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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Advancing radioactive material research method: the development of a novel in situ particle-attached microfluidic electrochemical cell

This study aims to develop a microgram-scale microfluidic electrochemical cell (E-cell) for investigating the redox behavior of uranium oxide (UO 2 ). The traditional bulk electrochemical methods may require shielded facilities to investigate the hazardous materials, e.g., spent nuclear fuel, due to high radiation levels. Microfluidic E-cells offer advantages such as reduced radiation exposure, control over fluid flow rates, and high-throughput capabilities. Methods: The design of the E-cell considers electrode morphology, adhesion to a thin membrane, electrode configuration, and vacuum compatibility. Three techniques, including FIB-SEM lift-out, Au coating, and polyvinylidene fluoride (PVDF) binder, are explored for fabricating and attaching microgram quantities of UO 2 as working electrodes. The PVDF binder method proves to be the most effective, enabling the creation of a vacuum-compatible microfluidic E-cell. Results and discussion: The PVDF binder method demonstrates successful electrochemical responses and allows for real-time monitoring of UO 2 electrode behavior at the microscale. It offers chemical imaging capabilities using in situ SEM/EDS analysis. The technique provides consistent redox outcomes similar to bulk electrochemical analysis. Conclusion: The development of a microgram-scale microfluidic electrochemical cell using the PVDF binder technique enables the investigation of UO 2 redox behavior. It offers a low-risk approach with reduced radiation exposure and high-throughput capabilities. The technique provides real-time monitoring and chemical imaging capabilities, making it valuable for studying spent nuclear fuel systems and material characterization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Criticality Accident Alarm System Modeling for the Uranium Processing Facility

The Criticality Accident Alarm System (CAAS) for the Uranium Processing Facility (UPF) provides coverage for two primary processing facilities. For UPF, detailed documentation and drawings of the design are readily available. While this design information allows for some CAAS modeling features to match the actual design with a high level of accuracy, assumptions and simplifications are necessary as it is unrealistic and unnecessary to replicate many aspects of the design. A set of modeling guidelines and techniques were used to develop consistency between the MCNP6 models. While ensuring conservatism, overly conservative choices must be avoided to demonstrate the minimum accident of concern can be detected. To ensure proper modeling choices were made for the final models, a group of parametric studies were performed to evaluate the impact on the dose rates detected at the individual detectors. Some of these modeling choices and/or associated parametric studies include: - Modeling the grid steel area above the detectors as homogeneous stainless steel and determining the impact on detectability with an increasing density. - Modeling the presence of equipment in close proximity to the CAAS detectors. - Determining an appropriate material density to represent miscellaneous items not specifically modeled. - Evaluating the impact of sprinkler activation. The results of UPF CAAS analyses show that the modeling assumptions made based on the detailed design and parametric studies are appropriate and demonstrate accurate yet conservative models have been developed for detecting minimum accident of concern within the UPF.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Acoustic Monitoring of Pyroprocessing Equipment

This paper provides an introduction to using acoustic monitoring to advance detection techniques for pyroprocessing in support of nuclear safeguards and non-proliferation. The usage of free air acoustic monitoring has been previously demonstrated at Idaho National Laboratory (INL) facilities such as the Advanced Test Reactor and the National Security Test Range. However, the proposed work revolves around a new deployment environment, the Fuel Conditioning Facility, that brings forward several questions regarding the performance of the technology in non-free air media. The confinement of the pyroprocessing equipment to a heavily shielded hot cell, the atmosphere of the hot cell containing argon gas, and the radiation dose inside the hot cell are all new environments for acoustic monitoring. To our knowledge, acoustic measurements have not been completed in such an environment before. This offers a new opportunity to study not only the acoustic signatures of the equipment inside of the hot cell, but also the propagation of the signals through the hot cell and at distances away from their origination. The objective of this paper is to explain the planned instruments to monitor the Fuel Conditioning Facility in order to evaluate acoustic signals emitted from equipment during various stages of operation. Identifying these signals can potentially enable the identification of specific pieces of equipment used in pyroprocessing and produce information of their operational status. If successful, this type of monitoring could offer a new method to aid in safeguards and proliferation detection of pyroprocessing activities.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Analyses of Dose Rates for Second Inner Reflector Plug Replacement at Spallation Neutron Source

The Inner Reflector Plug (IRP) is a central component of the Spallation Neutron Source (SNS) target monolith, which houses the liquid mercury target and four liter-sized neutron moderator units. It is exposed to high-level radiation fields during routine operation and builds up significant activity. The IRP needs to be replaced due to moderator neutron poison and decoupler burn-out, which is used for shaping neutron pulses. The first IRP exchange took place in March 2018 and next one, which is IRP2, is planned to be replaced during the facility maintenance period in the end of December 2025, beginning of January 2026 calendar year. The replacement of IRP is a complex task due to its location in the area receiving high irradiation, being under significant amount of shielding, and size, which requires removal in segments. For planning the replacement workflow to help to reduce radiation exposure to workers dose rates for each stage of extraction/replacement operation are predicted.

42 ENGINEERING↗
DOE OSTI2021

A Preliminary Investigation for the Use of Digital Gamma-Gamma Coincidence Spectrometry to Determine 239 Pu

Gamma-ray spectrometry is one of the most effective ways to determine the activity of 239 Pu, depending on activity levels. However, often high backgrounds in complex spectra with low amounts of 239 Pu can increase detection limits. The effectiveness of the use of gamma-gamma spectrometry in the characterization of 239 Pu was studied for the first time. Using the XIA Pixie-16 digital pulse processor, the gamma-gamma system was studied to establish unshielded background radiation, source to detector distances, Poisson statistics and gating of several gamma-rays from 239 Pu. The results clearly show that the distance between gamma ray spectrometers has a much bigger effort in counting statistics for the gamma-gamma coincidence detectors than a single detector. In addition, measurements demonstrated the background radiation to be virtually negligible, meaning gammagamma spectroscopy can still be very effective without requiring the usual lead shielding. Other measurements were taken to assure Poisson statistics were attained in the digital system. Preliminary measurements revealed 3 orders of magnitude background reduction for the measurement of 239 Pu using gated gamma-rays.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Reduced Geometry for the NB5 Velocity Selector Shielding Box

This report details the reduced NB5 velocity selector shielding box that has been isolated from the complete HFIR CGH model in order to speed up detailed calculations of the velocity selector shielding. This shielding model is intended for outsourcing shielding optimization calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Shield Design for 50 kW Three-Phase Wireless Charging System

In this paper, magnetic field emissions (MFE) from a 50 kW bipolar coil-based three-phase wireless charging system (WCS) is investigated. Then, a shield design is proposed to suppress the MFE below the safety limit identified by the International Commission of Non-Ionized Radiation Protection (ICNIRP) guidelines. Bipolar coil-based three-phase (3φ) WCSs provide extremely high power-density; therefore, the 3φ WCSs are highly promising for high-power fast-wireless-charging of the electric vehicles. For the high-power EV application, it is a critical challenge to design a shield to suppress the MFE below the safety limit. The traditional aluminum shield is found ineffective for the bipolar 3φ-WCS. The proposed shield is designed with high-permeability magnetic material, i.e. ferrite, nanocrystalline material, etc. The proposed shield is simulated in finite element analysis (FEA) and the FEA results show that the proposed magnetic shield effectively suppresses the magnetic field emission below the ICNIRP limit. The generalized design of the proposed shield can be used for even higher power bipolar coil based WCSs.

Mohammad, Mostak↗
DOE OSTI2021

Dose Rate Analysis of the WCS Consolidated Interim Storage Facility

This report describes a confirmatory calculation for a proposed consolidated interim storage facility (CISF) to support the US Nuclear Regulatory Commission’s review of the license application submitted by the Interim Storage Partners. The scope of this report is limited to an assessment of the annual dose from direct radiation associated with Phase 1 of the Waste Control Specialists (WCS) CISF and the determination of the minimum controlled area boundary based on the 10 CFR 72.104 limit of 25 mrem annual dose to the whole body. The confirmatory dose rate calculations used the source term and shielding calculation capabilities of the SCALE 6.2.3 computer code system. The calculation method employed in the confirmatory calculations uses a detailed Monte Carlo radiation transport simulation from source to dose rate. This method differs from the two-step method used in the safety analysis report (SAR), which requires determination of the photon and neutron energy and angular distributions on the cask external surface and then the use of this surface source in a new radiation transport calculation to determine dose rate as a function of distance from the storage facility. A simulation was made using the complete site geometry (all casks present) but with only one cask containing source, and a total of 467 independent calculations were performed to obtain the dose rate maps produced by each storage cask. Basic input data and assumptions used in the SAR (e.g., cask design parameters and design basis assembly characteristics) were used in this confirmatory calculation for consistency with the SAR. Local atmospheric conditions for skyshine calculations were simulated using US standard atmosphere data, which describe average air temperature, pressure, and density as a function of altitude. The determined minimum required distances from the WCS Phase 1 CISF to a contour defined by the 25 mrem annual dose limit were approximately 623 m (0.387 miles) in the N-NE direction, 594 m (0.369 miles) in the SSW direction, and 533 m (0.331 miles) in the E-SE direction and the W-NW direction. The annual dose results of the confirmatory calculation are in relatively good agreement with the annual dose values presented in the WCS SAR. The annual dose produced by the Phase 1 CISF at the nearest site boundary, 0.75 miles away from the CISF, was estimated by this study to be approximately 6.02E-02 ± 2.0E-04 mrem, which is significantly smaller than the 25 mrem annual dose limit provided in 10 CFR 72.104. The SAR annual dose rate value for the nearest site boundary was 7.52E-2 mrem, which included 7.73E-03 mrem due to postulated leakage of the FO-, FC-, and FF-canisters. A comparison with the dose rate values reported in the SAR for 16 detector locations, D1 through D16, on the protected area boundary is also included in this report. This comparison shows that the SAR determined higher annual dose values for detectors D1 through D14 by 12% to 45% and lower annual dose values for detectors D15 and D16 by 18% and 8%, respectively, as compared with this ORNL confirmatory calculation. Sensitivity studies have been performed to support a comprehensive review of the application. In these studies, the effects on dose rate of the modeled air volume, density, and humidity were analyzed. The sensitivity calculations showed that the air volume used in the Monte Carlo simulations was adequate and that a larger air volume would not further increase the dose rate due to skyshine contributions. The sensitivity analysis showed that simulation of local atmospheric conditions is important for the accuracy of dose rate estimates at large distances from the facility. Air humidity was demonstrated to have negligible effects on dose rate values for the simulated local atmospheric conditions.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

Radiation Effects in Concrete for Nuclear Systems

Concrete is massively used in the construction of critical safety-related components in nuclear power plants. Among those, the concrete biological shield is exposed to high levels of neutron and gamma irradiation exiting the reactor pressure vessel. This article provides a comprehensive overview of the state-of-the-art knowledge on irradiated concrete. First, the fundamental irradiation-induced degradation mechanisms occurring at the level of the concrete constituents such as rock-forming minerals and hydrated cementitious phases are examined. The critical role of the mineralogical and chemical composition regarding concrete resistance against radiation effects such as amorphization, irradiation-induced expansion and radiolytic gas release are detailed. Second, the interactions between the varied irradiated phases are comprehensively analyzed and put in perspectives with the effects of prolonged irradiation on the structural properties of concrete aggregates, cement paste, and concretes. Finally, the significance of irradiation for the long-term operation of nuclear power plants is discussed in the light of the most recent knowledge and its limitations.

Le Pape, Yann↗
DOE OSTI2019

Evaluation of Skyshine from an Accelerator Facility: Dependence on Distance and Angle

Neutron skyshine from Linac Coherent Light Source II 4 GeV electron beam operation at SLAC National Accelerator Laboratory can contribute to prompt radiation exposure to the public at distances far beyond the accelerator tunnel housing. One of the shielding design requirements at SLAC is that the annual dose to a member of the public is no more than 0.05 mSv y -1 . Our work uses Monte Carlo code FLUKA to simulate the generation of neutrons from 4 GeVelectron beam losses on a thick copper target inside a generalized geometry of the Linac Coherent Light Source II Beam Transport Hall accelerator tunnel section. The effective dose from neutron skyshine was characterized as a function of both distance from the tunnel wall (up to 1 km away) and angle relative to the beam direction (between 0° and 180°). Furthermore, this new methodology for evaluating neutron skyshine dose is applicable to high-energy GeV-range electron accelerator facilities both at SLAC and elsewhere.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Radiation Protective Coveralls for Fermilab’s SPOT Robot

Fermilab’s Accelerator Division recently purchased a SPOT robot from Boston Dynamics to be used to automatically measure the residual radiation in the beamline tunnels. Fermilab will work with the College of DuPage Fashion Studies Department to design and develop a prototype Tyvek-garments (coveralls and booties) that can be easily used and replicated by the Fermilab operators to shield the SPOT robot from collecting radioactive dust while deployed.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Microreactor Incident-Free Transportation Radiation Dose Assessment

This report, prepared by Pacific Northwest National Laboratory (PNNL) for the U.S. Nuclear Regulatory Commission, provides an estimate of the radiation dose related to the transportation of microreactors following the regulatory limits established in Title 10 of the Code of Federal Regulations (10 CFR) 71.47(b). The transportation of irradiated microreactors poses unique challenges. Microreactors may be challenged to meet the current 10 mrem/hr limit at 2 meters from the vehicle in 10 CFR 71.47(b). The need to modify packaging to be within practical weight and size limits for shipping may limit the amount of shielding that can be applied for a microreactor transportation package design. Operational strategies that would necessitate moving a reactor within a short time after its operating life may result in higher radiation dose rates associated with lower cooling times. This study estimates the radiation dose to maximally exposed individuals (MEIs) under the current regulatory radiation level limit of 10 mrem/hr at 2 meters from the vehicle as well as alternative limits of 10, 50, 100, and 200 mrem/hr at 2 meters from the vehicle. This analysis expands on prior incident free transportation radiation dose assessments conducted for spent nuclear fuel (SNF) and radioactive waste transportation. The key factors analyzed include the microreactor fuel type and burnup, shipment cooling periods, shielding designs integrated into transport packaging, and exposure scenarios for MEIs for highway and rail transport. This report considers additional distances that may be associated with compensatory measures to ensure safe transport conditions under varying radiation dose limits, including enhanced routing measures, escort protocols, vehicle inspections, and operational restrictions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Mass Optimization of a Multilayered Shield for Transportable Microreactors

The ability to easily transport microreactors is a major selling point for deploying microreactors to remote areas. However, this creates a unique shielding challenge, especially when the microreactor is being shipped after irradiation. A traditional reactor configuration utilizes a separate biological shield and pressure vessel to meet radiological shielding and pressure needs. The limited space available for transportable microreactors for both shielding and pressure vessels requires a revised assessment of separating out the biological shield and pressure vessel. To address these concerns, we examine a nuclear-grade sandwich composite (NGSC) that combines the reactor pressure vessel and biological shielding functions into a single component. Through a series of optimization problems for both transportation and operational use cases, the NGSC is able to minimize dose, minimize the vessel cost, and ensure that weight requirements are met for transportation. Initial results show that using a tungsten-tetraboride cermet in the first two layers of a six-layer NGSC provides adequate shielding for both use cases. These results show promise that an NGSC has enough overlap between operational and transportation cases to help reduce the design space for future analysis and assessment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Neutron Capture and Transmission Measurements and Evaluation of 54 Fe at the RPI LINAC [Slides]

This presentation covers neutron capture and transmission measurements and evaluation of 54 Fe (Iron) at the Rensselaer Polytechnic Insitute (RPI) Linear Accelerator (LINAC). Presentation conclusions find that 54 Fe is near its conclusion, covariance matrix generation is needed for both experiments before data is made available and released (Completion by April 2023). RPI will be conducting a full Resolved Resonance Region (RRR) evaluation following release of data (Completion by September 2023). Additionally, the neutron beam imager shows promise in providing an easy way to align samples in the most intense part of the neutron beam and there will be a new evaluation that will offer improvements in crit safety, shielding, and stellar applications.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Evidence of vapor shielding effect on heat flux loaded on flowing liquid lithium limiter in EAST

For this study, a lithium (Li) vapour layer was formed around a flowing liquid Li limiter to shield against the plasma incident power and reduce limiter heat flux in the EAST tokamak. The results revealed that after a plasma operation of a few seconds, the layer became clear, which indicated a strong Li emission with a decrease in the limiter surface temperature. This emission resulted in a dense vapour around the limiter, and Li ions moved along the magnetic field to form a green shielding layer on the limiter. The plasma heat flux loaded on the limiter, measured by the probe installed on the limiter, was approximately 52% lower than that detected by a fast-reciprocating probe at the same radial position without the limiter in EAST. Additionally, approximately 42% of the parallel heat flux was dissipated directly with the enhanced Li radiation in the discharge with the liquid metal infused trenches (LIMIT) limiter. This observation revealed that the Li vapour layer exhibited an excellent shielding effect to liquid Li on plasma heat flux, which is a possible benefit of liquid-plasma-facing components in future fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

CAAS-3S Radiation Testing for Y-12 and UPF with Godiva-IV

The Y-12 National Security Complex and the Uranium Processing Facility (UPF) selected the Mirion CAAS-3S as the Criticality Accident Alarm System for UPF and for Y-12 facilities replacing their legacy CAAS as part of efforts to extend the facility lifespans. As part of this process, the CAAS-3S system was exposed to a high radiation dose and dose rate during reactor testing with the Godiva-IV fast burst reactor. The reactor testing was designed around preliminary analyses that determined Y-12 and UPF requirements, and simulations of the radiation field within the reactor facility were used to determine reactor operating parameters, CAAS equipment locations, and the design of a neutron shield wall. This paper presents the design and results of the testing, and discusses how the test results were interpreted by criticality safety engineers at Y-12 and UPF.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

GEOMWRITER - A Workflow for Automatically Generating MCNP Geometry for Neutron Guide Simulations

This report describes the functionality of the geomwriter.py script, which is a python script for producing complex MCNP CSG geometries from the IGES CAD format. It is specifically written to handle neutron supermirror guides, but can handle more complex objects through utilizing a convex hull algorithm. This work flow reduces the amount of time spent on generating the surface definitions for a complex neutron guide system with hundreds of segments such that changes to the neutron guide system design in McStas can be rapidly incorporated into an MCNP model and integrated with the shielding geometry present in that model.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗