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At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Coaxial radio frequency connectors for high-power handling

Coaxial radio frequency (“RF”) connectors for high-power handling are disclosed. Specifically, a high-power male coaxial connector (“HPMC”) is disclosed. The HPMC includes a center conductor, an outer conductor disposed around the center conductor, an insulating layer positioned between the center conductor and the outer conductor, and a first elastomer. The outer conductor has an outer conductor front-end (“OCFE”) and the insulating layer has an insulating layer front-end (“ILFE”). The first elastomer is positioned between the center conductor and the insulating layer. The insulating layer may include an insulating layer cavity (“ILC”) extending inward into the insulating layer from the ILFE and the first elastomer may be within the ILC.

Farrell, James T.↗
DOE OSTI2021

Annual Summary Report for the Remote-Handled Low-Level Waste Disposal Facility—FY 2020

This Fiscal Year (FY) 2020 annual summary report documents the continued adequacy of the performance assessment (PA), the composite analysis (CA) and associated operating disposal-authorization statement (ODAS) technical basis documents for the Remote Handled (RH) Low-Level Waste (LLW) Disposal Facility at Idaho National Laboratory. Annual review of the adequacy of the PA and CA for the RHLLW Disposal Facility ensures that conclusions of the analyses remain valid in accordance with requirements of Department of Energy (DOE) Order 435.1 Change 1, “Radioactive Waste Management.” In FY 2020, no significant operational changes or other activities occurred that would cause deviation from the assumptions in the PA and CA pertaining to disposal geometry, verification of waste characteristics, tracking disposal inventories against total limits, facility closure design, or institutional controls. Fifteen total waste-canister shipments were received at the RHLLW Disposal Facility, and 15 total waste canisters were emplaced.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Westinghouse Quiver - A simple and safe system for the complete handling and storing of failed BWR and PWR fuel rods and fragments - 17049

Background: Extracted leaking fuel rods from fuel assemblies have been stored in the spent fuel pools at the Nuclear Power Plants for many years. Sending those fuel rods to an intermediate storage has been very difficult due to the uncertainty of the technical requirements. The Westinghouse Quiver is a container that can be used from the first moment when the leaking fuel rod is extracted from the Fuel Assembly. The leaking fuel rod can be placed in the Westinghouse Quiver. When the Quiver is filled, it can be sealed and dried and then treated as a regular fuel assembly during intermediate storage and final repository. Purpose of System: The purpose of the system is to store leaking fuel rods, broken rods and fuel rod fragments leak tight in the fuel pool and after vacuum drying and permanent sealing at an intermediate storage facility. The Quiver is designed to be fully compatible with a regular fuel assembly. That means all handling of the Quiver in a Nuclear Power Plant, during transportation or storage is also identical to that of a regular fuel assembly. Fulfilling these criteria together with functional testing, the Quiver is considered acceptable for long time storage of leaking fuel rods, both in dry or wet condition. The Quiver has been qualified in Switzerland for use in the spent fuel pool, intermediate storage, and for a final repository. The system will be implemented as well in Sweden during 2016-2019. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Remote-Handled TRU Storage Challenges at Sandia National Laboratories/ New Mexico - 16071

The events at the Waste Isolation Pilot Plant (WIPP) in February 2014 had a significant impact on the packaging, storage, and disposal options at Sandia National Laboratories/New Mexico (Sandia) as with other transuranic generator sites. In 2012, Sandia had identified accountable material for disposal, fresh and irradiated mixed oxide (MOX) and uranium oxide (UO{sub 2}), which had been used in experiments conducted in the 1970's and 1980's. During the 2012/2013 timeframe, discussions with WIPP determined the samples and experimental vessels qualified as remote-handled (RH) transuranic (TRU) waste for disposal at WIPP. Sandia began discussions with WIPP to determine the process to repackage and characterize the material for disposal at WIPP in addition to obtain funding to complete the project. By January 2014, Sandia had contracted with Nuclear Waste Partnerships, LLC (NWP) to provide Central Characterization Program (CCP) certified visual examination (VE) and dose-to-curie (DTC) support to package and certify the accountable material/waste. Work began on February 3, 2014 with the arrival of two VE operators. On February 5, 2014, a salt truck caught fire in the underground and on February 14, 2014, the radiological release occurred at WIPP initiating a suspension of shipments. The length of the suspension was not known at the time. Sandia was only a couple of days into this long-term project, had signed a contract with NWP for support, and was still responsible for meeting earlier negotiated milestones. There were 37 parent containers which included experimental vessels from 3 to 10 feet long and unused material pieces. The original plan was to package each of the 37 parent containers separately, but several issues, including the suspension of shipments, made it imperative that a new path forward be developed. Because space is limited in the Auxiliary Hot Cell Facility (AHCF), the storage configuration for criticality safety and material-at-risk was of the highest importance. Therefore, packaging in the fewest number of containers without exceeding WIPP limits, keeping the accountable items together, and not violating any safety basis requirements was Sandia's goal. (authors)

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

Development of the Waste Acceptance Criteria for the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility - 19143

The Idaho National Laboratory (INL) Remote-Handled Low-Level Waste (RH LLW) Disposal Facility has been constructed, has received operational approval from the Department of Energy (DOE), and is expected to receive its first waste shipment in FY 2019. The RH LLW Disposal Facility will support the needs of INL and the Naval Reactors Facility (NRF). RH LLW destined for the facility includes waste currently in storage from previous programs and projects conducted at INL; newly generated waste from programs and projects conducted at INL; and waste from NRF spent fuel processing activities. This new facility supports RH LLW disposal needs for the ongoing mission of INL and NRF. Waste to be disposed of at the RH LLW Disposal Facility may exhibit dose rates that range from 0.002 Sv/hr to 600 Sv/hr. Waste types include resins and activated metals as well as surface-contaminated debris generated from existing and new missions. The total number of canisters currently planned for disposal at the RH LLW Disposal Facility is 939 with the possibility of future expansion. The projected waste volume to be disposed of is 3,100 m{sup 3}. The RH LLW Disposal Facility will receive a wide variety of RH LLW and is in preparations to receive its first canisters in FY 2019. Waste acceptance criteria (WAC) and the radioactive waste management basis have been developed and implemented in accordance with the applicable DOE Order. In order to ensure that DOE requirements and guidance were appropriately implemented in the RH LLW Disposal Facility WAC, the project worked closely with the DOE Low Level Waste Disposal Facility Federal Review Group (LFRG) representatives and followed the applicable DOE standard. This paper provides an overview of the development of the WAC, including waste and waste generator certification requirements, for the RH LLW Disposal Facility. The paper illustrates the importance of maintaining strong ties to the development and revision of the RH LLW Disposal Facility Performance Assessment (PA), Documented Safety Analysis, Facility Procedures, Readiness Reviews, DOE Order, and applicable Technical Standards. The paper also provides an overview of the RH LLW Disposal Facility, WAC generator certification requirements, distinctive challenges to acceptance of legacy waste, and the unique approach the RH LLW Disposal Facility has taken to ensure compliance with PA driven limits. (authors)

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

Readiness Process and Transition to Operation for a Remote-Handled Transuranic Packaging Facility at Hanford - 19298

CH2M Hill Plateau Remediation Company manages the Sludge Retrieval Project, which has the mission to develop and deploy technologies to safely retrieve and package remote-handled transuranic sludge that is currently stored in the 105K West Reactor Fuel Storage Basin located in the 100-K Area of the U.S. Department of Energy's Hanford Site near Richland, Washington. In 2017, CH2M Hill Plateau Remediation Company installed and commissioned a process for removal of the sludge. A contractor Operational Readiness Review was completed in March 2018, and a U.S. Department of Energy, Richland Operations Office Operational Readiness Review was completed in May. The first sludge transfer was successfully completed on June 12, 2018, proving that the Engineered Container Retrieval and Transfer System process equipment worked as intended. This paper will describe the issues and successes along the path of readiness and startup, as well as the challenges that remain and CH2M Hill Plateau Remediation Company's plan to overcome them. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

A Preliminary Radiological Risk Assessment Model for Disposition of Remote-Handled Transuranic Wastes at Los Alamos National Laboratory Area G - 20116

The U.S. Department of Energy (DOE) operates a low-level radioactive waste (LLW) disposal site at Material Disposal Area G, in Los Alamos, New Mexico, USA. Area G has been the primary LLW disposal site for Los Alamos National Laboratory (LANL) since the 1960's. In addition to LLW, Area G is host to a variety of other wastes, the disposition of which must be determined before closure of the site. A probabilistic Radiological Risk Assessment (RRA) for Area G is used in order to support decision making regarding some wastes that are not addressed in the extant Area G Performance Assessment (PA) and Composite Analysis (CA). Between 1979 and 1987, 33 special shafts were augered into the Bandelier Tuff at Area G. This volcanic tuff is present across Pajarito Plateau on the eastern slopes of the Jemez Mountains, and varies widely in its consistency, from weakly indurated non-welded layers to welded layers that uphold the mesa cliffs of the Plateau. These mesas are home to LANL, Area G, and the townsites of Los Alamos and White Rock, with residences about 1400 m from Area G. The 33 Shafts were lined with steel casing, and contain remote-handled (RH) transuranic wastes (TRU) resulting from experiments and analysis performed in special glove boxes at the Chemistry and Metallurgy Research (CMR) facility at LANL. Some of these wastes originated as used nuclear fuel. The purpose of the Area G RRA is to evaluate the potential future risk to humans and the environment from the RH TRU in the 33 Shafts in the context of the risk associated with the surrounding wastes at Area G. The analysis is responsive to expectations outlined in DOE Order 458.1, Radiation Protection of the Public and the Environment, and is informed by the Manual and Guidance accompanying DOE O 435.1, Radioactive Waste Management. Because the waste meets the definition of TRU, the regulatory context necessarily takes into consideration the regulation governing the disposal of TRU from the U.S. Environmental Protection Agency (EPA): 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. Given the broader regulatory context for the RRA, the analysis is subject to different assumptions from those made in the existing DOE O 435.1 PA and CA, such as allowing for future occupation of the site. The analysis begins with a comprehensive evaluation of features, events, processes, and exposure scenarios (FEPS) for Area G and the wastes it contains. These FEPSs are screened to eliminate from further consideration those of extremely low probability and/or consequence, and a conceptual site model (CSM) is subsequently developed. The scope and structure of the Area G RRA Model is informed by this CSM, and the Area G RRA Model is developed using the GoldSim systems analysis modeling platform. This paper presents the initial version of a defensible, transparent, and reasonably realistic model, which is based on the state of knowledge of the wastes, the site, and the FEPSs that govern contaminant transport from wastes into the environment and subsequent exposures to humans and other biota. Probabilistic model input distributions represent uncertainties inherent in the real and modeled systems. The results of the Area G RRA Model inform decisions regarding the disposition of the RH TRU in the 33 Shafts. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Criticality safety analysis of spent fuel storage or handling disposals using burnup credit - loading curve calculation - 14445

MILLENNIUM and REEL SAS Companies have shared their own experience to develop a neutron and criticality calculations chain in order to perform isotopic composition predictions of spent fuel after irradiation in a reactor and criticality analysis of fuel storages taking into account burnup credit. The whole calculation sequence is divided into three steps, all calling SCALE 6.1.2 modules. The first two steps are dedicated to the evaluation of isotopic compositions of burnt fuel. As transport and depletion calculations in three dimensions appears to be way too long for the whole core (about several months), dealing with each fuel assembly one by one is the alternative solution MILLENNIUM and REEL SAS commonly chose to study. Therefore, the first step aims at determining the axial burnup profile of a given fuel assembly using a 3-D model with infinite reflections on lateral axes. The T6-depl sequence of TRITON module coupling KENO-VI and ORIGEN-S was used in this step. Sensitivity studies confirmed that addition of the 8 closest neighbors of the studied fuel assembly in the model has an insignificant impact on the axial burnup profile. However, fuel assemblies in the peripheral zone of the core must be treated separately with a model including the nearest fuel assemblies and the effect of peripheral water. Then, the second phase evaluates the isotopic compositions of each fuel assemblies at the center of the active height with a 2-D model under conservative core conditions. The T-depl sequence of the TRITON module coupling NEWT to ORIGEN-S is called in this step. Each fuel assembly must be studied separately to consider loading pattern changes occurring at each cycle. An infinite model with reflections on lateral axes is used and the closest neighbors of the fuel assembly are limited to 2 pins - which is around the size of the neutrons mean free path in the core. This is an innovative physical trick aiming at compensating for a limited computing power. Finally, isotopic compositions calculated are combined with the axial burnup profile of the first step to extrapolate the concentrations along the whole active length. Data can then be added in any criticality model, such as a spent fuel pit or any disposal handling fuel assemblies. Criticality analyses of burnt fissile material using burnup credit can be managed within a reasonable amount of time, on condition that sufficient proofs of code qualification for the studied range of parameters are provided, especially on fission products isotopic compositions calculation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Titanium Wire Arc Additive Manufacturing Inert Enclosure and Material Handling Safety Considerations

Wire arc additive manufacturing (WAAM) via metal inert gas (MIG)/gas metal arc welding (GMAW) is a viable option for fabrication of large-scale titanium parts; however, it introduces new safety hazards associated with both the material and the additional system hardware required for the process. Localized gas shielding of the weld arc via standard GMAW torch is inadequate for titanium due to its affinity for oxygen; thereby requiring the use of an inert enclosure to protect the weld from entraining oxygen. The use of the inert enclosure presents potential safety hazards such as operator asphyxiation and brings up discussion of confined space considerations. In addition, the titanium welding process creates pyrophoric titanium soot residue around the deposit, which can undergo deflagration during part cleaning and part removal. This paper provides an overview of the titanium WAAM process along with safety considerations for the design and operation of the inert enclosure as well as functional solutions for the safe handling of the titanium soot by-product.

Walters, Alex↗
DOE OSTI2024

Dynamic Simulation Modeling and Control of a Desiccant Assisted Direct-expansion Air Handling Unit

Desirable built environments demand simultaneous regulation of thermal comfort and indoor air quality (IAQ) with energy-efficient operation of heating, ventilation and air conditioning (HV AC) systems, which involves controls of temperature, humidity and airborne contaminants simultaneously. This paper presents the efforts of dynamic modeling and initial development control strategy for a desiccant-assisted multi-functional air handling unit (AHU) coupled with a direct-expansion rooftop unit (RTU) system, which aims to achieve multiple functions for indoor environment conditioning with energy efficient control. The RTU-AHU system includes a desiccant wheel for dehumidification and a conceptual direct air capture (DAC) filtering device for CO2 regulation. A Modelica-based dynamic model is developed for this conceptual system, and a simple decentralized control strategy is designed, which combines a differential-enthalpy based AHU return-air ratio control, a demand-controlled ventilation, and supply-air temperature humidity control via the RTU and DW controls. The proposed control method is evaluated with the Modelica simulation model for a selected set of scenarios

Pan, Chao↗
DOE OSTI2025

Restructuring and Optimizing Reactor Building Lifting Processes & Designing a Solution for an Overhead Door Handling Forklift Attachment

This poster presents two innovative projects aimed at enhancing operational efficiency and safety at the Advanced Test Reactor (ATR) Complex. The first project focuses on restructuring the existing lift book used for hoisting operations within the ATR Reactor Building. The current lift book, a cumbersome 150+ page document, is being transformed into a more efficient format using charts that allow for quick identification of maximum lift heights based on object footprint and weight. This new method also considers various parameters such as floor integrity and reactor status, dividing entries into four distinct charts to improve usability and safety during lifts. The second project involves designing a specialized forklift attachment for handling an overhead door at the ATR Reactor Building. The attachment is engineered to securely lift and lower a 1200lb, 14ft long door, ensuring safe replacement operations. The design process included research on forklift attachment codes, modeling in Autodesk Inventor, and testing through Finite Element Analysis (FEA) and hand calculations to confirm the attachment's structural integrity. Future work includes completing detailed drawings and an Engineering Calculations Analysis and Review (ECAR) document to proceed with manufacturing and assembly. Together, these projects demonstrate a commitment to optimizing reactor building operations through innovative engineering solutions and safety considerations.

42 - ENGINEERING↗
DOE OSTI2015

Human machine interface to manually drive rhombic like vehicles in remote handling operations

In the thermonuclear experimental reactor ITER, a vehicle named CTS is designed to transport a container with activated components inside the buildings. In nominal operations, the CTS is autonomously guided under supervision. However, in some unexpected situations, such as in rescue and recovery operations, the autonomous mode must be overridden and the CTS must be remotely guided by an operator. The CTS is a rhombic-like vehicle, with two drivable and steerable wheels along its longitudinal axis, providing omni-directional capabilities. The rhombic kinematics correspond to four control variables, which are difficult to manage in manual mode operation. This paper proposes a Human Machine Interface (HMI) to remotely guide the vehicle in manual mode. The proposed solution is implemented using a HMI with an encoder connected to a micro-controller and an analog 2-axis joystick. Experimental results were obtained comparing the proposed solution with other controller devices in different scenarios and using a software platform that simulates the kinematics and dynamics of the vehicle. (authors)

42 ENGINEERING↗
DOE OSTI2017

Assessment of the Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility Vault Concrete Data

This document provides a summary of the concrete performance data collected during vault fabrication and installation. The report includes the process by which quality data was collected during vault fabrication and installation, the inspection requirements, and a summary of resulting test data with the potential to impact vault system durability. This report also includes quantitative data collected on cured concrete samples poured for random concrete batches throughout the concrete component fabrication process. Quantitative data includes total porosity, bulk density, absorption, effective porosity, and gas-phase permeability. Chloride diffusivity data collected during the concrete mix selection process are also included for completeness. Data shows that the vault quality assurance program resulted in components received insignificant defects and damaged areas during the fabrication stage and during the vault installation stage. Data collected on the cured concrete samples show the gas-phase permeability and effective porosity are slightly higher than the data used to select the concrete mix designs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

Advanced Fuel Cycle Cost Basis Report: Module D1-7 Contact-Handled Pelletized Pressurized Heavy Water (PHWR) UOX Fuel Fabrication (Rev.1)

In addition to literature-based pressurized heavy-water reactor (PHWR) fuel price information in the 2017 AFC-CBR, the what-it-takes (WIT) unit cost data in this update is informed by new analysis and escalation of the 1978 PHWR-UOX fuel life cycle cost (LCC) data from ORNL reports prepared for the 1977–1980 Nonproliferation Alternative Systems Assessment Program (NASAP). (These reports are referenced and summarized in detail in Module D1-PR.) The PHWR fuel fabrication LCC data in these reports is scaled from a bottom-up cost estimate for a reference technology pressurized-water reactor (PWR)—uranium oxide (UOX) fuel fabrication plant by using algorithms that consider the manufacturing process complexity, fuel design complexity, plant floor space requirements, and the radiation and health, safety, and environmental (HS&E) regulatory environment of PHWR-UOX fuel production vis-à-vis light-water reactor (LWR)-UOX production (PWR fuel in this case). The module name has been changed from “Canadian Deuterium Uranium (CANDU)” to the more generic PHWR fuel fabrication in recognition that not all power reactors that might use this fuel type are considered. Unfortunately, the detailed algorithms and their design bases were not archived at the end of the NASAP effort of the commercial CANDU concept specifically developed in the middle of the last century by Atomic Energy of Canada Limited (AECL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1966

Handling qualities and structural characteristics of the hingeless-rotor helicopter

Several advantages and potential problem areas associated with the hingeless-rotor helicopter are discussed. The extent to which two of the advantages - increased control and damping moments - can be obtained with the hingeless-rotor helicopter is established. The upper limit on the available control and damping moments is associated with the excessive gyroscopic coupling that is obtained with heavy rotor blades. The poten­tial problem areas identified include a structural mode of oscillation very similar to that of a pendulum, which causes undesirably large roll accelera­tions under certain flight conditions, and large cyclic chordwise bending moments that occur during severe maneuvers (the primary structural problem of concern). Solutions are proposed which should eliminate future concern with these problems.

HELICOPTER↗