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At least 127 records · Page 7

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DOE OSTI2022

Power quality disturbances diagnosis: A 2D densely connected convolutional network framework

The fast and accurate diagnosis of power quality disturbances (PQD) aids in avoiding shutdowns and unnecessary procedures, concerning electric energy distribution systems. As such, a number of techniques have been tested and applied in order to reach this objective. Majority of the techniques applied are two-step based. On the first step, power quality disturbances features are extracted. Second step, considering features extracted, disturbance classification is implemented. Recently, relevant literature has presented data-driven signal processing-based approaches, as deep convolutional neural networks (DCNN), which can implement both processing steps while providing automated recognition of patterns and outliers in data. However, not considered by state-of-art, power quality disturbances are evolving in nature, while all possible regularities might not be represented in the dataset. In this work a 2 Dimension Densely Connected Convolutional Network (2D-DenseNet) framework is presented. Further, a case study with synthetic disturbance events are analyzed. Easy-to-implement formulation, built on the 2D-DenseNet, without hard-to-design parameters, highlight potential aspects for real-life implementation.

42 ENGINEERING↗

DOE OSTI2020

Automotive fuel cell stack and system efficiency and fuel consumption based on vehicle testing on a chassis dynamometer at minus 18 °C to positive 35 °C temperatures

In this work we present an in-depth laboratory technology assessment of a 2016 Toyota Mirai Fuel Cell (FC) vehicle based on chassis dynamometer testing. The 114.6 kW FC stack has a high dynamic response, which makes this powertrain a FC-dominant hybrid electric vehicle. The measured peak efficiency is 66.0% FC stack and 63.7% FC system with an idle hydrogen flow rate of 4.39 g/hr. The high FC system efficiencies at low loads match typical vehicle power spectrums, resulting in a high average vehicle efficiency of 62% compared to 45% and 23% for a hybrid electric vehicle and a conventional vehicle, respectively. An energy breakdown accounts for the FC stack losses, FC system losses, air compressor loads, and heater loads for different drive cycles and different thermal conditions. The cold-start North American city drive cycle (UDDS) energy consumption values are, respectively, 758, 581, 226, and 321 Wh/km at ambient conditions of -18 degrees C, -7 degrees C, -25 degrees C and 35 degrees C with 850 W/m 2 of solar loading. The FC system shutdown and startup processes at temperatures below the freezing point contribute to the increased hydrogen consumption. Additionally, the raw test data files are available for download, thus providing the research community with a public reference data on a modern production automotive FC system.

08 HYDROGEN↗

DOE OSTI2023

Design and requirements of a hydrogen component reliability database (HyCReD)

Hydrogen technologies are expected to play a key role in the decarbonization of several sectors including energy storage and transportation. Rigorous investigation and quantification of the risk and reliability issues associated with hydrogen technologies will be critical to ensuring both their wider adoption and safe, economical operations. Quantitative risk assessment (QRA) is an important tool that has been used to enable the safe deployment of many engineering systems, including hydrogen fueling stations and hydrogen storage systems. However, QRA studies require reliability data which is currently lacking for expanding applications of hydrogen systems. Here, to address this gap, we present a new structure for a hydrogen component reliability database (HyCReD) that can be used to generate reliability data to be used in QRA, reliability, safety studies, maintenance planning, and more. Building on our previous work examining four major hydrogen safety data collection tools (West et al., 2022) [1], our approach in this work was to consult scientific literature on reliability data collection as well as a number of existing reliability engineering databases in the oil & gas, chemical processing, and nuclear power plant sectors. The evaluation of these databases led to identifying best practices to be implemented in a data collection framework for a hydrogen component reliability database. Based on these best practices, a set of 24 requirements for the proposed database are presented, covering its characteristics and the types of data to be collected. We define the structure of the HyCReD database and 25 data elements to be collected, spanning system description, failure, shutdown, or near-miss events, and maintenance events. The data elements are then defined according to international standards used in the safety and reliability practice and potential choice lists are provided for each field. Since this database is being piloted for hydrogen fueling stations, a generic station component hierarchy developed by West (2021) [2] is used to standardize system data. Finally, we demonstrate populating the database with information extracted from five narrative reports on hydrogen fueling station incidents.

08 HYDROGEN↗

DOE OSTI2021

Modeling of fission and activation products in molten salt reactors and their potential impact on the radionuclide monitoring stations of the International Monitoring System

Molten Salt Reactors (MSRs) are one of six Generation IV reactor designs currently under development around the world. Because of the unique operating conditions of MSRs, which include molten fuel and the continuous removal of gaseous fission products during operation, work was performed to analyze the potential impact of emissions on the International Monitoring System (IMS) of the Comprehensive Nuclear-Test-Ban Treaty (CTBT). Simulations were performed to predict the production of IMS-relevant radionuclides in four MSR designs operating under two scenarios: (1) a sealed reactor with releases only during operational shutdown, and (2) continuous reprocessing or sparging of the fuel salt. From these production estimates the radioxenon and radioiodine signatures were extracted and compared to three current reactor designs (Pressurized Water Reactor (PWR), BWR, RBMK). In cases where continuous reprocessing of the fuel salt occurred, both the radioxenon and radioiodine signatures were nearly indistinguishable from a nuclear explosion. Estimates were also made of the potential emission rate of radioxenon for three reactor designs and it was found that MSRs have the potential to emit radioxenon isotopes at a rate of 10^15-8×10^16 Bq/d for 133Xe if no abatement is used. An assessment was made of activation products using a candidate fuel salt (FLiBe) mixed with corrosion products for the Thorium Molten Salt Reactor (TMSR-LF1).

Johnson, Christine M.↗

DOE OSTI2021

High-burnup fuel stress analysis prior to and during a LOCA transient

High-burnup fuel fragmentation and pulverization have been experimentally observed during simulated loss-of-coolant accident (LOCA) transient testing of high-burnup (>62 GWd/tU) nuclear fuel. The cause of this phenomenon is generally accepted to be the formation of stresses arising from the off-normal conditions. However, the primary mechanism(s) contributing to the stress formation is/are still being debated. Leading hypotheses suggest that the concurrence of fission gas bubbles and high burnup structure on the pellet periphery (i.e., the rim) leave the fuel susceptible to pulverization. Other hypotheses suggest that pulverization is artificially inflated during externally heated tests that do not reproduce prototypic fuel stresses and radial temperature profiles as the fuel transitions from full-power nuclear heating to decay heating. This work performs finite element analyses of fuel pellet temperatures and macroscopic stresses prior to and during a LOCA transient to identify (1) operating conditions that may exacerbate the fuel pellet stress state, (2) differences in predicted stresses with nuclear vs. furnace heating, and (3) how fuel cracking impacts the subsequent stress state in the pellet fragments. Analyses predict that macroscopic transient fuel stresses resulting from constrained thermal expansion will remain below the fracture stress of fresh UO 2 once the fragment size reaches ~4 mm or less. Calculated fuel stresses from simulated LOCA tests performed with representative nuclear vs. electrical heating rates are shown to be equal within 5% and essentially unchanged for all times after the first 27 seconds of the transients. Because the fuel stresses do not change once the temperature profile becomes uniform, macroscopic transient stresses would be no different than those following a normal shutdown. Because pulverization of high burnup fuel does not occur prior to a LOCA test, pulverization of high burnup UO 2 cannot be initiated by differential thermal stresses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Hierarchically structured metal carbides as conductive fillers in thermo-responsive polymer nanocomposites for battery safety

Integrating thermo-responsive polymer switching materials (TRPS) into lithium-ion batteries (LIBs) has been recognized as one of the most effective strategies to prevent thermal runaway under various abuse scenarios. However, the current methods to obtain TRPS cannot satisfy different practical applications. Herein, we develop a versatile strategy for the preparation of various metal carbides (e.g., tungsten carbide, molybdenum carbide) with controllable hierarchical structure that is featured with surface protrusion structure, which is critical for high conductivity and rapid thermal response. Systematic studies of the phase and morphology evolutions by advanced characterizations illustrate that the reducing agent and reduction rate are critical for developing the specific morphology. In conclusion, by using the above-mentioned carbides as the conductive fillers of TRPS, the resulting TRPS with a specially controlled shape exhibits over 5 order of conductivity improvement compared with common carbides with particulate morphology, in addition to reversible shutdown performance and effective thermal abuse protections toward safe LIB operation.

25 ENERGY STORAGE↗

DOE OSTI2023

Effect of inlet throttling on thermohydraulic instability in a large scale water-based RCCS: An experimental study

The objective of the present experimental study is to investigate the effect of inlet throttling on the thermohydraulic stability of a large scale water-based Reactor Cavity Cooling System (RCCS). The test was performed using the water-based Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne, which represented a ½ axial scale and 12.5° sector slice of the full scale Framatome 625 MW t SC-HTGR RCCS concept. A two-phase steady state was first established through direct condensate refill, followed by increased inlet throttling over 10 stages, corresponding to a loss coefficient K over the range of 0.05–653. With the inlet throttling gradually increased, the system experienced a unique transition process between stabilization and destabilization. Through a stability analysis, three instability mechanisms were identified in the present test, including a compound mechanism due to both natural circulation oscillations (NCOs) and density wave oscillations (DWOs), Type-II DWOs, and geysering.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Performance of prototype GE1/1 chambers for the CMS muon spectrometer upgrade

The high-luminosity phase of the Large Hadron Collider (HL-LHC) will result in ten times higher particle background than measured during the first phase of LHC operation. In order to fully exploit the highly-demanding operating conditions during HL-LHC, the Compact Muon Solenoid (CMS) Collaboration will use Gas Electron Multiplier (GEM) detector technology. The technology will be integrated into the innermost region of the forward muon spectrometer of CMS as an additional muon station called GE1/1. The primary purpose of this auxiliary station is to help in muon reconstruction and to control level-1 muon trigger rates in the pseudo-rapidity region 1.6 ≤ |η| ≤ 2.2. The new station will contain trapezoidal-shaped GEM detectors called GE1/1 chambers. The design of these chambers is finalized, and the installation is in progress during the Long Shutdown phase two (LS-2) that started in 2019. Several full-size prototypes were built and operated successfully in various test beams at CERN. Here, we describe performance measurements such as gain, efficiency, and time resolution of these prototype chambers, developed after years of R&D, and summarize their behavior in different gas compositions as a function of the applied voltage.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Layout and performance of HPK prototype LGAD sensors for the High-Granularity Timing Detector

The High-Granularity Timing Detector is a detector proposed for the ATLAS Phase II upgrade. The detector, based on the Low-Gain Avalanche Detector (LGAD) technology, will cover the pseudo-rapidity region of 2.4 < |n| < 4.0 with two end caps on each side and a total area of 6.4 m 2 . The timing performance can be improved by implanting an internal gain layer that can produce signals with a fast rising edge. It significantly improves the signal-to-noise ratio. The required average timing resolution per track for a minimum ionizing particle is 30 ps at the start and 50 ps at the end of the HL-LHC operation. This is achieved with several layers of LGAD. The innermost region of the detector would accumulate a 1MeV neutron-equivalent fluence up to 2.5 10 15 n eq /cm 2 including a safety factor of 1.5 before being replaced during the scheduled shutdowns. The addition of this new detector is expected to play an important role in the mitigation of high pile-ups at the HL-LHC. The layout and performance of the various versions of LGAD prototypes produced by Hamamatsu (HPK) have been studied by the ATLAS Collaboration. The breakdown voltages, depletion voltages, inter-pad gaps, collected charge as well as the time resolution have been measured and the production yield of large size sensors has been evaluated.

47 OTHER INSTRUMENTATION↗

DOE OSTI2021

A custom high-stability pulse generator for the test of a novel in-situ magnetic sensor developed to detect radiation damage in undulators

High-performance analog-to-digital converters (ADC) have been widely applied in many areas of science. For instance, magnetic field measurements based on the Faraday’s induction law require high-precision voltmeters to measure induced voltages. In this study in the context of free-electron lasers technology, the Magnetic Measurements Team at SLAC National Accelerator Laboratory proposed a novel in-situ radiation damage detection system (RDDS) for detecting small field variations in undulators. The system measures the flux change in a flexible printed-circuit coil attached to the magnet array during the undulator gap movement. The gap movement changes the magnetic field, which induces a voltage signal in the coil that is measured and integrated with an ADC. Although the system is capable of detecting relative flux changes better than 100 ppm, drift in the ADC’s gain or offset can cause apparent changes in the relative flux. This paper describes the first attempt to developed a high-precision verification circuit to perform ADC testing in the framework of the novel RDDS. The circuit generates a reference voltage pulse with a voltage–time integral relative precision better than 50 ppm for a few hundreds of mVs — the typical order of magnitude measured with the RDDS. The circuit’s design combines a fast and precise switch with a low-noise voltage reference. Long-term measurements allowed statistical analysis and showed that averaging the voltage–time integral of ten pulses gives the required 50 ppm stability. Moreover, reproducibility tests confirmed that the circuit’s output is invariable under small power supply instabilities and equipment shutdown. Instruments and applications designed to quantify the magnetic field by integrating voltage signals may use the pulse generator proposed in this paper for verification purposes.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

The CMS RPC system readiness for LHC Run-3 data taking

During Run-3, the LHC is preparing to deliver instantaneous luminosity in the range from 5 × 10 34 cm -2 s -1 to 7.5 × 10 34 cm -2 s -1 . To ensure stable data taking, providing redundant information for robust muon triggering, reconstruction and identification, the CMS RPC collaboration has used the opportunity given by the LHC long shutdown 2 (LS2), to perform a series of maintenance and preparation activities for the new data taking period. The overall performance of the RPC system after the LS2 commissioning period and the activities in preparation for future data taking will be presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Design and construction of the CMS Outer Tracker for the phase-2 upgrade

The High-Luminosity LHC (HL-LHC) is expected to deliver an integrated luminosity of 3000–4000 fb −1 over 10 years of operation with the peak instantaneous luminosity reaching about 5–7.5 × 1 0 34 cm −2 s −1 . During Long Shutdown 3, several components of the CMS detector will undergo major improvements, called Phase-2 upgrades, to be able to operate in the challenging environment of the HL-LHC. The current CMS tracker will be replaced. The Phase-2 Outer Tracker (OT) will have increased radiation tolerance, higher granularity, and the capability to handle higher data rates. Moreover, the OT will provide tracking information to the Level-1 trigger for the first time at a hadron collider, allowing trigger rates to be kept at a sustainable level without sacrificing physics potential. For this, the OT will be made of modules with two closely-spaced silicon sensors read out by front-end ASICs that can correlate hits in the two sensors to create short track segments, used in the Level-1 track finder. The modules come in two flavors: strip-strip and pixel-strip, containing different sensor configurations and multiple ASICs. This contribution presents the Phase-2 OT, the finalization of the OT module design, and the quality assurance and control procedures used to ensure that the modules fulfill both the specifications from the assembly steps as well as the proper communication among the ASICs.

Zoi, Irene [Fermilab] (ORCID:0000000257389446)↗

DOE OSTI2020

Impact of nuclear reactor radiation on the performance of AlN/sapphire surface acoustic wave devices

The performance of an AlN/sapphire surface acoustic wave (SAW) delay line device was characterized in real time under irradiation inside a nuclear reactor. Both its resonant frequency and transmission efficiency were observed to respond to a change in reactor power. The response follows an exponentially saturating behavior after a step power increase, followed by an exponentially decaying recovery after reactor shutdown. A sensitivity analysis based on the governing electro-mechanical equations shows that the frequency shift can be attributed to the softening of sapphire’s elastic constants under neutron radiation. Additionally, a kinetic rate equation is adopted to interpret device response and describe its microstructural evolution. These results suggest that the AlN/sapphire SAW device remains functional under irradiation, is sensitive to neutron and gamma ray fluxes, and offers an opportunity for remote sensing and in-situ measurement of material properties when exposed to nuclear reactor environment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2020

Lifetime estimation of a BWR core shroud in terms of IGSCC

The continued operation of aging Boiling Water Reactors (BWRs) worldwide requires gradually increasing cost of inspection, maintenance, and repair. Intergranular Stress Corrosion Cracking (IGSCC) in sensitized austenitic stainless steel piping first became a major issue for BWRs in the 1980s, resulting in recognition of the susceptibility of reactor internals to IGSCC. Shroud cracking identified in 1993–1994 confirmed that IGSCC of internals is a significant issue for BWRs. IGSCC is a time-dependent, material degradation process, which is caused and accelerated by the presence of residual stresses, material sensitization, irradiation, cold work, elevated temperature, and corrosive environments. This paper emphasizes the importance of accounting for corrosive environments, or more exactly, electrochemical phenomena in modelling IGSCC and predicting the service life of BWR in-vessel components, and stresses the necessity of performing such modeling not just for a single state point under full power conditions, but for the whole operating history of the reactor, including startups and shutdowns. Overall, this paper demonstrates that ignoring electrochemical considerations may result in underestimating component lifetime and lead to unnecessary expenses for inspection and repair.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Design of a prototypical natural circulation water-based reactor cavity cooling system (RCCS) for a pebble-bed generic FHR

A prototypical natural circulation water-based reactor cavity cooling system (RCCS) for the UC Berkeley Mark-1 pebble-bed gFHR is designed based on one-dimensional thermal hydraulics modeling and optimization implemented in an in-house MATLAB code. The model employs a lumped core consisting of fuel pebble and graphite pebble regions, while the graphite reflector, vessel, and RCCS are represented as separate but energy-coupled regions. The model is derived based on steady state energy balance equations accounting for conductive heat transfer from the lumped core to the vessel and subsequent radiative heat transfer to the RCCS and convective heat transfer in the water. Further, mass flow rate is calculated based on momentum and thermal energy balance in the RCCS. Starting with ANL’s water NSTF as baseline, effects of RCCS design parameters such as size of the plate between pipes, pipe diameter, source-to-sink distance, surface emissivity, and pitch from the core are examined in an integral effects framework which accounts for subsequent variation of the mass flow rate, number of pipes, convection coefficient, mean region and surface temperatures, radiative view factors, and water outlet temperature. The importance of including a prototypical reactor design in RCCS design calculations is emphasized as the calculations show that temperature drops from the core to the RCCS fluid to transfer the heat tend to be the limitation not the heat removal capacity of the fluid. A closed-loop design is obtained with capability to safely remove up to 0.72% of the nominal reactor power at maximum estimated peak conditions and 0.40% at shutdown. The design is based on physics calculations and does not account for economic optimization.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

The liquid-conduction, vapor-flow heat pipe model in Sockeye

A single-phase heat pipe flow model implemented in the heat pipe application Sockeye is described. This model solves one-dimensional, compressible flow equations for the vapor phase in the center of a heat pipe, which are coupled to the two-dimensional heat conduction equation for the wick, liquid, and cladding, as well as to an ordinary differential equation tracking the working fluid inventory in the evaporator section of the heat pipe. This model is demonstrated with several test problems, including comparisons to analytic solutions for the vapor flow fields, analytic curves for sonic and capillary limitations of heat pipe operation, and some experimental data. The numerical solution gives excellent agreement for verification problems and good agreement with experimental results. Also, demonstrations show that the model is very robust, allowing for full simulations of heat pipe transients, including frozen startup, sonic-limited (supersonic) flow, and heat pipe shutdown.

42 ENGINEERING↗

DOE OSTI2021

History and status of spent fuel treatment at the INL Fuel Conditioning Facility

Experimental Breeder Reactor II (EBR-II) had an illustrious operational history from 1964 to 1994 as one of the first sodium-cooled fast reactors to use sodium-bonded metallic driver fuels and blankets. EBR-II was a testbed for fuel development and reactor design principles. Following shutdown, it was decided to use electrometallurgical methods to treat the spent fuel and blanket materials for final disposition. Here, this paper provides a brief history of U.S. liquid metal cooled reactor development with emphasis on EBR-II, and a technical description of the electrometallurgical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Phenomena Identification and Ranking Table (PIRT) for heat pipes

Heat pipes are advanced passive thermal management devices that utilize phase change and capillary action to achieve efficient heat transfer. However, due to the complexity of the phenomena coupled in heat pipes, including capillary, phase change, turbulence, and compressibility effects, there are high uncertainties in the predictability of their operational regimes and performance. This PIRT exercise, conducted as a collaborative effort involving the Department of Energy (DOE) Microreactor Program (MRP), the Nuclear Regulatory Commission (NRC), and university partners systematically identifies, reviews, and prioritizes critical phenomena affecting the operation of heat pipes based on their importance and knowledge levels. Additional analyses and discussion are provided for phenomena with high importance and low knowledge, such as wick de-wetting, critical heat flux, contact angles, and pressure dynamics. The discussions included the recognizing challenges and proposing future research directions for both modeling and simulation and experimental efforts. Additionally, the report addresses phenomena with medium importance and low knowledge that could impact heat pipe operation during non-normal or transient operation, including frozen startup, laminar to turbulent transition, geyser boiling, wick priming, underfilling conditions, surface roughness of the wick, NCGs trapped in the wick, and the timescales of startup and shutdown. In conclusion, this comprehensive evaluation serves as a valuable resource for guiding future research and development efforts, supporting the successful integration of heat pipes into critical applications such as nuclear reactors, and contributing to the advancement of heat pipe technologies in safety-critical industries.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗