on the engineering and construction cost and time must be substantially mitigated. Seismic isolation is a mature technology that has been used for more than 30 years in non-nuclear sectors to substantially reduce earthquake demands in buildings and other infrastructure. In this paper, seismic isolation is used to enable standardization of advanced reactor designs, aimed at the complete re-use of a site-independent, certified design and repeated procurement of safety-class equipment. A pathway to standardized designs using seismic isolation is demonstrated for two fundamentally different advanced reactors: a molten salt reactor and a high temperature gas reactor. Each reactor building is equipped with three specialized pieces of safety-class equipment, namely, a reactor vessel, a steam generator, and a control rod drive mechanism housing that is attached to the reactor head. Analysis is performed per ASCE and ASME standards to design the buildings and the equipment for two base conditions: conventional (fixed base) and base isolated. The impact of the seismic load case is characterized for the reinforced concrete walls in the buildings and for the equipment, measured using vessel wall thickness and horizontal accelerations. Here the analysis results show that the fixed-base buildings, designed for a site of low seismic hazard (peak ground acceleration, PGA = 0.15 g) could be constructed at a site of much greater seismic hazard (PGA = 0.7 g) if seismic base isolation is employed. Importantly, the scope of the site-specific analysis, design, and qualification would be limited to the seismic isolators and the isolated substructure, drastically reducing plant-specific engineering, review, and licensing, and time to construction start. Regulatory challenges and opportunities with standardized reactor designs are identified.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Codes and standards for ceramic composite core materials for High Temperature Reactor applications

Fiber-reinforced ceramic matrix composites are attractive for high-temperature nuclear applications due to excellent thermal and mechanical properties as well as reasonable-to-outstanding radiation resistance. Over the past 20 years, the use of ceramic matrix composite applications expanded to many commercial non-nuclear industries as fabrication and application of the technologies mature. The ASME Boiler Pressure Vessel Code, under Section III Division 5, provides the design and construction rules for High Temperature Reactor components. It published the first rules for ceramic matrix composites to be used for reactor core components. The rules lay out the quality requirements together with the design and materials criteria for the use and application of silicon carbide- and carbon-based matrix material technologies. As with the established graphite rules, the ceramic composite material rules are structured in Subsection HH (from Section III), that addresses the criteria for class SN nonmetallic core components. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in ASTM Committee C28 on Advanced Ceramics, with a current focus on ceramic composite tubes. This article describes the detail of the composites code, the design methodology and similarities to the graphite code, the guidance for the development of specifications for ceramic composites (for nuclear applications) including recent standard developments, and it mentions the next steps to support licensing aspects by validating the code with benchmarking data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗