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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Status and prospects for inertial fusion energy via lasers

Fusion energy is the ultimate clean and limitless energy source, but its development requires overcoming many scientific and technological challenges. For the first time in the 60-year long history of fusion research, ignition and target gain above unity (G>1) were demonstrated in the laboratory at the National Ignition Facility at the Lawrence Livermore National Laboratory in 2022. Turning laser fusion into an energy source requires that the results from single shot experiments be replicated at much higher repetition rates of many shots per second to produce high average power at relatively low cost. Furthermore, this requires the development of new laser technologies, mass production of suitable targets, accurate injection and tracking systems and new materials for the reactor chamber components and final optics. Ultra broadband and deep UV light are laser advances that can dramatically improve the laser energy coupling to the target thereby reducing the laser energy and power requirements.

Fusion energy↗
DOE OSTI2024

MARVEL Decommissioning Plan

The purpose of this document is to update and formally document decommissioning plans for the Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor after two years of operational testing. The operational testing consists of 3 continuous days of operation followed by 4 days of non-operation. This plan is also based on the assumption that all MARVEL fuel will be transported to long-term intermediate storage at the CPP-603 used fuel storage facility, sodium-potassium (NaK) coolant can be disposed as Class B radioactive waste and that the balance of the reactor primary components can be disposed of at an available disposal facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Summary of LWRS Research in Addressing RPV Research Gaps in NRC EMDA Report

Reactor Pressure Vessels (RPVs) are critical components in nuclear reactors, housing the reactor core and coolant under extreme conditions of temperature, pressure, and radiation. These harsh environments contribute to the degradation of RPV materials over time, presenting challenges for extending reactor operations beyond their original design lifespans. The NRC Expanded Materials Degradation Assessment (EMDA) report volume 3 have been instrumental in guiding research to support extending the operational life of light water reactors (LWRs) up to 80 years. It provides a comprehensive framework to address technical challenges related to aging and degradation mechanisms in RPVs. The LWRS program has played a key role in advancing this research, supporting projects such as the UCSB ATR-2 Experiment, material testing from Zion and Palisades reactors, and the development of advanced mini-compact tension testing techniques. These efforts have been crucial in identifying and addressing gaps in our understanding of RPV aging, contributing to the successful subsequent license renewals of eight LWR units in the U.S. The EMDA report volume 3, built on the Phenomena Identification and Ranking Table (PIRT) analysis from earlier versions of the EPRI Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs), provides a detailed assessment of RPV degradation mechanisms. However, as EPRI has updated the MDM and IMT, it is important to revisit research priorities and methodologies to reflect these changes. The revised MDM and IMT may introduce new factors affecting long-term RPV performance and safety, highlighting the need for continued research and updated guidance to ensure the reliable and safe operation of reactors beyond 80 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Survey of baffle plate samples, specimen machining, and hydrogen measurements

This report outlines preliminary results on the survey and characterization of irradiated baffle plate, baffle former, and flux thimble tube specimens as part of the ongoing investigation into irradiation-induced embrittlement in austenitic stainless steels at ambient (room temperature) conditions. The specimens originated from commercial pressurized water reactor (PWR) components, covering displacement damage doses ranging from approximately 0.065 to 75 dpa. Initial scanning electron microscopy (SEM) surveys revealed that specimen surfaces exhibited fine machining marks and in-service-formed oxide layers on the side surfaces of the analyzed specimens. The oxide layers revealed specific features resembling localized "pitting-like" corrosion. An additional set of specimens representing in-service oxidation is being preserved for future microstructure analysis work. As believed, this will provide additional insights into long-term material degradation in PRWs. Specimen machining challenges emerged due to a complex failure of the electric discharge machining (EDM) system located in hot area. Despite partial restoration, issues persist with the EDM’s secondary power supply, necessitating the exploration of alternative EDM or computer numerical control (CNC) machining approaches to facilitate tensile specimen preparation. Currently, low speed saw cutting is in progress to prepare specimens for hydrogen measurements and general microstructure analysis. The near-term goals include completing tensile specimen machining for advanced mechanical testing and characterizing fracture mechanisms, and stress-corrosion cracking testing, ultimately aiming to identify and mitigate the ambient-condition intergranular cracking through targeted post-irradiation annealing strategies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Implications of Microstructure in Helium-Implanted Nanocrystalline Metals

Helium bubbles are known to form in nuclear reactor structural components when displacement damage occurs in conjunction with helium exposure and/or transmutation. If left unchecked, bubble production can cause swelling, blistering, and embrittlement, all of which substantially degrade materials and—moreover—diminish mechanical properties. On the mission to produce more robust materials, nanocrystalline (NC) metals show great potential and are postulated to exhibit superior radiation resistance due to their high defect and particle sink densities; however, much is still unknown about the mechanisms of defect evolution in these systems under extreme conditions. Here, the performances of NC nickel (Ni) and iron (Fe) are investigated under helium bombardment via transmission electron microscopy (TEM). Bubble density statistics are measured as a function of grain size in specimens implanted under similar conditions. While the overall trends revealed an increase in bubble density up to saturation in both samples, bubble density in Fe was over 300% greater than in Ni. To interrogate the kinetics of helium diffusion and trapping, a rate theory model is developed that substantiates that helium is more readily captured within grains in helium-vacancy complexes in NC Fe, whereas helium is more prone to traversing the grain matrices and migrating to GBs in NC Ni. Our results suggest that (1) grain boundaries can affect bubble swelling in grain matrices significantly and can have a dominant effect over crystal structure, and (2) an NC-Ni-based material can yield superior resistance to irradiation-induced bubble growth compared to an NC-Fe-based material and exhibits high potential for use in extreme environments where swelling due to He bubble formation is of significant concern.

36 MATERIALS SCIENCE↗
NASA NTRS1975

Nonlinear seismic analysis of a reactor structure impact between core components

The seismic analysis of the FFTF-PIOTA (Fast Flux Test Facility-Postirradiation Open Test Assembly), subjected to a horizontal DBE (Design Base Earthquake) is presented. The PIOTA is the first in a set of open test assemblies to be designed for the FFTF. Employing the direct method of transient analysis, the governing differential equations describing the motion of the system are set up directly and are implicitly integrated numerically in time. A simple lumped-nass beam model of the FFTF which includes small clearances between core components is used as a "driver" for a fine mesh model of the PIOTA. The nonlinear forces due to the impact of the core components and their effect on the PIOTA are computed.

Hill, R. G.↗
NASA NTRS2006

Stainless Steel NaK Circuit Integration and Fill Submission

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed to hold a eutectic mixture of sodium potassium (NaK), was redesigned to hold lithium; but due to a shift in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature loop include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This document summarizes the integration and fill of the pumped liquid metal NaK flow circuit.

Garber, Anne E.↗
NASA NTRS2006

Design, Fabrication and Integration of a NaK-Cooled Circuit

The Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system, which was originally designed for use with a eutectic mixture of sodium potassium (NaK), was redesigned to for use with lithium. Due to a shi$ in focus, it is once again being prepared for use with NaK. Changes made to the actively pumped, high temperature circuit include the replacement of the expansion reservoir, addition of remotely operated valves, and modification of the support table. Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a fill design) was selected for fabrication and test. This paper summarizes the integration and preparations for the fill of the pumped liquid metal NaK flow circuit.

Garber, Anne↗
NASA NTRS2007

Assembly and Thermal Hydraulic Test of a Stainless Steel Sodium-Potassium Circuit

Early Flight Fission Test Facilities (EFF-TF) team has been tasked by the NASA Marshall Space Flight Center Nuclear Systems Office to design, fabricate, and test an actively pumped alkali metal flow circuit. The system was originally built for use with lithium, but due to a shift in focus, it was redesigned for use with a eutectic mixture of sodium potassium (NaK). Basic circuit components include: reactor segment, NaK to gas heat exchanger, electromagnetic (EM) liquid metal pump, load/drain reservoir, expansion reservoir, instrumentation, and a spill reservoir. A 37-pin partial-array core (pin and flow path dimensions are the same as those in a full design) was selected for fabrication and test. This paper summarizes the first fill and checkout testing of the Stainless Steel NaK-Cooled Circuit (SNaKC).

Garber, A.↗
NASA NTRSDate not supplied

Hot Hydrogen Testing of W-coated dUN Kernels for Nuclear Thermal Propulsion

Tungsten-coated kernels of uranium nitride (UN) are a possible reactor fuel component for nuclear thermal propulsion. Hot hydrogen testing of such kernels (approx. 250μm diameter, coating thickness 5μm) has been performed at temperatures from 1800°C to 2300°C for 30min each, using depleted uranium nitride (dUN). In addition to direct observations and mass loss measurements, the samples were analyzed by XRD and SEM/EDS after each run. Decomposition of dUN and the formation of molten uranium happened at all temperatures tested with reaction rates increasing with temperature, despite the tungsten coating. At 2300°C, the kernels disintegrated and molten uranium seeped through the wall and bottom of the tungsten crucible employed. These hot hydrogen tests allowed to test the integrity of the tungsten coating and addressed the operating limits of the loose tungsten-coated dUN kernels. Improved results are expected upon consolidation of the tungsten-coated dUN kernels in a Mo-W alloy matrix.

nuclear fuels↗
DOE OSTI2023

NOTCH EFFECT ON CREEP-FATIGUE BEHAVIOR OF ALLOY 617 AT ELEVATED TEMPERATURE

High-temperature reactor structural components are often under the complex multiaxial creep-fatigue (CF) loading conditions throughout the lifetime because of geometric and/or metallurgical discontinuities and complex loading paths. To assess the multiaxial CF deformation behavior and to evaluate the CF design rules in the ASME BPVC Section III, Division 5, Subsection HB, Subpart B, experimental and numerical studies are performed on Alloy 617 at 950°C using notch specimen geometries under CF loading in this study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Nuclear reactor core support system providing radial and axial support

A nuclear reactor core mechanical support bracket is disclosed. The support bracket includes a housing, a spring disposed internally within the housing, a shaft slidingly disposed within the housing, a shaft travel pin, and a flange. The shaft is configured to engage the spring to compress and decompress the spring as the shaft travels in and out of the housing. The shaft travel pin controls the travel of the shaft. The flange is configured to mount the nuclear reactor core mechanical support bracket to a canister of a nuclear reactor. The shaft includes an inset configured to interface with a nuclear reactor core component.

Aleshin, Yuriy↗
DOE OSTI2023

Micro-reactor core mechanical support

A nuclear reactor core mechanical support bracket is disclosed. The support bracket includes a housing, a spring disposed internally within the housing, a shaft slidingly disposed within the housing, a shaft travel pin, and a flange. The shaft is configured to engage the spring to compress and decompress the spring as the shaft travels in and out of the housing. The shaft travel pin controls the travel of the shaft. The flange is configured to mount the nuclear reactor core mechanical support bracket to a canister of a nuclear reactor. The shaft includes an inset configured to interface with a nuclear reactor core component.

Aleshin, Yuriy↗
NASA NTRS1983

Constitutive Equations for Use in Design Analyses of Long-life Elevated Temperature Components

Design analysis needs and procedures relative to elevated temperature components in liquid metal fast breeder reactor (LMFBR) system were examined. The effects of the thermal transients on the pressure boundary components are enhanced by the excellent heat transfer properties of the liquid sodium coolant. Design criteria for high temperature nuclear reactor components recognize the potential occurrence of inelastic structural response. Specifically, criteria and limits were developed which reflect a recognition of this potential and employ design by analysis concepts that requires that inelastic (elastic-plastic and creep) analyses be performed. Constitutive equations to represent multiaxial time-dependent responses of LMFBR alloys are established. The development of equations applicable under cyclic loading conditions are outlined.

Pugh, C. E.↗
DOE OSTI2024

Contactless Position Measurement System for Remote Alignment of Highly Reflective Objects

The Contactless Position Measurement System (CPMS) permits automatic alignment of highly-reflecting components that cannot or should not be touched by human hands, such as those used in SRF cavities. CPMS also has application during maintenance of fusion reactors, when the components in the reactor are radioactive. The length of time for which humans can handle the components in order to carry out maintenance or repairs is limited. We would like to measure the relative positions of such objects so we can bring them into alignment with motorized stages, thus eliminating the need for human contact. The CPMS is a computer vision system that measures the location of highly reflective components to within a fraction of a millimeter by finding the edges of silhouette images. This is done by placing a uniform, infrared backlight behind the components, and viewing each component with two low-aberration, infrared cameras. Each of these stereoscopic cameras are located within a coordinate system we set up with reference platforms, light sources, and survey cameras distributed around the perimeter of the string assembly room, or fusion reactor. We obtain stereoscopic silhouette images of each component, and we use these to determine the position of the component within our string assembly coordinate system. The CPMS combines the silhouette images with knowledge of the dimensions of the objects to obtain the relative positions of components in its field of view. These position measurements then allow us to mechanically maneuver the components into contact. Once we know where they are, we can move the components to where they are supposed to be with motorized stages or other mechanical means, check they are in the right place, and bolt them together. In Phase I, we built a prototype infrared backlights and cameras, tested it, wrote an analysis program to fit the images. At a range of 50 cm, we are able to measure the relative positions of two stainless steel flanges with an accuracy of 150 μm rms. The next steps are to test the analysis program with motorized stages, build a full-size prototype, and implement it in an cleanroom with SRF cavity assembly. This can be done in a cost effective way given the materials with which the system is designed. As Open Source Instruments achieved more in Phase I than anticipated, despite not receiving a Phase II SBIR grant, we hope to move forward with product development and sales with a testing partner.

43 PARTICLE ACCELERATORS↗
DOE OSTI2022

Designing Cladded Components for High Temperature Nuclear Service: Part II—Design Rules

The challenge of using existing ASME Section III, Division 5, class A metallic materials for the construction of structural components of advanced reactors with corrosive coolants could be mitigated by allowing designers to use cladding to protect the base material from corrosion. However, the existing Section III, Division 5 rules provide no guidance on the evaluation of strain accumulation and creep-fatigue damage in cladded components. The availability of design rules for cladded components that do not require long-term clad material testing could promote the application of the cladding approach to accelerate the deployment schedule of these advanced reactor systems. To avoid long-term properties for the clad materials Part I of this work proposes two approximate design analysis methods for two types of clad materials—soft clads that creep much faster and have lower yield stress than the class A base material, and hard clads that creep much slower and have higher yield stress than the class A base material. The proposed analysis methods approximate the response of a soft clad by treating it as perfectly compliant and of a hard clad by treating it as linear elastic. Based on these approximate design analysis strategies this Part II develops a complete set of design rules for class A components cladded with either soft or hard clad materials. In conclusion, Part II discusses the reasoning behind the proposed design rules and uses example finite element analyses of representative reactor components to illustrate the use of these design methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Demonstration of CFD to support the coupled analysis of a reactor pressure vessel subjected to pressurized thermal shock

The structural components that comprise nuclear reactors and their supporting structures are subjected to harsh operating environments that can challenge their integrity, especially after exposure for extended duration or under accident condition. As one of the most significant components of a reactor, the Reactor Pressure Vessel (RPV) is exposed to an aggressive environment during the operation time (e.g. more than 40 years). Aging degradation mechanisms (e.g. thermo-fatigue) could grow initial defects up to a critical size, increasing the susceptibility to failure in the RPV. The conventional methods are mostly based on simple crack and structure geometries. Very limited studies consider the real conditions of the RPV subjected to a thermal shock due to a Loss of Coolant Accident (LOCA). During a LOCA event, the most severe conditions take place when the emergency core cooling (ECC) water is injected inside the cold legs filled initially with hotter water and/or steam. The rapid cooling of the down-comer and the internal RPV surface followed probably by re-pressurization of the RPV causes large temperature gradients and variation of pressure which induces thermal-mechanical stresses. In order to develop the model for integrity assessment of a reactor pressure vessel (RPV) subjected to pressurized thermal shock (PTS), a multi-physics simulation, which includes the thermo-hydraulic, thermo-mechanical and fracture mechanics analyses is necessary. The prediction of the temperature field is achieved by using computational fluid dynamics (CFD) simulation. In this report, a demonstration CFD standalone simulation is performed to support coupled analysis for Reactor Pressure Vessel (RPV) subjected to Pressurized Thermal Shock (PTS). The study use a simplified computational domain to represents a real RPV. The purpose of the study is to demonstrate the transient temperature response of RPV to ECC injection. The CFD model is built in a robust and efficient way for further coupled calculation. The next steps of this work, including the coupled thermal and tensor mechanics capabilities using Cardinal are expected to be complete by the end of FY21 for the demo problem. After this, into FY22, the capability will be demonstrated for a realistic RPV.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗