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 1,279 records · Page 71
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Fast neutron irradiation capability in existing thermal test reactors

In today’s nuclear industry, momentum towards the design, licensing, and construction of advanced nuclear demonstration plants, including fast reactors, is at a remarkably high level. However, there are currently no dedicated fast spectrum irradiation test facilities in the United States to support the development of fast spectrum technologies. As a result, a unique situation is developing where most of these plants will likely be designed by leveraging historic nuclear material technologies, but where the further optimization and advancement is impeded by the lack of fast neutron irradiation test facilities. While these circumstances present a challenge, there are some near-term opportunities that, if seized, can still help develop advanced fast reactor materials to a meaningful level of readiness to support future commercial fast reactors. Here, in this paper, we assess the feasibility of using thermal neutron filtering materials in existing experiment positions in the Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory to simulate fast reactor test environments for nonfuel test specimens. Items investigated include the incident neutron flux (both fast and thermal), the total neutron fluence and cumulative atom displacements, helium production rate due to thermal neutron capture in nickel, and the potential impact that the thermal neutron filter material has on the cycle length of a given reactor. It is concluded that while HFIR provides the highest fast flux of all the options investigated, it is limited in the amount of thermal neutron filtering material that can be introduced into an experiment position without significantly affecting the operation of the reactor. Irradiation in Outboard-A positions in the ATR was found to be the most realistic near-term experiment avenue due to having ample space for several capsules in a moderately fast flux.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Large tetragonality and room temperature ferroelectricity in compressively strained CaTiO 3 thin films

We report that ferroelectricity and piezoelectricity are desirable for a variety of high-temperature applications such as actuators and sensors in heat engines, high-temperature manufacturing, and space technologies; however, the number of such material candidates is currently limited. Here, we demonstrate that CaTiO 3 , the first discovered perovskite mineral that is abundantly found in the Earth, as a non-polar bulk form, becomes a high-temperature ferroelectric oxide under compression. A strain-phase-temperature diagram of CaTiO 3 is created by growing a series of thin films with a range of compressive and tensile strains. Using temperature dependent optical second harmonic generation analysis, we show that tensile strained films exhibit predominantly in-plane polarization with orthorhombic-like point group symmetry with a phase transition below room temperature. On the other hand, compressively strained CaTiO 3 films exhibit a neartetragonal unit cell with a c/a ratio of 1.03, larger than that of classic ferroelectric, e.g. BaTiO 3 (c/a~1.01). These films exhibit a robust and switchable out-of-plane polarization at room temperature, with a ferroelectric transition temperature up to ~800 K. Density functional theory calculations reveal that compressive strain gives rise to a large out-of-plane displacement of Ti-cations inside the TiO 6 octahedral cages responsible for the strain induced polarization of ~ 9μC/cm 2 . Given that nearly half of the perovskites exhibit the bulk symmetry of CaTiO 3 , compressive strain tuning of this family may prove to be a fertile ground for the discovery of strain-induced piezoelectrics and ferroelectrics at high-temperatures. High-temperature piezoelectric sensors are important for many industries such as automotive, aircraft, and aerospace, as well as nuclear and geothermal energy, where dynamic structural monitoring of high-speed rotors in engines, turbines, and power generators is critical for safety and reliability.[1, 2] Modern sensors are however limited, where the main challenges are maximizing operating temperatures while maintaining good sensing capability and low electrical leakage.[3] High-temperature piezoelectric materials are typically derived from BaTiO 3 or PbTiO 3 based ceramics and single crystals, as these materials are also ferroelectric, meaning that they have a built-in spontaneous polarization (Ps). In bulk, these standard ferroelectric perovskite materials yield maximum operating temperatures ranging from 250-450 °C.[4], where above these temperatures the spontaneous polarization (P) and piezoelectric properties disappear due to the onset of a ferroelectric-to-paraelectric phase transition (TC). As next-generation engines, turbines, and reactors require hotter operating temperatures above 500 °C, materials with enhanced ferroelectric properties and increased ferroelectric transition temperatures will be needed in such extreme conditions. A promising way of boosting ferroelectric transition temperatures and polarization is through epitaxial strain engineering,[5] where tremendous in-plane distortions can be applied to thin film crystal materials, well beyond the point at which their bulk ceramic and single crystal counterparts would crack. Strain engineering has proven to drastically enhance ferroelectric polarizations and transition temperatures in classic ferroelectric materials such as BaTiO 3 [6] and PbTiO 3 [7], and strain can even turn on ferroelectricity in otherwise non-polar materials such as SrTiO 3 and CaTiO 3 ,[8] although reported strain-induced ferroelectric transition temperatures are only at room temperature or below.[9-12] In this work, we will focus on strain-induced ferroelectricity in CaTiO 3 , the first discovered perovskite mineral and long known to be a simple nonpolar insulating crystal.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Graphite Degradation Modeling and Analysis

A graphite component in a nuclear reactor core is subjected to variety of stresses and can experience degradation during normal and off-normal operation. Understanding how a graphite component will behave in service is essential to ensuring core structural stability and safe reactor operation. This report summarizes a graphite modeling tool currently under development at Idaho National Laboratory. The model incorporates several of the anticipated stresses during service and includes the effects of oxidation and irradiation prior to turnaround. This tool is intended to be used to help assess the design of graphite components by utilizing design code rules found in Section III, Division 5 of the American Society of Mechanical Engineering Boiler and Pressure Vessel Code. Specifically, the tool uses the methodologies found within the Full and Simplified assessments from Article HHA-3000 to verify that a graphite component has an acceptably low probability of failure.

36 MATERIALS SCIENCE↗
DOE OSTI2023

System Study: High-Pressure Core Spray 1998-2022

This report presents an unreliability evaluation of the high-pressure core spray (HPCS) at eight U.S. commercial operating boiling water reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from 1998–2022 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10-year period while yearly estimates for system unreliability are provided for the entire active period. Statistically significant increasing trends were identified in both the HPCS system start-only unreliability and 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

System Study: High-Pressure Coolant Injection 1998-2022

This report presents an unreliability evaluation of the high-pressure coolant injection (HPCI) system at 22 U.S. commercial operating boiling water reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from 1998–2022 for selected components were obtained from the Institute of Nuclear Power Operations Industry Reporting and Information System. The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCI results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2024

Validation of the SCALE/Polaris–PARCS Code Procedure With the ENDF/B-VII.1 AMPX 56-Group Library: Boiling Water Reactor

The SCALE/Polaris–PARCS code procedure has been used in the confirmatory analysis for boiling water reactors by the US Nuclear Regulatory Commission. In this study, the SCALE/Polaris v6.3.0–PARCS v3.4.2 code procedure with the Evaluated Nuclear Data File (ENDF)/B-VII.1 AMPX 56-group library was validated by comparing the simulated results with the measured data for operating boiling water reactors, including Peach Bottom Unit 2 cycles 1–3, Hatch Unit 1 cycles 1–3, and Quad Cities Unit 1 cycles 1–3. The uncertainties and biases of the SCALE/Polaris–PARCS code package for boiling water reactor physics analysis were evaluated in the validation for key nuclear parameters such as reactivity and traversing in-core probe data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Fabrication and Testing of Solid-Solution Strengthened Corrosion Resistant Alloys For Service in Molten Fluoride Environments

The demand for higher system thermal efficiencies requires the operation of power generation cycles and heat conversion systems at progressively higher temperatures. As the system operating temperature increases, existing materials may not provide adequate mechanical properties or environmental compatibility or both. There is an increasing commercial interest in the development and deployment of liquid-fueled Molten Salt Reactors (MSRs). Hastelloy®N, the highest performing candidate MSR structural alloy, is not capable of operations at temperatures above 700°C, thus limiting the performance of these systems. Using an Integrated Computational Materials Engineering (ICME)-approach and small laboratory scale heats, ORNL developed a class of patented alloys covered by U.S. Patent 9, 435, 011 B2, “Creep-resistant, Cobalt-free alloys for high temperature, liquid-salt heat exchanger systems,” similar to Hastelloy®N in that they are primarily solid solution strengthened. In contrast to precipitation strengthened alloys, the microstructure of solid solution alloys and hence the high temperature mechanical properties are stable for extended periods of time allowing long reactor operating life. The new alloys have shown to possess good resistance to liquid fluorides at temperatures up to 850°C and have significantly improved creep properties when compared to Hastelloy®N. The purpose of the CRADA project was for ORNL to collaborate with Haynes International- a materials producer, MetalTek International- a foundry, and Kairos Power – an advanced reactor developer – to scale-up selected alloys, evaluate their properties, and identify one solid solution strengthened alloy that can meet the property requirements for the reactor being developed by Kairos Power and other similar liquid fluoride-salt cooled reactors. As part of the project, eight alloys were down-selected and fabricated in larger industrial scale heats by Haynes International. Resistance to molten salt was evaluated in flowing FLiNaK and FLiBe by Kairos Power using their Rotating Cage Loop (RCL) system. Accounting for iron deposition during these tests, the new alloys displayed very low net mass change showing excellent corrosion performance in molten salt. Creep properties evaluated at ORNL were found to be better than that of Hastelloy®N and 316 stainless steel. Long-term stabilities of the alloys evaluated by Haynes International showed that these alloys have excellent thermal stability in the temperature range 704.4-815.6°C, with the change in strength and ductility being less than 10-15% after a 4000-hour exposure at 815.6°C. Autogenously Gas Tungsten Arc Welding (GTAW) welded samples showed less than 10% change in yield strength / ultimate tensile strength / total elongation compared to the basemetal, indicating that the alloys have excellent weldability. Three parts were successfully investment-cast using one alloy with very little voiding showing feasibility of fabricating parts using the casting process. This project enabled extensive interaction between the material producer Haynes International, casting supplier MetalTek, and reactor developer Kairos Power. This facilitated testing of materials and components produced using the newly developed alloys by the end-user. This allowed the generation of critical dataset required for down-selection of a few promising alloys for further development. This data is also currently being shared with other reactor designers for them to evaluate the suitability of this alloy for their reactor design. The availability of this alloy will ultimately enable the design and development and deployment of MSRs with increased temperature of operation and thus, improved efficiencies.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

TREAT Transient Test Reactor Restart Status

Transient testing involves placing nuclear fuel or material into the core of a nuclear reactor designed to operate at high power for a short time, and subjecting the nuclear fuel or material to short bursts of intense, high power radiation. After the experiment is completed, the fuel or material is analyzed to determine the effects of the radiation. The results are then used in fuel or material design and/or qualification. Transient testing is required for essentially all nuclear fuel design and qualification efforts to learn how nuclear fuel will respond during accidents involving transient overpower and/or under cooling events. For example, nuclear fuel may fragment when exposed to higher than normal power. This fragmentation can cause unacceptable performance. Transient testing is also needed to validate performance models for nuclear fuel and materials. These models, when validated, will dramatically shorten the development and qualification life cycle for nuclear fuels, supporting rapid development of low emissions, reliable power generation. In the past, transient testing was primarily done in facilities that no longer exist. The U.S. Department of Energy evaluated how to provide the required transient testing capability and, following completion of the National Environmental Policy Act process, they selected the Transient Reactor Test (TREAT) facility to resume transient testing. TREAT is located at Idaho National Laboratory (INL). The TREAT reactor was used from 1959 to 1994 to conduct more than 2,500 nuclear fuel transient tests, and was placed in standby in 1994. TREAT is capable of transients in excess of 19 GW. The paper describes the progress towards restart. The restart efforts have revealed that the plant fuel and equipment is acceptable for future operations. The safety basis development, procedures, and training are well under way. Experiment capability preparations are in progress. Reactor startup is anticipated no later than 2018. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Improving Predictions Under Uncertainty of Material Plasma Device Operations

Understanding the properties of materials when exposed to various plasma temperatures and fluxes is essential to the building and operating of fusion reactors. The Material Plasma Exposure eXperiment (MPEX) is an instrument currently being developed by the Department of Energy (DOE) for this purpose. MPEX is expected to come online in stages over the next five years. Proto-MPEX, the predecessor to MPEX, operated from 2014 to 2021, and was designed to understand the generation of plasma temperatures and fluxes at orders of magnitude below what will be obtained by MPEX. This work uses the recently developed stochastic neural network (SNN), a machine learning technique capable of operating under uncertainty to provide a surrogate model for the Proto-MPEX device. We demonstrate that SNN outperforms Bayesian neural network (BNN), a standard in the field of machine learning with uncertainty. The development of a robust surrogate of the Proto-MPEX will aid in the commissioning and operation of the MPEX device.

Archibald, Rick↗
DOE OSTI2017

DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program – Joint Research and Development Plan, Rev 6

Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1998

Design, Operation, and Modeling of a Vertical APCVD Reactor for Silicon Carbide Film Growth

An atmospheric pressure chemical vapor deposition (APCVD) reactor utilizing a unique vertical geometry which enables 3C-SiC films to be grown on two, 4-inch diameter Si wafers has been constructed. Contrary to expectations, 3C-SiC films grown in this reactor are thickest at the downstream end of the substrates. To better understand the reason for the thickness distribution on the wafers, an axisymmetric finite-element model of the gas flow in the reactor was constructed. The model uses the ANSYS53 Flowtran package and includes compressible and temperature-dependent fluid properties in laminar or turbulent flow. It does not include reaction chemistry or unsteady flow. The ANSYS53 results predict that the cool, inlet fluid falls through the inlet pipe and the warm, diffuser region like a jet. This jet impinges on top of the susceptor and gets diverted to the reactor side walls, where it flows to the bottom of the reactor, turns, and slowly rises along the face of the susceptor. This may explain why the SiC films are thickest at the downstream side of the wafers, as gas containing fresh reactants first passes over this region. Modeling results are presented for both one atmosphere and one half atmosphere reactor pressure.

DeAnna, Russell G.↗
DOE OSTI2019

System for the highly autonomous operation of a modular liquid-metal reactor with steam cycle

The invention relates to a nuclear plant in which the power of a nuclear reactor is controlled via demand of a connected electric grid. A naturally circulating nuclear reactor coolant loop is linked to a water/steam loop by means of a steam generator. The water/steam loop consists of an electric power generating unit and a water recirculating and steam control system. The generator is coupled to an external power grid. As power requirements of the grid change, a controller linked to the generator and a three way valve divides steam flow between the expansion turbine and a feedwater heater to boost or retard the power output. Altering the steam flow changes the pressure and temperature in the water/steam system and thus the coolant flow rate. The change in coolant flow allows the reactor core to regulate its reactivity to reach a state of equilibrium to the demand for electric power.

Vilim, Richard B.↗
DOE OSTI2019

Regulatory Review of Micro-Reactors: Initial Considerations

The U.S. Nuclear Regulatory Commission (NRC) is working to have an effective and efficient mission readiness for reactors that differ considerably from those currently licensed. Micro-reactors, that is, reactors that have a power of roughly no more than 10 MW, are one class of these advanced reactors. This report is one of the first steps taken to articulate the many technical and regulatory issues that will need to be addressed for NRC to have the ability to review license applications for micro-reactors. Many of the issues center around the fact that a) these reactors may be operated remotely or semi-autonomously and b) it will be difficult to analyze risk from new, unique, technologies. Initial thoughts are given on how probabilistic methods could be used to determine risk and how the current approach for reviewing non-power reactors could be useful for micro-reactors.

42 ENGINEERING↗
DOE OSTI2023

System Study: Isolation Condenser 1998-2022

This report presents an unreliability evaluation of the isolation condenser (ISO) system at three U.S. commercial operating boiling water reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hour, and failure data from calendar year 1998–2022 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Industry Reporting and Information System (IRIS). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the ISO results.

99 GENERAL AND MISCELLANEOUS↗