Search NASA⌕ Search

SEARCH · Search NASA

Results for “REACTOR VESSELS”

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 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Research and Development Roadmaps for Nondestructive Evaluation of Cables, Concrete, Reactor Pressure Vessels, and Piping Fatigue

The purpose of the Materials Aging and Degradation Pathway is to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants and to provide data and methods to assess the performance of systems, structures, and components essential to safe and sustained nuclear power plant operations. The understanding of aging-related phenomena and their impacts on systems, structures, and components is expected to be a significant issue for any nuclear power plant planning for long-term operations (i.e., service beyond the initial license renewal period). Management of those phenomena and their impacts during long-term operations can be better enabled by improved methods and techniques for detection, monitoring, and prediction of systems, structures, and components degradation.

Materials Aging↗
DOE OSTI2024

Concept of Operations for Advanced Reactor Spent Nuclear Fuel Management

This presentation presents a preliminary description of a concept of operations to incorporate advanced-reactor spent nuclear fuel (SNF) into an integrated waste-management system (IWMS). The evaluation includes SNF from four advanced-reactor concepts with the following fuel types: (1) small modular reactors using oxide fuels, (2) tristructural-isotropic (TRISO) fuels, (3) metallic fuels, and (4) fuel salts. To provide context for the proposed concept of operations for advanced reactors, a comparison is made with traditional light-water reactors (LWRs) to identify potential gaps in the IWMS. The technical differences between advanced reactors and LWRs are assessed to determine the feasibility of managing advanced-reactor waste streams using existing operations and technology. This presentation emphasizes fuel types from Advanced Reactor Demonstration Program reactors: Xe 100, which uses TRISO fuel and Natrium, with its metallic fuels while also analyzing management options for molten-salt reactors and advanced light-water reactors (ALWRs). Understanding the storage, transportation, and disposal requirements of SNF is dependent on both the quantity and characteristics of the SNF generated by nuclear reactors. This presentation provides a high level overview, comparing the anticipated concept of operations for different SNFs from advanced reactors. The IWMS includes at-reactor storage, transportation, potential off-site storage, potential treatment, and disposition. To assess the potential effect of advanced-reactor concept of operations on the IWMS, estimates were made for fuel characteristics that contribute to storage, transportation, disposal, and possible treatment of advanced-reactor SNF. These include canister heat load, dose rates, and criticality-control limits, which are important for determining the condition and configuration of the advanced reactor SNF. At-reactor storage of LWR SNF traditionally involves a spent-fuel pool (SFP) before transfer to an independent spent-fuel storage installation. However, some advanced-reactor concepts, particularly those using TRISO and salt fuels, do not anticipate the use of an SFP. This difference in at-reactor storage could impact the IWMS. Additionally, transportation of advanced-reactor SNF may include additional processes tied to potential off gassing, and transportation of microreactor SNF may occur within the reactor vessel. Some advanced-reactor SNF could also undergo treatment to meet requirements of an acceptable waste form for disposition, and the treatment location will be a major contributor to efficiently performing IWMS responsibilities. Moreover, the quantity of SNF generated is an important consideration for IWMS because it could affect the size of the transportation fleet and potential off-site storage requirements. Additionally, volume and heat load are the primary drivers for SNF disposition. This presentation compares potential packaging options for advanced reactor SNF with traditional LWR packaging to provide a high-level comparison for the needs of the IWMS for advanced-reactor SNF.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Coupled Deterministic and Monte Carlo Neutronics for Vessel Fluence Calculations

Throughout core life, the reactor vessel is continually bombarded with neutron flux embrittling the vessel and limiting its useful life. As a result, it is imperative to adequately predict how much flux has passed through the vessel during all stages of its life, as a means of ensuring it has adequate ductility to preclude a brittle fracture even under design accident conditions. At present many of these predictions are made either by purely deterministic or purely stochastic means, each offering their own advantages and disadvantages. This paper introduces a coupled method for calculating fluence through the VERA framework, which enables modeling incremental accumulations of fluence while simultaneously accounting for several factors as they evolve through core life. The intention of this method is to simplify the calculation process and provide better resource utilization in obtaining the required accuracy.

Henderson, Shane↗
DOE OSTI2017

LOCA Analysis for the NIST Research Reactor

The research reactor (aka NBSR) at the National Institute of Standards and Technology (NIST) is a heavy water cooled, moderated, and reflected tank-type research reactor that operates at atmospheric pressure and a design power of 20 MW. Core power is generated in 30 fuel elements, each with 17 fuel plates in an upper section and a separate lower section. Power is removed using forced convection and heat exchangers in the cooling loop. The licensing analysis of the loss-of-coolant accident (LOCA) in the NBSR was addressed in the current Safety Analysis Report (SAR). The SAR discusses the low probability of a significant pipe break: 'the main piping is located in protected areas, system pressures are low, and flow rates are small so that wear is not an issue.' Therefore, the probability of a large break, including a double-ended guillotine break, is extremely low. For smaller breaks where the operator has time to take action, procedures are in place to 1) mitigate a loss of water by tripping pumps and closing control valves to isolate the reactor vessel after the falling water level in the vessel is detected by instrumentation, and 2) ensure that emergency cooling water continues to flow for as long as needed. The SAR also refers to an analysis of why there is sufficient water without forced cooling to cool the fuel elements (FEs) in a LOCA if the flow channels between two fuel plates (or a fuel plate and an outside plate) remain filled with water. That analysis was based on a flooding-limited critical power but assumed that reactor trip had taken place. A more recent study showed that some scenarios may lead to the draining of water from the coolant channels. In these cases, the cooling on the inside of the fuel element is from emergency cooling water that falls only along the inner surface of one of the side plates in the fuel elements; the remaining surfaces within the fuel element being exposed to gas. The analysis therein was done for both the present high-enriched uranium fuel (HEU) and the proposed low-enriched uranium fuel (LEU). The LOCA analyses done previously assumed that there is sufficient cooling during the short-term, i.e., before the water level reaches the top of the fuel plates and begins to drain from the coolant channels or before decay heat becomes low enough such that the flooding-limited critical power shows that there is no problem. The present study is to determine if that assumption is correct, i.e., to consider the situation immediately after the break, before and after reactor trip when the power is still relatively high.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Mid–height seismic isolation of equipment in nuclear power plants

An innovative seismic isolation solution for designers of safety-class equipment in advanced nuclear power plants is introduced. The test specimen was a tall, slender, carbon steel vessel that could represent a reactor vessel, steam generator, or a heat exchanger in a nuclear power plant: 240 inches tall, outer diameter of 60 inches, and wall thickness of 1 inch. The vessel was supported by three radial mounts at its mid-height, near its center of gravity, on a steel frame. The vessel was subjected to three-component ground motions using a 6DOF earthquake simulator. The specimen was filled with water for testing to indirectly account for the fluid and internal equipment present inside a prototype vessel. Three configurations were tested: non-isolated, isolated using single Friction Pendulum (SFP) bearings, and isolated using triple Friction Pendulum (TFP) bearings. The test results demonstrate that mid-height seismic isolation is practical and enables a significant reduction in horizontal spectral accelerations. Furthermore, these outcomes are not specific to the spherical sliding bearings used in the experiments but are broadly applicable to mid-height, seismically isolated equipment.

42 ENGINEERING↗