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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Floating Offshore Wind US Manufacturing and Commercialization: Cooperative Research and Development (Final Report)

NREL assessed the supply chain and workforce considerations for the OCG-Wind floater technology, a floating semi-submersible offshore wind substructure, as well sharing vessel needs to inform their installation strategy. This technical assistance was in support of the FLoating Offshore Wind ReadINess (FLOWIN) Prize Phase 2 submission. NREL provided an assessment of domestic supplier capabilities for the main components of their floating offshore wind platform design and analyzed US regional and national supply chain constraints and gaps. Thirteen interviews with companies including steel distributors, forges, foundries, ports, large component fabricators, subcomponent fabricators, and secondary suppliers provided key insights such as 1) assembly ports are the key infrastructure barrier standing in the way of unlocking the domestic assembly and component fabrication for steel-based FOW platforms, 2) domestic steel producers can supply the types and quantities of steel necessary for FOW platforms, and 3) coordination between stakeholders will be a vital part of successfully developing the supply chain and infrastructure needed to domestically produce FOW platforms. In the workforce assessment, NREL documented a step-by-step approach to conduct a place-based assessment of the foundational workforce consideration for recruiting, upskilling, and retaining a workforce, such as supportive local and state policy, nearby education and training programs, and existing relevant industry. This approach was applied to Tacoma, Washington. Tacoma was indicated to have the potential be a successful location for fabrication and assembly of floating offshore wind energy in terms of workforce development. To share data on vessel requirements to install the OCG-Wind floater, NREL compiled resources that help answer the questions related to anchor handling tug vessels, shared a database of cable laying vessels, and answered questions on complying with the Jones Act.

17 WIND ENERGY↗
DOE OSTI2020

Passive integral isolation valve

A valve assembly includes a flange connected to a vessel penetration of a reactor pressure vessel of a nuclear reactor. A valve is disposed inside the flange or protrudes from the flange into the vessel penetration. The valve includes a valve seat and a movable valve member positioned so that pressure inside the reactor pressure vessel urges the movable valve member against the valve seat to close the valve. The valve assembly further includes a plenum having an inlet via which the plenum can be pressurized to apply pressure to the movable valve member that urges the movable valve member away from the valve seat to open the valve. The plenum may be defined in part by a surface of the movable valve member. The valve assembly preferably does not include a valve actuator.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Steam generator with inclined tube sheet

A steam generation system may include a plurality of heat transfer tubes configured to circulate a secondary coolant of the steam generation system. The steam generation system may be thermally coupled to a reactor vessel, and the reactor vessel may be configured to house a primary coolant. Heat generated from within the reactor vessel may be transferred from the primary coolant to the secondary coolant. The steam generation system may further include an inclined tube sheet fluidly coupled to the plurality of heat transfer tubes. The inclined tube sheet may be attached to a wall of the reactor vessel in a non-horizontal orientation.

42 ENGINEERING↗
DOE OSTI2020

Methods and systems for producing a metal chloride or the like

Systems and methods for producing metal chloride M ICl x from metal M I without the use of HCl and/or Cl 2 gases, including: a bath vessel holding conductive fluid; an anode disposed in the conductive fluid, the anode including metal M.sup.I; a cathode assembly disposed in the conductive fluid, the cathode assembly including a cathode vessel including porous and non-porous portions, the non-porous portion holding sacrificial metal chloride M IICl y substantially separate from metal chloride M ICl x , wherein the cathode assembly includes a center lead disposed within the cathode vessel operable for delivering charge to sacrificial metal chloride M IICl y ; and a power supply coupling the anode and the cathode assembly, the power supply polarized to produce current flow in a direction that causes anodic dissolution of metal M.sup.I into the conductive fluid and deposition of metal M.sup.II within the cathode vessel. The systems and methods apply equally to producing metal halide M IX x .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Method of separating components of a gas

The disclosure provides a method for separating components of a gas. A feed gas stream is cooled in a first vessel. The feed gas stream includes methane, water, carbon dioxide, and Natural Gas Liquids. The feed gas stream is cooled in a first vessel. A portion of the water condenses to form a primary liquid stream, resulting in a first depleted gas stream, which is cooled in a second vessel. A portion of the NGLs condense to form a secondary liquid stream, resulting in a second depleted gas stream, which is cooled in a condensing exchanger. A first portion of the methane condenses to form a liquid methane stream, resulting in a third depleted gas stream, which is cooled in a third vessel. A portion of the carbon dioxide condenses, desublimates, or condenses and desublimates as a final product stream, also resulting in a fourth depleted gas stream.

Baxter, Larry↗
DOE OSTI2021

Electrochemically modulated molten salt reactor

An electrochemically modulated molten salt reactor (EMMSR) that contains a vessel and a power source. The vessel houses a fuel salt, at least a portion of a neutron moderator, and at least a portion of an insulator. The fuel salt includes enough dissolved fissile isotopes to cause continued self-sustaining fission reactions during the operation of the EMMSR. The neutron moderator is configured to slow down fast neutrons produced by the dissolved fissile isotopes. The insulator is configured to electrically isolate the neutron moderator from the vessel. The power source has a positive potential and a negative potential. The positive potential is received by the neutron moderator and the negative potential is received by the vessel.

Jiang, Junhua↗
DOE OSTI2021

Method for solid-liquid separations

Devices, systems, and methods are disclosed to separate solids from liquids. A vessel includes a screw, an inlet, a first outlet, and a second outlet. The inlet receives a first slurry having an incoming solids concentration. The first slurry consists of solids and a contact liquid. The first outlet is disposed toward the first end of the vessel. The second outlet is disposed toward a second end of the vessel. The screw preferentially conveys the solids over the contact liquid towards the second end of the vessel, the solids displacing at least a portion of the contact liquid, causing that portion of the contact liquid to flow toward the first outlet. A melting device is included.

Baxter, Larry↗
DOE OSTI2022

TCR Central Shutdown Rod Fine Motion Control

Transformational Challenge Reactor (TCR) is a Helium cooled 3 MWt test reactor that leverages advances in materials and manufacturing, computing, and AI in its design. Classical design of a shutdown rod uses either gravity, pneumatic, or springs to quickly release the Central Shutdown Rod (CSR) containing neutron absorber into the reactor core stopping nuclear reaction. Generally, the motor/actuator resides outside the reactor, but the motion is transmitted through a penetration into the pressure vessel. There are also designs where the control rod drive incorporates magnetic latches with coil residing outside the pressure boundary for precise position control of the rod. We are proposing a magnetic coupling to position the shutdown rod without any penetration into the pressure vessel for the entire drive length of TCR shutdown rod. Electromagnets are currently used in non-power nuclear reactors for shutdown rods, but these electromagnets are resident inside the reactor pressure vessel. This paper will describe the use of an electromagnet outside the pressure vessel to position and release the shutdown rod. A prototype was developed at ORNL to demonstrate the concept, and a design optimization of the ferritic core and material was conducted to maximize the lift force of the electromagnet. An elevated temperature testing was also performed to ensure that the system will perform under the temperature conditions inside an operating reactor.

Fountain, Eliott J.↗
DOE OSTI2022

Controlling a power output of a nuclear reaction without control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above the reactor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗
DOE OSTI2022

Steam generator for fuel cell applications

A steam vaporizer assembly includes an internal steam generator having a vessel configured to hold water, a vaporizer unit having a heating element configured to heat the water to generate saturated steam; and a controller configured to: cause the heating element to heat the water to a stand-by temperature; and while maintaining a water level of the water in the vessel between two control points: maintain the water in the vessel at the stand-by temperature until steam generation is required, and when steam generation is required, heating the water in the vessel from the stand-by temperature to a temperature at or above a vaporization temperature of the water using a heating element, to generate the steam.

Dobek, Frank↗
DOE OSTI2022

Heat transfer systems for nuclear reactor cores, and related systems

A system for transferring heat from a nuclear reactor comprises a nuclear reactor comprising a nuclear fuel and a reactor vessel surrounding the nuclear reactor and a heat transfer system surrounding the nuclear reactor. The heat transfer system comprises an inner wall surrounding the nuclear reactor vessel, first fins coupled to an outer surface of inner wall, an outer wall between the inner wall and a surrounding environment, and second fins coupled to an inner surface of the outer wall and extending in a volume between the outer surface of the inner wall and the inner surface of the outer wall, the outer surface of the inner wall and the first fins configured to transfer heat from the nuclear reactor core to the second fins and the inner surface of the outer wall by thermal radiation. The heat transfer system may be directly coupled to the nuclear reactor vessel, or may be coupled to an external reflector surrounding the nuclear reactor vessel. Related heat transfer systems and systems for selectively removing heat from a nuclear reactor are disclosed.

Ougouag, Abderrafi Mohammed El-Amine↗
DOE OSTI2023

A New Offering for the Seaman Status Labyrinth - Seaman Status for Nuclear Reactor Operators on Floating Nuclear Power Plants

Floating nuclear power plants present a unique operating environment for land-based nuclear reactor operators. Traditionally located in the control room of a nuclear power plant on land, development of floating nuclear power plants exposes the traditional land-based employees to the marine environment. With the extension of nuclear power generation facilities into the maritime domain, do nuclear reactor operators working on a floating nuclear power plant qualify as seaman under maritime law? Applying existing maritime law, the answer is no, a nuclear reactor operator who operates the nuclear reactor on a floating nuclear power plant does not qualify as a seaman because their work is not in support of the mission of the vessel and the reactor is not connected to a vessel because a floating nuclear power plant is not a vessel. Applying the analysis developed by the Supreme Court in Chandris v. Latsis and the recent Sanchez v. Smart Fabricators of Texas, L.L.C. en banc decision by the Fifth Circuit, a nuclear reactor operator on a floating nuclear power plant does not qualify for seaman status under the Jones Act because their function supports the operation of the reactor and the structure on which the reactor resides does not meet the reasonable person standard established in Lozman v. City of Riviera Beach. Further, existing case law highlights that rendering a structure practically impossible to move eliminates the structure from consideration as a vessel. Because a floating nuclear power plant may be anchored at a seaport or anchored offshore but connected via transmission cables and protected by physical protection barriers, a floating nuclear power plant, with no current means of propulsion is rendered a power plant on water, which is its true function. Recognizing that technological change may alter the conclusion presented in this Article, current designs and structures that exist illustrate the intersection between nuclear and maritime law and the ever-evolving concepts that underpin seamen status in maritime law.

Fialkoff, Marc↗
DOE OSTI2020

Innovative Approaches to the Decommissioning of a Redundant Plutonium Processing Facility - 20147

At Sellafield, in the North West of the United Kingdom (UK), there are a number of redundant plutonium processing facilities. Over the past three decades several decommissioning strategies have been deployed to clean up these redundant facilities. Whilst there have been many successes previously reported, there remain many challenges. As previously presented at WM2017 (Paper Ref 17081) some of these challenges have recently been addressed through a new approach which introduced a number of innovations. The purpose of these innovations is to seek ways to significantly reduce risks to both decommissioning personnel and the environment by minimising the extent of physical 'hands on' decommissioning activities and by simplifying the waste production process. This paper provides an update on the progress of these innovations since 2017. Additionally, the paper also provides details of the new decommissioning technologies which have been introduced over the past 2 years to enable the most challenging aspects of the redundant plutonium processing facility to be decommissioned. These new decommissioning technologies address challenges associated with the preparation of process vessels and pipework to enable their safe in-situ size reduction by remote means: - New decommissioning techniques have been introduced to allow penetrations to be remotely cut into process vessels and pipework to provide a means of accessing and controlling the venting of any potential hydrogen gas present. - Additionally, a new large scale bespoke remote diamond wire cutting machine has been designed and manufactured to enable the process vessels to be size reduced in situ. Both of these new cutting technologies have involved collaborative working with specialist equipment suppliers including extensive development, trialling demonstration and commissioning prior to bringing the equipment into operational service. Several years of work have now culminated in reaching the final stages of decommissioning of the redundant plutonium processing facility with decommissioning of the final process vessels now underway and completion expected by the end of 2019. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Residual Dose Rates Analyses from Vent Lines Shield Block on the Target Cart at Spallation Neutron Source

The target carriage supports the mercury target vessel in the Spallation Neutron Source (SNS) in a cantilevered fashion and supplies the liquid mercury flowing to and from the target vessel. The target carriage is designed to move the target vessel into its target monolith position to receive the proton beam and then remove it horizontally back into an adjacent service cell, where the target vessel can be exchanged about 2-3 times per calendar year. At the time of the performed studies, the SNS mercury target can accept a proton beam at 1 GeV energy and 1.4 MW power; however, it will be upgraded to receive 1.3 GeV energy at 2 MW by the end of 2024 year [1]. The cart is practically a shield block on wheels to minimize fluxes and dose rates in the service cell downstream of the carriage. There are vent lines located inside a passage on top of the target carriage, and they are covered by a stepping-up steel shield block (vent lines shield block). Vent lines are permanently attached to the shield block. New vent lines with a different design are going to be installed. These lines are shielded by the new steel block to which the vent lines are attached. The estimate of the residual dose rates from the block after it is extracted from the target carriage is performed to estimate the radiation fields and to decide on the type of storage container to be used.

Popova, Irina I↗
DOE OSTI2024

Open aperture flow cells for on-line optical analysis of process fluids

A flow cell system includes a vessel and a fluid located in the vessel. A fluid surface of the fluid can be vented to a first gas pressure. The fluid surface can have a first cross-sectional area. The flow cell system includes a conduit in fluid communication with the vessel and positioned downstream of the vessel. The conduit can have a region that includes one or more orifices and has a second cross-sectional area. The second cross-sectional area can be less than the first cross-sectional area. The one or more orifices can be vented to a second gas pressure. The second gas pressure can be equal to or greater than the first gas pressure. Methods for analyzing a process fluid can include characterizing the fluid in the conduit.

Launiere, Cari A.↗
DOE OSTI2024

Controlling a power output of a nuclear reactor without using control rods

A nuclear power system includes a reactor vessel that includes a reactor core that includes nuclear fuel assemblies configured to generate a nuclear fission reaction. A representative nuclear power system further includes a riser positioned above there actor core and a primary coolant flow path that extends from a bottom portion of the reactor vessel, through the reactor core, and through an annulus between the riser and the reactor vessel. A primary coolant circulates through the primary coolant flow path to receive heat from the nuclear fission reaction and release the heat to a power generation system configured to generate electric power. The nuclear power system further includes a control rod assembly system positioned in the reactor vessel and configured to position control rods in only two discrete positions.

Callaway, Allyson↗
DOE OSTI2024

NSTXU Diagnostic Disruption Dynamic Loading Represented by Response Spectra

This article presents the results of transient dynamic simulations of loads due to disruption eddy currents on the NSTXU vacuum vessel. Dynamic loading at diagnostic mounting locations is expressed as response spectra derived from the time history results of the dynamic structural simulations of a variety of disruption scenarios. The disruption simulations draw on a history of the project assessments of worst case disruptions for specific components. Major efforts to assess disruption loading have included the vacuum vessel which is the major structural support for the machine, as well as the passive plates (PPs), high harmonic fast wave (HHFW) antenna, and centerstack casing. Each one of these efforts included transient electromagnetic simulations producing time-dependent eddy current Lorentz loads (and in some cases halo loads) which then were applied to time-dependent structural dynamic analyses intended to obtain the proper dynamic amplification factors. In some instances, the EM model and structural model were identical allowing direct transfer of EM forces to the structural model. In other cases, the EM and structural model were not identical and the vector potential (VP) transfer method was used. The results files from these analyses were available (or re-run) to post process in ANSYS Classic time history postprocessor. In conclusion, the ANSYS command is used to create response spectra from time history data at desired points on the vessel.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI1994

Nuclear Safety [Vol. 35, No. 2, July-December 1994]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 179 Consideration of Postaccident Consequences in the Determination of Safety Objectives for Future Nuclear Power Plants in France, D. Queniart, A. Sugier, and J. Lochard; ACCIDENT ANALYSIS: 187 Nuclear Safety Research: The Phebus FP Severe Accident Experimental Program, P. von der Hardt, A. V. Jones, C. Lecomte, and A. Tattegrain; 205 Containment Performance Analysis of the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory, S. H. Kim, R. P. Taleyarkhan, and V. Georgevich; 213 Assessment of Fission Product Deposits in the Reactor Coolant System: The DEVAP Program, G. Le Marois and M. Megnin; 222 Erratum to “A Review of the Available Information on the Triggering Stage of a Steam Explosion," Vol. 35, No. 1, CONTROL AND INSTRUMENTATION: 223 Effects of Normal Aging on Calibration and Response Time of Nuclear Plant Resistance Temperature Detectors and Pressure Sensors, H. M. Hashemian; DESIGN FEATURES: 235 Defense in Depth Against the Hydrogen Risk—A European Research Program, F. Fineschi; ENVIRONMENTAL EFFECTS: 246 Technical Note: A Preliminary Analysis of the Risks to Hong Kong Resulting from Potential Accidents of Daya Bay Nuclear Power Plant, Z. Shi and X. Wei; OPERATING EXPERIENCES: 253 Reactor Shutdown Experience, Compiled by J. W. Cletcher; SPECIAL SECTION ON TMI-2 VESSEL INVESTIGATION PROJECT: 256 Three Mile Island—New Findings 15 Years After the Accident A. M. Rubin and E. Beckjord; 269 Relocation of Molten Material to the TMI-2 Lower Head, J. R. Wolf, D. W. Akers, and L. A. Neimark; 280 Insight Into the TMI-2 Core Material Relocation Through Examination of Instrument Tube Nozzles, L. A. Neimark; 288 Physical and Radiochemical Examinations of Debris from the TMI-2 Lower Head, D. W. Akers and B. K. Schuetz; 301 Results of Metallographic Examinations and Mechanical Tests of Pressure Vessel Samples from the TMI-2 Lower Head, D. R. Diercks and G. E. Korth; 313 Margin-to-Failure Calculations for the TMI-2 Vessel, J. Rempe, L. Stickler, S. Chavez, G. Thinnes, R. Witt, and M. Corradini; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 328 1993 Accident Sequence Precursor (ASP) Program Results, L. N. Vanden Heuvel, J. W. Cletcher, D. A. Copinger, J. W. Minarick, B. W. Dolan, and P. D. O’Reilley; RECENT DEVELOPMENTS: 339 Reports, Standards, and Safety Guides, D. S. Queener; 345 Proposed Rule Changes as of June 30, 1994; ANNOUNCEMENTS: 234 Thirty-First Annual Meeting of the National Council on Radiation Protection and Measurements; 356 1995 International Incineration Conference; 358 Fifth International Controls and Instrumentation Conference; 359 ANS International Topical Meeting on Safety of Operating Reactors; 359 Fifth International Conference on Nuclear Criticality Safety; 360 International Conference on Probabilistic Safety Assessment Methodology and Applications; 351 The Authors; 357 Reviewers of Nuclear Safety, Vol. 35.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗