Search NASA⌕ Search

SEARCH · Search NASA

Results for “REACTOR OPERATION”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

The role of material modeling on strain range estimation for elevated temperature cyclic life evaluation

High temperature nuclear reactors operating in the creep regime are designed to withstand numerous cyclic events. Current ASME code rules provide two basic paths for evaluating creep fatigue and ratcheting under these conditions; one based on full inelastic analysis intended to provide a representative stress and strain history and the other based on elastic material models with adjustments of varying complexity to account for inelastic stress and strain redistribution. More recent developments have used elastic-perfectly plastic analysis to bound the effects of cyclic service. However, these methods still rely on the separate evaluation of fatigue and creep damage utilizing a damage interaction diagram. There is a procedure under current development that uses creep-fatigue data from key feature test articles directly without the use of the damage interaction diagram. However, it requires a reasonable representation of the strain range in a structure as an input. This work develops a simplified procedure based on elastic perfectly-plasticity analysis that can be used to represent the strain range in a structure in the steady state under cyclic loading conditions.

Messner, M. C.↗
DOE OSTI2019

Interaction of Long-lived Reactive Species from Cold Atmospheric Pressure Plasma with Polymers: Chemical Modification by Ozone and Reactive Oxygen-Nitrogen Species

Atmospheric pressure plasma (APP) sources are able to generate a variety of reactive species that have different effects on materials, such as functionalization, etching and deposition. In this article, we study the effect of long-lived reactive neutral species on polymers using a model plasma-surface interaction (PSI) system that consists of ultra-thin (~10 nm) polystyrene (PS) films and a surface micro-discharge (SMD) reactor operated with various N 2 /O 2 working gas mixtures. We characterized and quantified the reactive species generated by SMD using IR and UV absorption, and we found that O 3 , N 2 O 5 , N 2 O and HNO 3 are the dominant long-lived reactants near the target surface. When exposing PS films to these reactive species, we observed material responses including film thickness expansion, surface and bulk oxidation, and surface organic nitrate formation. The quantity of these changes varied with the N 2 /O 2 working gas composition. By correlating material response with gas phase species, we find that the chemical modification of PS strongly depends on the density of O 3 in the gas phase, which is indicative of an essential role of O 3 in the remote APP treatment of polymers. Our results show that O 3 causes polymer surface oxidation, participates in the diffusion-reaction process in the polymer bulk, and results in aromatic ring cleavage and the formation of carbonyl groups. In contrast, we did not find a correlation between surface organic nitrate and individual long-lived reactive species mentioned above. This indicates that surface organic nitrate formation on polymer surfaces might result from the interaction of multiple reactive species, including O 3 and nitrogen containing species. A model for the interphase mass transfer of reactive species from gas to solid was also described.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Physical Parameters of a Reactor-Stellarator with Small Ripples of the Helical Magnetic Field

The paper describes the calculation data on the physical parameters of a reactor-stellarator, where the nonuniformities of the helical field are smaller than the toroidal magnetic field nonuniformities: ε{sub h} < ε{sub t}. Unlike the previous studies, where the ion-component transport coefficients had the collision frequency dependence proportional to ν{sup 1/2}, this being equivalent to the ε{sub h} > ε{sub t} case, in the present calculations, these coefficients were assumed to be in proportion to the first power of the collision frequency, D{sub i} ∝ ν for ν{sub eff} < 2ω{sub E}, and to D{sub i} ∝ ν{sup −1} for the inverse inequality. Here, ω{sub E} is the rotation frequency of plasma in the radial electric field. As before, the plasma electrons corresponded to the mode of D{sub e} ∝ ν{sup −1}. As initial parameters for numerical calculations, a reactor with R = 8 m, r{sub p} = 2 m, and B{sub 0} = 5 Т was taken. A numerical code was used to solve the set of equations that describes the plasma space−time behavior in the reactor-stellarator under the conditions of equal diffusion fluxes. The start of reactor operation in the mode of thermonuclear burning was provided by heating sources with a power of several tens of megawatts. Steady-state operating conditions of a self-sustained thermonuclear reaction were attained by maintaining the plasma density through DT fuel pellet injection into the plasma.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Acoustic Data Analytics For Predictive Technologies

The software utilizes frequency, amplitude, and phase features from sensor signals (either raw or processed) that are extracted using recursive frequency analysis, enabling (1) understanding of reactor operations under different operating conditions, (2) identifying of quiescent frequency ranges for sensor development, (3) system controls (4) predictive analytics, (5) structural health monitoring, and (6) loose part monitoring.

Smith, JamesA.↗
DOE OSTI2024

Radiological Safety Analysis Computer Program V 7.2

Users generates a fission product inventory from either reactor operating history or a nuclear criticality event. RSAC-7 models the effects of high-efficiency particulate air filters or other cleanup systems and calculates the decay and ingrowth during transport through processes, facilities, and the environment. Doses are calculated for inhalation, air immersion, ground surface, ingestion, and cloud gamma pathways. RSAC-7 is used as a tool to evaluate accident conditions in emergency response scenarios, radiological sabotage events, and safety basis accident consequences.

Jordan, RobertaA. [Idaho National Laboratory (INL)↗
DOE OSTI2025

Divertor Plasma Detachment Control Neural Network

DivControlNN is a state-of-the-art software tool that leverages advanced machine learning techniques to predict and control divertor plasma behavior in fusion reactors. Plasma, a highly energetic and electrically charged gas, requires meticulous management to protect reactor components and maintain optimal energy production. Conventional simulation methods, although extremely detailed, typically demand extensive computational time-making them unsuitable for real-time control scenarios. DivControlNN addresses this challenge by learning from tens of thousands of high-fidelity simulations, thereby creating a rapid surrogate model that can deliver near-instantaneous predictions. At the core of its functionality is a sophisticated technique known as latent space mapping, which condenses complex, high-dimensional plasma data into a compact, lower-dimensional representation. This streamlined representation enables the system to quickly forecast essential plasma properties and determine the precise conditions required for effective detachment. Detachment is a crucial process in which the plasma is cooled before reaching the divertor plates, thereby reducing heat loads and mitigating material erosion. In recent experiments conducted on the KSTAR tokamak in South Korea, DivControlNN successfully guided the detachment process without any fine-tuning-even when applied to a new tungsten divertor configuration. By achieving a computational speed-up of over one hundred million times compared to traditional simulation methods while maintaining low prediction errors, DivControlNN stands to significantly enhance real-time control and diagnostic capabilities in future fusion reactors. This breakthrough paves the way for safer, more reliable reactor operation and represents a major advancement toward realizing fusion energy as a practical, sustainable, and clean power source.

Xu, Xueqiao [Lawrence Livermore National Laborator↗
DOE OSTI2025

ADVANCED REACTOR ARCHETYPE LIBRARY

SF-25-142 An archetype library for Cyclus that can model more complex reactor operations.

Bachmann, Amanda [Argonne National Laboratory (ANL↗
DOE OSTI2024

Data for Pilot-Scale Processing of Miscanthus x giganteus for Recovery of Anthocyanins Integrated with Production of Microbial Lipids and Lignin-Rich Residue

Chemical-free hydrothermal pretreatment of Miscanthus x giganteus (Mxg) at the lab scale using high liquid-to-solid ratios resulted in the recovery of anthocyanins and enhanced enzymatic digestibility of residual biomass. In this study, the process is scaled up by using a continuous hydrothermal pretreatment reactor operated at a low liquid-to-solid ratio (50 % w/w solids) as an important step towards commercialization. Anthocyanin yield was 70 % w/w at the pilot scale (50 kg of Mxg), compared to the 94 % w/w yield achieved at the lab scale (0.5 g of Mxg). The pretreated biomass was subsequently refined mechanically using a disc mill to increase the accessibility of cellulose by cellulases. Enzymatic saccharification of the pretreated and disc-milled residue yielded 238 g/L sugar concentration by operating in fed-batch mode at 50 % w/v solids content. Two strains of Rhodosporidium toruloides were evaluated for converting the hydrolysate sugars into microbial lipids, and strain Y-6987 had the highest lipid titer (11.0 g/L). Further, the residue left after enzymatic saccharification was determined to be enriched 1.7-fold in the lignin content. This lignin-rich residue has value as a feedstock for the production of sustainable aviation fuel precursors and other high-value lignin-based chemicals. Hence the proposed biorefinery based on Mxg creates an opportunity for generating revenue from multiple high-value products. As the demand for biofuels and biobased products is rising, the biorefinery products from Mxg would create a niche in the industrial sector.

Conversion↗
DOE OSTI2023

Development of a Scalable Risk-informed Predictive Maintenance Cloud-based Strategy at Nuclear Power Plants

The fact that light-water reactor operation and maintenance costs are prohibitively expensive and contribute to the premature decommissioning of nuclear power plants is partly due to how the equipment is monitored. In recent years, cloud computing has emerged as a dominant technology, as its low cost, computing and storage adaptability, and ability to host applications across numerous virtual infrastructures potentially make it a cost-effective alternative to onsite storage and diagnostics. In this paper, a technological assessment is carried out on a provisional cloud deployment architecture for a nuclear power plant predictive monitoring system. This cloud-based monitoring system would enable maintenance and diagnostic analysts and other authorized plant users to remotely monitor equipment functionality, thus enabling early fault detection and effective predictive maintenance practices. To provide data processing and storage, sensor device networking, and database management, the Microsoft Azure cloud platform is utilized as part of the proposed cloud architecture; however, this analysis could be extended to other cloud computing service providers as well. The focus of this paper is on application of cloud resources for enabling predictive maintenance, identification of technological hurdles associated with moving to a cloud-computing-based architecture, and potential benefits from moving to a centralized cloud system.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Fuzzy logic model of irradiated aggregates

The worldwide need for nuclear power plant (NPP) lifetime extension to meet future national energy requirements while reducing greenhouse gases raises the question of the condition of concrete structures exposed to ionizing radiation. Although research into the effects of radiation has a long history and the phenomenon of deterioration of concrete due to irradiation is not yet completely understood, the main assumed degradation mode is radiation-induced volumetric expansion of aggregates. There are experimental data on irradiated concrete obtained over decades under different conditions; however, the collection of data exhibits considerable scatter. Fuzzy logic modeling offers an effective tool that can interconnect various data sets obtained by different teams of experts under different conditions. The main goal of this work is to utilize available data on irradiated concrete components such as minerals and aggregates that expand upon irradiation. Furthermore, aggregate radiation-induced volumetric expansion gives an estimate of the change in mechanical properties of aggregate after years of reactor operation. The mechanical properties of irradiated aggregate can then be used for modeling irradiated concrete in the actual NPP structure based on the composition of concrete, the average temperature on the surface of the biological shield structure, and the neutron dose received by biological shield.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Materials and Fuels Complex FY-18--FY-23 Five-Year Investment Strategy

The Department of Energy Office of Nuclear Energy (DOE-NE) mission is to provide the research, development, and demonstration foundation to extend the lives of the current operating reactor fleet, develop the next generation of nuclear reactors, and provide integrated nuclear fuel cycle solutions. This document is a complimentary document to the Materials and Fuels Complex Integrated Five-Year Science Strategy (INL/EXT-17-41117, Rev. 1) that defines specific implementation strategies for increasing research capability and throughput through targeted investments in research facilities, research instruments, and research staff.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

The H-Modified Process

Construction of the H-Area canyon separation facility was completed in the early 1950's. Processing of irradiated depleted Uranium fuel using the Purex flowsheet began in July 1955. The introduction of enriched uranium fuels into Savannah River reactor operations was to produce a greater variety of products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Modeling Radionuclide Releases from TRISO Particles by Simultaneous Diffusion Through and Corrosion of the Silicon Carbide Barrier Layer

TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 μm thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 °C). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10 -20 m 2 /s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10 -21 m 2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Materials and Fuels Complex FY19 - FY23 Five-Year Investment Strategy

The Department of Energy Office of Nuclear Energy (DOE-NE) mission is to provide the research, development, and demonstration foundation to extend the lives of the current operating reactor fleet, develop the next generation of nuclear reactors, and provide integrated nuclear fuel cycle solutions. The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear research, development, and demonstration (RD&D) enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2019

Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems 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. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using 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 SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. 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 for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Materials and Fuels Complex Five-Year Investment Strategy (FY2020-FY2024)

The Department of Energy Office of Nuclear Energy (DOE-NE) mission is to provide the research, development, and demonstration (RD&D) foundation to extend the lives of the current operating reactor fleet, develop the next generation of nuclear reactors, and provide integrated nuclear fuel cycle solutions. The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy. This document is a complimentary document to the Materials and Fuels Complex Five-Year Mission Strategy FY-20 – FY-24 (INL/EXT-20-57224) that defines specific implementation strategies for increasing research capability and throughput through targeted investments in research facilities, research instruments, and research staff. Refer to INL/EXT-20-57224 for details about MFC, its capabilities, and the overall mission strategy for the Materials and Fuels Complex.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Demonstration of the Coating Analysis Capabilities in BISON

In order to improve the accident tolerance of light water reactor (LWR) fuel, coated cladding materials have been proposed which exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high-temperature steam, making accident mitigation more likely. In order to determine how these coating materials will adjust the behavior of the Zr-alloy cladding during normal reactor operation, simulations have been performed using the Bison fuel performance code. This analysis focused on modeling and comparing the integral thermo-mechanical performance of coated and uncoated zircaloy cladding with uranium dioxide fuel under normal operating conditions. This analysis demonstrates the ability to model a discrete meshed coating bonded to the cladding exterior and provides a comparison with an uncoated fuel rod. A generic range of coating material properties are implemented given the differences and unknowns regarding the proprietary coating properties used by the vendors. A main-effects analysis was performed to identify trends and sensitivities in how coating material properties affect key fuel performance metrics. Lastly, because residual stresses are expected to form in the coating, arising from the high temperature coating application process and application techniques, an additional analysis was performed to identify the impact of these stresses on the cladding behavior. Future analysis efforts will target more comprehensive documentation of this main effects analysis by sampling the results at the beginning and middle-of-life in order to capture important behavior before and during the onset of gap closure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Pitch Variation Experiments in Water-Moderated Square-Pitched U(6.90)O 2 Fuel Rod Lattices with Fuel to Water Volume Ratios Spanning 0.08 to 0.67

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad square-pitched U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMPTHERM- 096, LEU-COMP-THERM-097, and LEU-COMP-THERM-101. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The purpose of these experiments was to measure the effects of decreasing the fuel-to-water volume ratio on the critical array size. This was accomplished by removing fuel rods from fully fueled configurations, effectively increasing the pitch of the fuel arrays in the assembly. The fuel rod pitch variations provided assembly configurations that ranged from strongly undermoderated to slightly overmoderated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗