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 1,189 records · Page 66

Integration of Safety, Security, and Safeguards During Design and Operations: A Technical Assessment and Regulatory Considerations for Advanced Reactor and Advanced Fuel Fabrication Facilities

This report presents the current state of knowledge, technology, methodologies, and tools that could be implemented to realize the robust integration of safety, security, and safeguards (3S) for advanced nuclear reactors (ARs) and advanced nuclear fuel cycle facilities. This report was motivated by the global development of ARs which are expected to play a key role in meeting domestic energy and climate objectives. Domestically, with many ARs in the early design phase, the integration of 3S provides an opportunity to achieve risk reduction while using less resources than traditional light water reactors by leveraging interdependencies and synergies between each domain. In addition, domestic policy considerations encourage the convergence of each 3S domain through facility design and operations. Therefore, there is a need to better understand the interdependencies and integration between 3S across ARs and advanced reactor fuel cycle facilities’ lifecycles including design, construction, and operational phases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

System Study: High-Pressure Coolant Injection 1998-2024

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 to 2024 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 and 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.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

System Study: High-Pressure Core Spray 1998-2024

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 to 2024 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 and yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the HPCS results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Crystal River 3 Cable Materials for Thermal and Gamma Radiation Aging

The Expanded Materials Degradation Assessment Volume 5: Aging of Cables and Cable Systems (EMDA) summarizes the state of knowledge of materials, constructions, operating environments, and aging behavior of low voltage and medium cables in nuclear power plants (NPPs) and identifies potential knowledge gaps with regard to cable operation beyond 60 years. The greatest area of uncertainty relates to how well the accelerated aging used in the original equipment qualification (EQ) processes predicts the performance of cable materials in extended operation. General opinion and utility experience have indicated that actual operating environments of in-plant cables are not as severe, however, as the operating and design basis environments used in the qualification process. Better understanding of the long term aging behavior of cable insulation materials in service conditions and the analysis of actual cable operating environments are the objectives of ongoing research to support subsequent license renewal activities in particular and long term cable aging management in general. A key component of the effort to better understand cable material aging behavior is the availability of representative samples of cables that have been installed in operating light water reactors and have experienced long term service. Unique access to long term service cables, including relatively rich information on cable identity and history, occurred in 2016 through the assistance of the Electric Power Research Institute (EPRI). EPRI facilitated DOE receipt of harvested cables from the decommissioned Crystal River Unit 3 (CR3) pressurized water reactor representing six of the nine most common low voltage cable manufacturers (EPRI 103841R1): Rockbestos, Anaconda Wire and Cable Company (Anaconda), Boston Insulated Wire (BIW), Brand-Rex, Kerite and Okonite. Cable samples received had been installed in the operating plant for durations ranging from 10 years to 36 years. These cables provide the opportunity to assess actual in-plant material aging and compare it to the expectations for service aging implied in original equipment qualification. The received samples are from cables manufactured as early as 1971 and as late as 1998. Of the original manufacturers, BIW, Anaconda and Kerite no longer supply low-voltage cables to the nuclear industry. Okonite, Rockbestos, and Brand-Rex do still supply nuclear-grade low-voltage cables, but most cable insulation formulations have changed over the years. Thus the availability of the CR3 samples representative of cables installed in existing U.S. NPPs also presents the opportunity for additional aging studies on the most relevant insulation and jacketing materials. This report describes the cables received from CR3 through EPRI assistance, some of the specific knowledge gaps that study of these cable materials can be used to address, and experimental plans for addressing those gaps using these materials. Harvested cables from CR3 and other NPPs that have experienced long term service, new old stock cables (manufactured before 2000, but never put in service), and relevant modern nuclear cables and materials from cable manufacturers are enabling research to address identified knowledge gaps and better understand long term aging behavior for cable materials currently installed in NPPs. This research, combined with refined understanding of actual service environments and conditions, will both support subsequent licensing activities and more efficient plant cable aging management.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

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↗

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↗

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↗

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↗

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↗

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↗

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↗

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-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↗

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.↗

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.↗