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 577 records · Page 32
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Weight function procedure for reduced order fracture analysis of arbitrary flaws in cylindrical pressure vessels

Fracture mechanics calculations using the finite element method can be computationally expensive, which makes them challenging to use in engineering evaluations of flaws in pressure vessels. The weight function (WF) technique is a reduced order fracture modeling method that is widely used to greatly reduce these computational costs. Its computational efficiency is essential for use in probabilistic fracture mechanics evaluations of embrittled nuclear reactor pressure vessels (RPVs), due to the large number of sampled flaws that must be evaluated. Although the WF technique is general, it is typically used only for axis-aligned flaws. Recent discoveries of off-axis flaws in operating nuclear reactors have necessitated detailed simulations of such flaws and the interactions between neighboring flaws. This paper presents a generalized WF approach applicable for analyzing arbitrary flaw geometries in thick-walled cylindrical vessels, including surface-breaking and subsurface flaws, which can either axis-aligned or off-axis, and can account for interactions with other flaws. Herein, this approach is demonstrated on representative simulations of multiple flaw geometries in a RPV subjected to transient loading conditions. In all cases, the WF technique gives good comparison with benchmark results from direct simulation with greatly reduced computational effort.

42 ENGINEERING↗
DOE OSTI2024

Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1971

Fission engine concepts.

Discussion of numerous approaches to exploitation of fission energy for special applications in more beneficial ways than can be achieved by conventional reactor technology. It is *hown that, for space propulsion, the high temperature plasma core reactor is most desirable. But fissioning plasmas may find applications beyond this. Because of the criticality to be achieved in high temperature gaseous uranium fuel, plasma core reactors would operate at very high pressures, up to a thousand atmospheres, and their power would range from 5000 to 50,000 megawatts. However, apart from obvious technological difficulties, basic physics problems have to be solved before the feasibility of plasma core reactors can be demonstrated. These problems and their implications are discussed.

Thom, K.↗
NASA NTRS1978

Low cost silicon solar array project: Feasibility of low-cost, high-volume production of silane and pyrolysis of silane to semiconductor-grade silicon

Silicon epitaxy analysis of silane produced in the Process Development Unit operating in a completely integrated mode consuming only hydrogen and metallurgical silicon resulted in film resistivities of up to 120 ohms cm N type. Preliminary kinetic studies of dichlorosilane disproportionation in the liquid phase have shown that 11.59% SiH4 is formed at equilibrium after 12 minutes contact time at 56 C. The fluid-bed reactor was operated continuously for 48 hours with a mixture of one percent silane in helium as the fluidizing gas. A high silane pyrolysis efficiency was obtained without the generation of excessive fines. Gas flow conditions near the base of the reactor were unfavorable for maintaining a bubbling bed with good heat transfer characteristics. Consequently, a porous agglomerate formed in the lower portion of the reactor. Dense coherent plating was obtained on the silicon seed particles which had remained fluidizied throughout the experiment.

Breneman, W. C.↗
DOE OSTI2026

System Study: Auxiliary Feedwater 1998-2024

This report presents an unreliability evaluation of the auxiliary feedwater (AFW) system at 62 U.S. commercial operating nuclear 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 years 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 industry-wide estimates of AFW system start-only unreliability, but a statistically significant decreasing trend was identified in the industry-wide estimates of AFW system 24-hour mission unreliability.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

System Study: Residual Heat Removal 1998-2022

This report presents an unreliability evaluation of the residual heat removal (RHR) system in two modes of operation (low-pressure injection in response to a large loss-of-coolant accident and in response to post-trip shutdown cooling) at 93 U.S. commercial operating nuclear 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 (INPO) Industry Reporting and Information System (IRIS), formerly the INPO Consolidated Events Database. 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 RHR results.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

System Study: Emergency Power System 1998-2022

This report presents an unreliability evaluation of the emergency power system (EPS) at 93 U.S. commercial operating nuclear 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 industry-wide estimates of EPS system start-only unreliability, but a highly statistically significant decreasing trend was identified in the industry-wide estimates of EPS system 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
NASA NTRS2007

Oxygen production System Models for Lunar ISRU

In-Situ Resource Utilization (ISRU) seeks to make human space exploration feasible; by using available resources from a planet or the moon to produce consumables, parts, and structures that otherwise would be brought from Earth. Producing these in situ reduces the mass of such that must be launched and doing so allows more payload mass' for each mission. The production of oxygen from lunar regolith, for life support and propellant, is one of the tasks being studied under ISRU. NASA is currently funding three processes that have shown technical merit for the production of oxygen from regolith: Molten Salt Electrolysis, Hydrogen Reduction of Ilmenite, and Carbothermal Reduction. The ISRU program is currently developing system models of, the , abovementioned processes to: (1) help NASA in the evaluation process to select the most cost-effective and efficient process for further prototype development, (2) identify key parameters, (3) optimize the oxygen production process, (4) provide estimates on energy and power requirements, mass and volume.of the system, oxygen production rate, mass of regolith required, mass of consumables, and other important parameters, and (5) integrate into the overall end-to-end ISRU system model, which could be integrated with mission architecture models. The oxygen production system model is divided into modules that represent unit operations (e.g., reactor, water electrolyzer, heat exchanger). Each module is modeled theoretically using Excel and Visual Basic for Applications (VBA), and will be validated using experimental data from on-going laboratory work. This modularity (plug-n-play) feature of each unit operation allows the use of the same model on different oxygen production systems simulations resulting in comparable results. In this presentation, preliminary results for mass, power, volume will be presented along with brief description of the oxygen production system model.

Santiago-Maldonado, Edgardo↗
NASA NTRS2019

Overview of the Main Propulsion System for a Nuclear Thermal Propulsion Flight Demonstrator

A demonstration of a Nuclear Thermal Propulsion (NTP) engine has not been conducted in over 50 years. Several tests were conducted during the NERVA program but no NTP engine was ever flown in space. In the last several years there has been a considerable amount of conceptual design work on NTP engines conducted. With the prospect of human Mars missions in the 2030’s there has been a renewed interest in NTP engines. A concept design study was conducted with the intent to design 2 flight demonstrator vehicles that would buy down programmatic and technical risks associated with launching and operating nuclear reactors in space. The intent of the first demonstrator mission would be to employ a simplified NTP engine and buy down programmatic risks whereas the second demonstrator would buy down technical risks with a NTP engine designed to be similar to an operational NTP model. The results of the study showed that a simplified NTP engine demonstrator could be feasibly built and flown in the near term with mostly high TRL, commercial off-the-shelf components.

Simpson, S. P.↗
DOE OSTI2023

System Study: High-Pressure Safety Injection 1998-2022

This report presents an unreliability evaluation of the high-pressure safety injection system (HPSI) at 62 U.S. commercial operating nuclear reactors. New Standardized Plant Analysis Risk (SPAR) models with the most recent SPAR parameter update results were used in this report. Demand, run hours, and failure data from calendar years 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. Highly statistically significant increasing trends were identified in both the HPSI system start-only unreliability and 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

System Study: Auxiliary Feedwater 1998-2022

This report presents an unreliability evaluation of the auxiliary feedwater (AFW) system at 62 U.S. commercial operating nuclear 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 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 industry-wide estimates of AFW system start-only unreliability, but a highly statistically significant decreasing trend was identified in the industry-wide estimates of AFW system 8-hour mission unreliability.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2021

Solid waste gasification: comparison of single- and multi-staged reactors

Interest in converting waste into renewable energy has increased recently due to concerns about sustainability and climate change. This solid waste is mainly derived from municipal solid waste (MSW), biomass residue, plastic waste, and their mixtures. Gasification is one commonly applied technology that can convert solid waste into usable gases, including H2, CO, CH4, and CO2. Single- and multi-staged reactors have been utilized for solid waste gasification. Comparison in reactor dimensions, operating factors (e.g., gasification agent, temperature, and feed composition), performance (e.g., syngas yield and selectivity), advantages, and disadvantages are discussed and summarized. Additionally, discussion will include economic and advanced catalysts which have been developed for use in solid waste gasification. The multi-staged reactor can not only be applied for gasification, but also for pyrolysis and torrefaction.

Zhao, Xianhui↗
DOE OSTI2025

INL_intern_poster_RashedA

Advanced nuclear reactors often operate at elevated temperatures, novel fuel with different chemical and thermal properties that may interact differently with cladding, new coolants (e.g., molten salt, liquid metal, or sodium) that are chemically aggressive, requires cladding materials with superior oxidation, corrosion, and creep resistance. Developing fuel cladding for advanced reactors is essential, as the cladding plays a critical role in reactor safety, performance, and longevity. In this study, innovative Non-Concentrated Alloys (NCAs) were developed as novel fuel cladding materials for advanced nuclear reactors. Multiple elements were chosen for the alloys. NCAs were manufactured and tested mechanically, showing promise with the arc-melting method, and setting a clear path for the next step of NCAs' development as a fuel cladding for advanced nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2023

Volume 9: Critical Facility with Add-On Ion Beam

Completion of the originally planned critical facility has been considered as part of the High Flux Isotope Reactor (HFIR) Sustaining and Enhancing Neutron Science Initiative at Oak Ridge National Laboratory (ORNL). A working group of ORNL staff members was formed to develop the idea and to recommend one or more configurations to best support future HFIR operations and scientific capacities. HFIR was designed with a critical pool in the reactor bay that was projected to be outfitted as a critical facility. The primary purpose of the planned critical facility was to measure the subcritical worth of fresh fuel elements to support startup requirements. However, the Y-12 National Security Complex already housed a critical facility that was used for this purpose. This report presents the working group’s efforts, including determination of the proposed critical facility’s high-impact benefits. A generic low-power critical facility would be employed for reactor physics measurements, code and data validation, reactor operator and staff training, and education. This facility would be instrumental in supporting current HFIR operations, conversion of HFIR to low-enriched uranium, existing light water reactor operations, and advanced reactor development and deployment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Prospects and Limitations of Predicting Fuel Ignition Properties from Low-Temperature Speciation Data

Using chemical kinetic modeling and statistical analysis, we investigate the possibility of correlating key chemical “markers”–typically small molecules–formed during very lean (φ ~ 0.001) oxidation experiments with near-stoichiometric (φ ~ 1) fuel ignition properties. One goal of this work is to evaluate the feasibility of designing a fuel-screening platform, based on small laboratory reactors that operate at low temperatures and use minimal fuel volume. Buras et al. [Combust. Flame2020,216, 472–484] have shown that convolutional neural net (CNN) fitting can be used to correlate first-stage ignition delay times (IDTs) with OH/HO2 measurements during very lean oxidation in low-T flow reactors with better than factor-of-2 accuracy. In this work, we test the limits of applying this correlation-based approach to predict the low-temperature heat release (LTHR) and total IDT, including the sensitivity of total IDT to the equivalence ratio, φ. We demonstrate that first-stage IDT can be reliably correlated with very lean oxidation measurements using compressed sensing (CS), which is simpler to implement than CNN fitting. LTHR can also be predicted via CS analysis, although the correlation quality is somewhat lower than for first-stage IDT. In contrast, the accuracy of total IDT prediction at φ = 1 is significantly lower (within a factor of 4 or worse). Furthermore, these results can be rationalized by the fact that the first-stage IDT and LTHR are primarily determined by low-temperature chemistry, whereas total IDT depends on low-, intermediate-, and high-temperature chemistry. Oxidation reactions are most important at low temperatures, and therefore, measurements of universal molecular markers of oxidation do not capture the full chemical complexity required to accurately predict the total IDT even at a single equivalence ratio. As a result, we find that φ-sensitivity of ignition delay cannot be predicted at all using solely correlation with lean low-T chemical speciation measurements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Impact of Heaters on Molten Salt Reactors Dynamics

Recently, there has been renewed interest in the Molten Salt Reactor (MSR) concept. This interest is mainly due to its advantages over Light Water Reactors (LWRs). Among these advantages are the flexibility in the fuel choice and the possibility to burn actinides [1]. Currently, two molten research reactors are under development. A 2 MW molten salt experimental reactor has been constructed and is planned for operation in China. A 1 MW Molten Salt Research Reactor is under construction in the United States [2]. Thus, Research and Development (R&D) programs are needed for the MSR technology. Reactor dynamics studies are crucial safety evaluation studies. The development of reactor dynamics tools for MSRs started during the Molten Salt Reactor Program (MSRP) at ORNL, focusing on the Molten Salt Reactor Experiment (MSRE) [3]. Although various dynamic simulation tools are available, R&D efforts to develop specialized tools for MSRs are still ongoing [4]. The primary motivation for this interest is the goal of commercializing MSRs. Developing a dynamics tool for MSRs is crucial for analyzing their safety and supporting their demonstration efforts. A crucial safety issue in MSRs is ensuring that the fuel salt in the primary loop remains molten. This concern becomes particularly important when the reactor operates at lower power levels. Most chloride and fluoride salts used as fuel have a melting point of approximately 450 °C [5]. Therefore, maintaining fuel salt temperatures above this melting point is essential to prevent it from freezing. Freezing of the fuel salt can lead to volume expansion, potentially causing damage to reactor components [6]. Moreover, it can create local blockages within fuel salt channels, reducing cooling efficiency and resulting in hot spots [7]. Various methods can prevent fuel salt from freezing at low power levels. These include reducing the fuel mass flow rate in the primary loop, decreasing heat exchange between the primary and secondary loops in the reactor, and using electric fuel salt heaters. Reducing the mass flow rate of fuel salt in the primary loop can increase the fuel salt transit time in the core, enhancing heat production and increasing temperature. However, it can potentially lead to challenges. These include increasing the potential for fuel salt deposition on the reactor channel walls [8], affecting the overall heat transfer processes. Thus, alternative methods should be used [9]. This research investigates the effects of fuel salt heaters on the dynamics and stability during MSR operation. The investigation includes studying the reactor stability with the fuel salt heaters on and off. In addition, the reactor response to transients is compared with and without the operation of fuel salt heaters. These transients include reactivity insertion, primary and secondary pump failure, and heat sink temperature increase. The results of the reactivity insertion transient are presented in the summary.

73 - NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS1998

High-Density Plasma Reactors: Simulations for Design

The development of improved and more efficient plasma reactors is a costly process for the semiconductor industry. Until five years ago, the Industry made most of its advancements through a trial and error approach. More recently, the role of computational modeling in the design process has increased. Both conventional computational fluid dynamics (CFD) techniques like Navier-Stokes solvers as well as particle simulation methods are used to model plasma reactor flowfields. However, since high-density plasma reactors generally operate at low gas pressures on the order of 1 to 10 mTorr, a particle simulation may be necessary because of the failure of CFD techniques to model rarefaction effects. The direct simulation Monte Carlo method is the most widely accepted and employed particle simulation tool and has previously been used to investigate plasma reactor flowfields. A plasma DSMC code is currently under development at NASA Ames Research Center with its foundation as the object-oriented parallel Cornell DSMC code, MONACO. The present investigation is a follow up of a neutral flow investigation of the effects of process parameters as well as reactor design on etch rate and etch rate uniformity. The previous work concentrated on silicon etch of a chlorine flow in a configuration typical of electron cyclotron resonance (ECR) or helical resonator type reactors. The effects of the plasma on the dissociation chemistry were modeled by making assumptions about the electron temperature and number density. The electrons or ions themselves were not simulated.The present work extends these results by simulating the charged species.The electromagnetic fields are calculated such that power deposition is modeled self-consistently. Electron impact reactions are modeled along with mechanisms for charge exchange. An bipolar diffusion assumption is made whereby electrons remain tied to the ions. However, the velocities of tile electrons are allowed to be modified during collisions and are not confined to a Maxwellian distribution. The interaction between the neutral flow and plasma is examined, and results for etch rate uniformity from the previous research and the present plasma simulations are compared.

Hash, David B.↗
DOE OSTI2025

Structural Design and Modeling of MARVEL Primary Coolant System Using the ASME Section III, Division 5, Code

This paper presents the structural design and supporting analysis for the Microreactor Applications Research Validation and Evaluation (MARVEL) primary coolant system (PCS) using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section III, Division 5, rules. MARVEL is a liquid metal–cooled microreactor intended to provide experimental capabilities for the rapid testing and development of microreactor technologies. The PCS utilizes high-temperature sodium-potassium liquid metal as the primary coolant and operates at a design temperature of 570°C, necessitating the consideration of creep-related failure mechanisms. The base metal for the PCS is 316H stainless steel, and the weldments are made with a 16-8-2 filler. The design approach incorporates the current base code rules along with ASME code cases N-924, N-861, and N-862 to address primary load, ratcheting, and creep-fatigue evaluations, respectively. The reactor’s operation involves complex thermal and mechanical interactions due to natural convective flow and differential thermal expansion between components. In conclusion, this paper discusses the structural engineering challenges encountered, such as managing thermal stresses in the distribution plenum and guard vessel, and outlines the strategies implemented to meet the code requirements, including design modifications and operational constraints.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗