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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Fundamental Understanding of Ambient and High-Temperature Plasticity Phenomena in Structural Materials in Advanced Reactors

The goal of this research project is to develop the methods and tools necessary to link unit processes analyzed using atomistic simulations involving interaction of vacancies and interstitials with dislocations, as well as dislocation mediation at sessile junctions and interfaces as affected by radiation, with cooperative influence on higher-length scale behavior of polycrystals. These tools and methods are necessary to design and enhance radiation-induced damage-tolerant alloys. The project will achieve this goal by applying atomistic simulations to characterize unit processes of: 1. Dislocation nucleation, absorption, and desorption at interfaces 2. Vacancy production, radiation-induced segregation of substitutional Cr at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels 3. Investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S) 4. Time evolution of swelling (cluster growth) phenomena of irradiated materials 5. Energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip This project will consider the Fe, Fe-C, and Fe-Cr ferritic/martensitic material system, accounting for magnetism by choosing appropriate interatomic potentials and validating with first principles calculations. For these alloys, the rate of swelling and creep enhancement is considerably lower than that of face-centered cubic (FCC) alloys and of austenitic Fe-Cr-Mo alloys. The team will confirm mechanisms, validate simulations at various time and length scales, and improve the veracity of computational models. The proposed research's feasibility is supported by recent modeling of radiation effects in metals and alloys, interfacial dislocation transfer reactions in nano-twinned copper, and dislocation reactions at general boundaries, along with extensive modeling cooperative effects of dislocation interactions and migration in crystals and polycrystals using continuum models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

A High Temperature-Tolerant and Radiation-Resistant In-Core Neutron Sensor for Advanced Reactors. Final report

The objectives of this project are to develop a small and reliable gallium nitride (GaN) neutron sensor that is capable of withstanding high neutron fluence and high temperature, isolating gamma background, and operating in a wide dynamic range. The first objective will be the understanding of the fundamental materials properties and electronic response of a GaN semiconductor materials and device in an environment of high temperature and intense neutron field. To achieve such goal, an in-situ study of electronic properties of GaN device such as I-V, leakage current, and charge collection efficiency (CCE) in high temperature using an external neutron beam will be designed and implemented. We will also perform in-core irradiation of GaN up to the highest yet fast neutron fluence and an off-line performance evaluation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Feasibility and Safety Assessment for Advanced Reactor Concepts Using Vented Fuel

Recent interest in fast reactor technology has led to renewed analysis of past reactor concepts such as Gas Fast Reactors and Sodium Fast Reactors. In an effort to make these reactors more economic, the fuel is required to stay in the reactor for extended periods of time; the longer the fuel stays within the core, the more fertile material is converted into usable fissile material. However, as burnup of the fuel-rod increases, so does the internal pressure buildup due to gaseous fission products. In order to reach the 30 year lifetime requirements of some reactor designs, the fuel pins must have a vented-type design to allow the buildup of fission products to escape. The present work aims to progress the understanding of the feasibility and safety issues related to gas reactors that incorporate vented fuel. The work was separated into three different work-scopes: 1. Quantitatively determine fission gas release from uranium carbide in a representative helium cooled fast reactor; 2. Model the fission gas behavior, transport, and collection in a Fission Product Vent System; and, 3. Perform a safety analysis of the Fission Product Vent System. Each task relied on results from the previous task, culminating in a limited scope Probabilistic Risk Assessment (PRA) of the Fission Product Vent System. Within each task, many key parameters lack the fidelity needed for comprehensive or accurate analysis. In the process of completing each task, the data or methods that were lacking were identified and compiled in a Gap Analysis included at the end of the report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Radiation behavior of high-entropy alloys for advanced reactors. Final report

In the first task, we have demonstrated the radiation damage and the recrystallization behaviors in multicomponent alloys through molecular-dynamics simulations. It is found that by alloying with atoms of different sizes, the atomic-level strain increases, and the propensity of the radiation-induced crystalline to amorphous transition increases as the defects cluster in the cascade body. Recrystallization of the radiation induced supercooled or glass regions show that by tuning the composition and the equilibrium temperature, the multicomponent alloys can be healed. The crystalline-amorphous-crystalline transitions predict the potential high radiation resistance in multicomponent alloys. In the second task, three types of high-entropy alloys (HEAs) were fabricated from AlCoCrFeNi and AlCuCrFeNi quinary alloys. Hardness and reduced contact modulus were measured using nanoindentation tests. Heavy ion irradiation were performed using 10 MeV gold and 5 MeV nickel to study radiation effects. Al 0.5 CrCuFeNi 2 shows phase separation upon the presence of copper. Both hardness and contact modulus exhibit the same trend as increasing the applied load, and it indicates that excessive free volume may alter the growth rate of the plastic zone. The as-cast Al 0.1 CoCrFeNi specimen undergone the hot isostatic pressing (HIP) process and steady cooling rate which mitigate the quenching effect. The swelling behavior was characterized by the atomic force microscopy (AFM), and the swelling rate is approximately 0.02% dpa. Selected area diffraction (SAD) patters show irradiation-induced amorphization throughout the ion projected range. Within the peak damage region, an amorpous ring is observed, and a mixture of amorphous/ crystalline structure at deeper depth is found. The Al 0.3 CoCrFeNi HEAs shows good radiation resistance up to 60 peak dpa. No voids or dislocations are observed. The crystal structures remain face-centered-cubic (FCC) before and after 5 MeV Ni irradiation. Higher dpa might be required to study defects formation mechanisms. In the third task, all the constituent binary and ternary systems of the Al-Co-Cr-Fe-Ni system were thermodynamically modeled within the whole composition range. Comparisons between the calculated phase diagrams and literature data are in good agreement. The multi-component thermodynamic database of the Al-Co-Cr-Fe-Ni system was then obtained via extrapolation. The current Al-Co-Cr-Fe-Ni thermodynamic database enables us to carry out the calculations of phase diagrams, which can be used as useful guidelines to identify the Al-Co-Cr-Fe-Ni HEAs with desirable microstructures. In the fourth task, we discuss how as-cast and homogenized phases can be identified, what phases are usually found in the as-cast and homogenized conditions, and what the thermodynamics and kinetics of phase transformations are in the AlCoCrFeNi HEA. The microstructure and phase composition were studied in as-cast and homogenized conditions. It showed the dendritrical structure in the as-cast condition consisting primarily of a nano-lamellar mixture of A2 [disorder body-centered-cubic (BCC)] and B2 (ordered BCC) phases, in addition to a very small amount of A1 [disorder face-centered-cubic (FCC)] phases. The homogenization heat treatment resulted in an increase in the volume fraction of the A1 phase and formation of a Sigma phase. Tensile properties in as-cast and homogenized conditions are reported at 700 °C. Thermodynamic modeling of non-equilibrium and equilibrium phase diagrams for the AlCoCrFeNi HEA gave good agreement with the experimental observations of the phase contents. The reasons for the improvement of ductility after the heat treatment are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O 2 cannot be ignored, especially for the FHR, in which environment the product, SiO 2 , can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

FHR↗
DOE OSTI2016

Novel Engineered Refractory Materials for Advanced Reactor Applications

This report summarizes the results of DOE-NEUP grant 10-853. The project spanned 48 months (36 months under the original grant plus a 12 month no cost extension). The overarching goal of this work was to fabricate and characterize refractory materials engineered at the atomic scale with emphasis on their tolerance to accumulated radiation damage. With an emphasis on nano-scale structure, this work included atomic scale simulation to study the underlying mechanisms for modified radiation tolerance at these atomic scales.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Proposed Activities to Address Regulatory Gaps and Challenges for Licensing Advanced Reactors Using Seismic Isolation

Over the last decade, particularly since implementation of the certified design regulatory approaches outlined in 10 CFR 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” interest has been increasing in the use of seismic isolation (SI) technology to support seismic safety in nuclear facilities. In 2009, the United States (U.S.) Nuclear Regulatory Commission (NRC) initiated research activities to develop new guidance targeted at isolated facilities because SI is being considered for nuclear power plants in the U.S. One product of that research, which was developed around a risk-informed regulatory approach, is a draft NRC NUREG series (NUREG/CR) report that investigates and discusses considerations for use of SI in otherwise traditionally founded large light water reactors (LWRs). A coordinated effort led to new provisions for SI of LWRs in the American Society of Civil Engineers standard ASCE/SEI 4-16, “Seismic Analysis of Safety Related Nuclear Structures.” The risk-informed design philosophy that underpinned development of the technical basis for these documents led to a set of proposed performance objectives and acceptance criteria intended to serve as the foundation for future NRC guidance on the use of SI and related technology. Although the guidance provided in the draft SI NUREG/CR report and ASCE/SEI 4 16 provides a sound basis for further development of nuclear power plant designs incorporating SI, these initial documents were focused on surface-founded or near-surface-founded LWRs and were, necessarily, limited in scope. For example, there is limited information in both the draft NUREG/CR report and ASCE/SEI 4-16 related to nonlinear analysis of soil-structure systems for deeply-embedded reactors, the isolation of components, and the use of vertical isolation systems. Also not included in the draft SI NUREG/CR report are special considerations for licensing of isolated facilities using the certified design approach in 10 CFR 52 and a detailed discussion of seismic probabilistic risk assessments for isolated facilities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Advanced reactor site feasibility evaluation

The report provides a description of the project, summarizes the methodology and results for each task of the project, and ends with the overall project conclusions and recommendations. In addition, the report contains a descriptive index of the technical reports generated during the course of the project.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Fission Fragment Yield Data in Support of Advanced Reactor Technology

Within the 3 year POP we propose to continue to test and further develop the fission spectrometers, to do development tests and full data acquisition run at the national laboratory neutron beam facilities, to measure correlated fission fragment yields at low neutron energies with 235 U fission targets, and make these data available to the nuclear community. The spectrometer development will be both on the university based r\prototype and on the National Laboratory Spectrometer, and measurements will be performed with both. Over the longer time frame of the collaboration, we will take data over a range of low energies, and use other fission targets available to the laboratory. We will gather energy specific fragment distributions and reaction cross sections. We will further develop the data acquisition capabilities to take correlated fission fragment'gamma ray/neurton data, all on an event-by-event basis. This really is an enabling technology.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Tritium Mitigation/Control for Advanced Reactor System

A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent the residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: To estimate tritium permeation behavior in FHRs; To design a tritium removal system for FHRs; To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities were designed and laboratory-scale experiments were proposed for the validation of the proposed tritium removal facilities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Risk-Informed Performance-Based Technology. Inclusive Guidance for Advanced Reactor Licensing Basis Development (NEI 18-04)

This guideline presents a modern, technology-inclusive, risk-informed, and performance-based (TI-RIPB) process for selection of Licensing Basis Events (LBEs); safety classification of structures, systems, and components (SSCs) and associated risk-informed special treatments; and determination of defense-in-depth (DID) adequacy for non-LWRs. This guidance document provides one acceptable means for addressing the aforementioned topics as part of demonstrating a specific design provides reasonable assurance of adequate radiological protection.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

First Annual Report on Development of Microwave Resonant Cavity Transducer for Fluid Flow Sensing: Development of Sensor Performance Model of Microwave Cavity Flow Meter for Advanced Reactor High Temperature Fluids

We are investigating a microwave cavity-based transducer for in-core high-temperature fluid flow sensing in molten salt cooled reactors (MSCR) and sodium fast reactors (SFR). This sensor is a hollow metallic cylindrical cavity, which can be fabricated from stainless steel, and as such is expected to be resilient to radiation, high temperature and corrosive environment of MSCR and SFR. The principle of sensing consists of making one wall of the cylindrical cavity flexible enough so that dynamic pressure, which is proportional to fluid velocity, will cause membrane deflection. Membrane deflection causes cavity volume change, which leads to a shift in the resonant frequency. Feasibility of the sensor was initially investigated with analytical derivations and with COMSOL RF Module computer simulations of resonant frequency spectral shift due to uniform load. We also investigated the mechanical integrity of the flowmeter’s membrane through analytical modelling and COMSOL Structural Mechanics Module computer simulations. Both the analytic model and COMSOL model showed that maximum stresses on the plate, which are at the radial boundary of the plate, are three orders of magnitude smaller than the material’s yield strength and ultimate tensile strength. This indicates that the sensor is at a low risk of mechanical failure. Using results from models, we have developed an initial design for a microwave K-band sensor. A cylindrical resonator prototype was fabricated from brass for the initial tests. The external dimensions of the cavity are matched to the flange of a standard WR-42 waveguide. Microwave field is coupled into the resonant cavity through a subwavelength-size aperture. A test article was developed consisting of a piping Tee with a bulkhead WR-42 microwave waveguide installed in a leak-proof assembly. A microwave waveguide circulator was installed in the setup to suppress the effect of reflections at the cavity entrance by increasing the isolation between the input and the output port. Preliminary spectral characterization of cavity spectral response was performed with a portable PXIe chassis microwave VNA with a custom GUI. Preliminary dry tests of the transducer response were conducted with a set of calibrated weights. Transducer frequency shift was shown to be monotonically increasing with increasing pressure. The next steps will involve investigation of the transducer performance for water flow sensing.

42 ENGINEERING↗
DOE OSTI2021

Enhancing the Operational Resilience of Advanced Reactors with Digital Twins by Recurrent Neural Networks

Because of a lack of operational data and uncertainty in evaluation model for abnormal and accident scenarios, the established operating procedures can be biased in characterizing the reactor states and ensuring operational resilience. To reduce uncertainty associated with actual plant conditions, digital twin (DT) technology is suggested to support operator’s decision-making by effectively extracting and using knowledge of the current and future plant states from the knowledge base. This study first builds a knowledge base based on the characterization of issue space and the simulation tool. Next, this study discusses diagnosis and prognosis DTs for enhancing operational resilience by recovering the complete states of reactors and by predicting the future reactor behaviors. Finally, the decision-making module of the control system can determine the optimal control strategy that meets operational goals during loss-of-flow scenarios. To demonstrate and evaluate the DTs capability for supporting the operations of nuclear reactors, this study develops and assesses both the diagnosis and prognosis DTs in a nearly autonomous management and control system for an Experimental Breeder Reactor-II simulator during different loss-of-flow scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗