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DOE OSTI2024

Electrolyte and Cutoff Potential Effects on Cycle Life of Li4Ti5O12/LiNi0.9Mn0.1O2 Batteries for Behind-the-Meter Storage Applications

Behind-the-Meter Storage (BTMS) is a stationary battery energy storage system that is connected to the electrical distribution system on the customer's side of the utility's service meter. BTMS systems are used to store electrical energy from the grid as well as inconstant, renewable energy, such as local solar and wind generation. A successful BTMS system will allow the customer to pair their energy generation and storage to optimize electrical consumption from the grid, improving reliability and minimizing cost. For BTMS applications, batteries must be designed and optimized with different set of criteria from other leading segments of the Li-ion battery market, like transportation, due the system being stationary and proximal to the residential or commercial building it's benefitting. BTMS applications prioritize safety, cost (low/no-critical materials), reliability (20-year calendar life), and durability (10,000 cycle life), while having the ability to (minimally) compromise energy density and rate capability. Lithium titanate (Li4Ti5O12-, LTO) is a promising anode candidate for BTMS applications due to its high safety and capacity retention, while maintaining a reasonable 160 mAhg-1 reversable capacity and composition of relatively abundant materials. (1) Specifically, LTO has a high working voltage which helps to prevent Li dendrite formation, improving safety. Furthermore, LTO also has negligible lithiation-based volume change, leading to less mechanical pulverization, or loss of active material, upon cycling. For the cathode, materials with little or no Co are of high interest due to the high cost and low abundance of Co. LiMn2O4 (LMO) has been paired with LTO for BTMS applications in the past due to its safety, low cost (abundancy), and reasonably high operating voltage. (2-4) However, the low capacity of LMO limits energy density and specific energy. While not the highest priority for BTMS applications, increasing energy density will enable deployment in space constrained BTMS applications and decrease total cost. LiNi0.9Mn0.1O2 (LN-MO) is a recently developed material with promise due to its high operating voltage and relatively low price. (5) However, Ni-rich layered oxides, including LNMO, tend to struggle with capacity retention during high-voltage cycling due to mechanical pulverization, irreversible phase transitions, and unstable solid-electrolyte interphase. The study presented here focuses on building an understanding of how electrolyte solvent and varied cutoff potentials will impact the cycle life of LTO/LN-MO cells. Specifically, a comparison is provided between ethylene carbonate (EC), ethyl methyl carbonate (EMC), fluoroethylene carbonate (FEC), and Gen2 electrolyte solvents with 1M Lithium hexafluorophosphate (LiPF6) salt, cycling to two upper termination potentials, 2.6V and 2.7V. Electrochemical testing and diagnostics (e.g., differential capacity analysis, area specific impedance, constant voltage hold, and rate capability) and post-mortem characterization will be used to understand the aging behavior and failure mechanisms of the 8 cell combinations (four electrolytes and two voltage cutoffs). Cells with FEC electrolyte showed a lower initial capacity compared to cells with Gen2, EMC, and EC cycling at both voltages; however, the cells with FEC showed consistent trends in capacity retention with 2.6V and 2.7V termination potentials, while the cells with the other electrolytes showed much higher rates of capacity loss when cycling to the higher voltage. These results indicate that FEC may play a role in improving durability of high-voltage, Ni-rich electrode systems for use in high-cycle applications, such as BTMS.

electrolyte↗

DOE OSTI2026

Fuel-Cladding Eutectic Study of Legacy Fast Flux Test Facility (FFTF) MFF HT9/U-10Zr Metallic Fuel

This report presents the first systematic investigation of fuel-cladding eutectic interaction (FCEI) in irradiated HT9/U-10Zr metallic fuel from the Fast Flux Test Facility (FFTF) Materials Fuels Form (MFF) program, using differential scanning calorimetry (DSC) coupled with scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). Two irradiated fuel cross-sections, MNT07H (9.5 at% burnup, x/L = 0.78) and MNT08H (7.0 at% burnup, x/L = 0.93), were subjected to three successive isothermal annealing rounds (R1–R3) at 820°C for 20 minutes each, yielding a cumulative transient duration of one hour. This study directly addresses a recognized gap in the existing FCEI database, which previously lacked irradiated HT9/U-10Zr data at burnup levels above 8 at%. Two principal findings emerge from the study. First, for both samples, FCEI remained spatially confined within the pre-existing fuel-cladding chemical interaction (FCCI) zone boundaries after R3, with no measurable eutectic penetration into unaffected cladding beyond the original FCCI layer. This self-limiting behavior is consistent with historical Fuel Behavior Test Apparatus (FBTA) results and is attributed to the near-eutectic phase composition of the FCCI zone, which rapidly absorb the available eutectic-forming constituents and then stall penetration once the FCCI zone is consumed. A comparison with unirradiated surrogate data further supports this mechanism: whereas a U–34 at.% Fe sample would be expected to show ~176 µm of iron penetration under comparable conditions, the irradiated samples exhibited only ~20 µm, a discrepancy attributed to irradiation-induced interfacial porosity and pre-existing FCCI composition gradients. Second, for MNT08H, FCEI was observed exclusively on the half of the cladding circumference where pre-existing steady-state FCCI was present, with no detectable FCEI on the opposite half. Three hypotheses are proposed to explain this asymmetry: the inhibiting role of a zirconium-rich rind at the fuel-cladding interface; the chemical sequestration of iron by redistributed zirconium within the fuel matrix; and the persistence of fuel-cladding gaps on the FCEI-free half that preclude direct contact. All three hypotheses require further experimental investigation. The results extend the empirical FCEI database into higher-burnup territory and demonstrate the viability of DSC-based testing as a substitute for the no-longer-available FBTA apparatus. Future work will include additional cross-section testing, compilation of the full FCEI dataset, model evaluation, and DSC testing of ternary fuel compositions to broaden the experimental basis for safety assessment of sodium-cooled fast reactor systems.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Modeling of Lead-210 Data Documents Benefits in Proactively Understanding TENORM Deposition - 18515

Radioactive decay of naturally occurring Radon-222 present in just-produced natural gas can lead to deposition of its radiogenic daughter, Lead-210, and subsequent radiogenic progeny within natural gas handling, treatment, and distribution facilities. Worker safety issues and disposal challenges arise when these accumulate to radiologically problematic activity levels. This is due to the 22.3-year half-life of Lead-210, allowing for multi-decade period of accumulation to occur and to persist, even after facility operations have ceased. Due to the far shorter half-lives of Lead-210 radiogenic progeny, especially that of Polonium-210, secular equilibrium within the Lead-210-Bismuth-210-Polonium-210 decay chain is quickly approached during facility operations and achieved after facility operations end. Lead-210 that accumulates in pipe scale can complicate any subsequent intrusive facility modifications, such as disassembly and dismantlement, potentially exposing workers to adverse radioactivity and impeding recycling and/or disposal of radiologically impacted equipment or other material. Experience has shown it is not strictly the presence of Lead-210 that is problematic but rather its 'unexpected' presence when this Technologically Enhanced Naturally Occurring Radioactive Material (TENORM) is discovered during intrusive work not incorporating radiological health and safety and material handling plans. Conversely, when implementing a program that proactively considers the potential for Lead-210, this form of TENORM simply becomes another 'special waste' that can be safely, efficiently, and effectively addressed during a decommissioning effort, without associated adverse worker exposures, project cost over-runs, and schedule delays. This paper will discuss how proactive radioactive scanning quickly and safely identified Lead-210 and its radiogenic progeny within a natural gas sweetening and compressor station located in northern Montana, a portion of which was undergoing decommissioning, aiding in safe and compliant material handling and recycling. The location of this operation necessitated that the decommissioning program be completed on schedule to prevent weather-related impacts due to the onset of adverse winter conditions. Further, the radioactive data acquired during this project will be shown to provide important insight into historic behavior of Lead-210 in terms of its disposition and its more recent decay. From these data, it became possible to back-model the average activity of Radon-222 within the natural gas being processed during the life of the operation, how the estimated Radon-222 activity within the processed natural gas conformed to measured Radon-222 at various other natural gas fields, and how this information can be leveraged in developing safe and successful plans for future intrusive work at this and companion site operations. This presentation will also note the apparent preferential accumulation of Lead-210 within both rubber-based materials, such as gaskets and seals, and grease-based products found within natural gas flow lines. Thoughts regarding processes that may lead to this finding will be explored. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2012

Principles of Product Quality Control of German Radioactive Waste Forms from the Reprocessing of Spent Fuel: Vitrification, Compaction and Numerical Simulation - 12529

The German product quality control is inter alia responsible for control of two radioactive waste forms of heat generating waste: a) homogeneous vitrified HLW and b) heterogeneous compacted hulls, end-pieces and technological metallic waste. In either case, significantly different metrology is employed at the site of the conditioning plant for the obligatory nuclide inventory declaration. To facilitate an independent evaluation and checking of the accompanying documentation numerical simulations are carried out. The physical and chemical properties of radioactive waste residues are used to assess the data consistency and uncertainty margins, as well as to predict the long-term behavior of the radioactive waste. This is relevant for repository acceptance and safety considerations. Our new numerical approach follows a bottom-up simulation starting from the burn-up behavior of the fuel elements in the reactor core. The output of these burn-up calculations is then coupled with a program that simulates the material separation in the subsequent dissolution and extraction processes normalized to the mass balance. Follow-up simulations of the separated reprocessing lines of a) the vitrification of highly-active liquid and b) the compaction of residual intermediate-active metallic hulls remaining after fuel pellets dissolution, end-pieces and technological waste, allows calculating expectation values for the various repository relevant properties of either waste stream. The principles of the German product quality control of radioactive waste residues from the spent fuel reprocessing have been introduced and explained. Namely, heat generating homogeneous vitrified HLW and heterogeneous compacted metallic MLW have been discussed. The advantages of a complementary numerical property simulation have been made clear and examples of benefits are presented. We have compiled a new program suite to calculate the physical and radio-chemical properties of common nuclear waste residues. The immediate benefit is the independent assessment of radio-active inventory declarations and much facilitated product quality control of waste residues that need to be returned to Germany and submitted to a German HLW-repository requirements. Wherever possible, internationally accepted standard programs are used and embedded. The innovative coupling of burn-up calculations (SCALE) with neutron and gamma transport codes (MCPN-X) allows an application in the world of virtual waste properties. If-then-else scenarios of hypothetical waste material compositions and distributions provide valuable information of long term nuclide property propagation under repository conditions over a very long time span. Benchmarking the program with real residue data demonstrates the power and remarkable accuracy of this numerical approach, boosting the reliability of the confidence aforementioned numerous applications, namely the proof tool set for on-the-spot production quality checking and data evaluation and independent verification. Moreover, using the numerical bottom-up approach helps to avoid the accumulation of fake activities that may gradually build up in a repository from the so-called conservative or penalizing nuclide inventory declarations. The radioactive waste properties and the hydrolytic and chemical stability can be predicted. The interaction with invasive chemicals can be assessed and propagation scenarios can be developed from reliable and sound data and HLW properties. Hence, the appropriate design of a future HLW repository can be based upon predictable and quality assured waste characteristics. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Modeling of Water Film Formation on a Stay-Cable

With the increasing popularity of cable-stayed bridges in the past few decades, longer spans are being constructed using ever-longer stay cables. Under certain conditions, especially during construction, and wind and rain-wind events, the cables of cable-stayed bridges may undergo large amplitude vibrations. This type of vibration has been of concern to researchers and engineers because large amplitude vibration can induce excessive stresses and therefore threaten the safety and integrity of bridge structures. Figure 1-1 presents an example of a stay cable bridge structure, The Cooper River Bridge. Many countermeasures for the wind-rain induced vibrations have been developed to decrease the amplitude of the vibrations and consequently, the fatigue of the cables. One of the methods is to install helical fillets on the cable surface. A sketch of a cable with two fillets wrapped around it is shown in Figure 1-2. The main objective of the present study is to develop and test computational methodologies, with the use of computational fluid dynamics (CFD), that model the interaction of rain with stay cables including: the formation of water film, development of rainwater rivulets, and rivulet behavior under various conditions with and without wind. Interaction of water with cables with and without fillets is analyzed, and results are presented in this report. This study is divided into two parts. The first part of this study, Section 4, explores methods of modeling water film formation and flow on cables. A series of simulations using the Reynolds averaged Navier-Stokes equations turbulence model (RANS) and large eddy simulation (LES) model are performed to analyze water film formation and flow on a cable using the fluid film model in STARCCM+ [(b)1]. Various methods of setting up initial and boundary conditions are tested, especially boundary conditions for introducing rain into the domain. The second part of the study, Section 5, explores upper rivulet formation in the presence of wind, based on the findings from the first part of the study. In that section, LES simulation with the Volume Of Fluid model for the free surface is used. Also, various results for cases with and without water film on the cable are presented and compared.

42 ENGINEERING↗

DOE OSTI2012

Measurement of interfacial structures in horizontal air-water bubbly flows

In order to predict multi-dimensional phenomena in nuclear reactor systems, methods relying on computational fluid dynamics (CFD) codes are essential. However, to be applicable in assessing thermal-hydraulic safety, these codes must be able to accurately predict the development of two-phase flows. Therefore, before practical application these codes must be assessed using experimental databases that capture multi-dimensional phenomena. While a large database exists that can be employed to assess predictions in vertical flows, the available database for horizontal flows is significantly lacking. Therefore, the current work seeks to develop an additional database in air-water horizontal bubbly flow through a 38.1 mm ID test section with a total development length of approximately 250 diameters. The experimental conditions are chosen to cover a wide range of the bubbly flow regime based upon flow visualization using a high-speed video camera. A database of local time-averaged void fraction, bubble velocity, interfacial area concentration, and bubble Sauter mean diameter are acquired throughout the pipe cross-section using a four-sensor conductivity probe. To investigate the evolution of the flow, measurements are made at axial locations of 44, 116, and 244 diameters downstream of the inlet. In the current work, only measurements obtained at L/D = 244 are presented. It is found that increasing the liquid superficial velocity tends to reduce both the bubble size and the degree of bubble packing near the upper wall. However, it is observed that the position of the maximum void fraction value remains nearly constant and is located approximately one bubble diameter away from the upper wall. It is also found that the bubble velocity exhibits a power law behavior resembling a single phase liquid turbulent velocity profile. Moreover, the local bubble velocity tends to decrease as the local void fraction increases. Conversely, increasing the gas superficial velocity is found to have little effect on the bubble size or velocity, but increases the bubble packing near the upper pipe wall. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2010

The role of science supporting the Waste Isolation Pilot Plant.

The presentation briefly addresses three topics. First, science has played an important role throughout the history of the WIPP project, beginning with site selection in the middle 1970s. Science was a key part of site characterization in the 1980s, providing basic information on geology, hydrology, geochemisty, and the mechanical behavior of the salt, among other topics. Science programs also made significant contributions to facility design, specifically in the area of shaft seal design and testing. By the middle 1990s, emphasis shifted from site characterization to regulatory evaluations, and the science program provided one of the essential bases for certification by the Environmental Protection Agency in 1998. Current science activities support ongoing disposal operations and regulatory recertification evaluations mandated by the EPA. Second, the EPA regulatory standards for long-term performance frame the scientific evaluations that provide the basis for certification. Unlike long-term dose standards applied to Yucca Mountain and proposed repositories in other nations, the WIPP regulations focused on cumulative releases during a fixed time interval of 10,000 years, and placed a high emphasis on the consequences of future inadvertent drilling intrusions into the repository. Close attention to the details of the regulatory requirements facilitated EPA's review of the DOE's 1996 Compliance Certification Application. Third, the scientific understanding developed for WIPP provided the basis for modeling studies that evaluated the long-term performance of the repository in the context of regulatory requirements. These performance assessment analyses formed a critical part of the demonstration that the site met the specific regulatory requirements as well as providing insight into the overall understanding of the long-term performance of the system. The presentation concludes with observations on the role of science in the process of developing a disposal system, including the importance of establishing the regulatory framework, building confidence in the long-term safety of the system, and the critical role of the regulator in decision making.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

High Temperature Fluoride Salt Test Loop

Effective high-temperature thermal energy exchange and delivery at temperatures over 600°C has the potential of significant impact by reducing both the capital and operating cost of energy conversion and transport systems. It is one of the key technologies necessary for efficient hydrogen production and could potentially enhance efficiencies of high-temperature solar systems. Today, there are no standard commercially available high-performance heat transfer fluids above 600°C. High pressures associated with water and gaseous coolants (such as helium) at elevated temperatures impose limiting design conditions for the materials in most energy systems. Liquid salts offer high-temperature capabilities at low vapor pressures, good heat transport properties, and reasonable costs and are therefore leading candidate fluids for next-generation energy production. Liquid-fluoride-salt-cooled, graphite-moderated reactors, referred to as Fluoride Salt Reactors (FHRs), are specifically designed to exploit the excellent heat transfer properties of liquid fluoride salts while maximizing their thermal efficiency and minimizing cost. The FHR s outstanding heat transfer properties, combined with its fully passive safety, make this reactor the most technologically desirable nuclear power reactor class for next-generation energy production. Multiple FHR designs are presently being considered. These range from the Pebble Bed Advanced High Temperature Reactor (PB-AHTR) [1] design originally developed by UC-Berkeley to the Small Advanced High-Temperature Reactor (SmAHTR) and the large scale FHR both being developed at ORNL [2]. The value of high-temperature, molten-salt-cooled reactors is also recognized internationally, and Czechoslovakia, France, India, and China all have salt-cooled reactor development under way. The liquid salt experiment presently being developed uses the PB-AHTR as its focus. One core design of the PB-AHTR features multiple 20 cm diameter, 3.2 m long fuel channels with 3 cm diameter graphite-based fuel pebbles slowly circulating up through the core. Molten salt coolant (FLiBe) at 700°C flows concurrently (at significantly higher velocity) with the pebbles and is used to remove heat generated in the reactor core (approximately 1280 W/pebble), and supply it to a power conversion system. Refueling equipment continuously sorts spent fuel pebbles and replaces spent or damaged pebbles with fresh fuel. By combining greater or fewer numbers of pebble channel assemblies, multiple reactor designs with varying power levels can be offered. The PB-AHTR design is discussed in detail in Reference [1] and is shown schematically in Fig. 1. Fig. 1. PB-AHTR concept (drawing taken from Peterson et al., Design and Development of the Modular PB-AHTR Proceedings of ICApp 08). Pebble behavior within the core is a key issue in proving the viability of this concept. This includes understanding the behavior of the pebbles thermally, hydraulically, and mechanically (quantifying pebble wear characteristics, flow channel wear, etc). The experiment being developed is an initial step in characterizing the pebble behavior under realistic PB-AHTR operating conditions. It focuses on thermal and hydraulic behavior of a static pebble bed using a convective salt loop to provide prototypic fluid conditions to the bed, and a unique inductive heating technique to provide prototypic heating in the pebbles. The facility design is sufficiently versatile to allow a variety of other experimentation to be performed in the future. The facility can accommodate testing of scaled reactor components or sub-components such as flow diodes, salt-to-salt heat exchangers, and improved pump designs as well as testing of refueling equipment, high temperature instrumentation, and other reactor core designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2013

Opportunities and challenges for structural health monitoring of radioactive waste systems and structures

Radioactive waste systems and structures (RWSS) are safety-critical facilities in need of monitoring over prolonged periods of time. Structural health monitoring (SHM) is an emerging technology that aims at monitoring the state of a structure through the use of networks of permanently mounted sensors. SHM technologies have been developed primarily within the aerospace and civil engineering communities. This paper addresses the issue of transitioning the SHM concept to the monitoring of RWSS and evaluates the opportunities and challenges associated with this process. Guided wave SHM technologies utilizing structurally-mounted piezoelectric wafer active sensors (PWAS) have a wide range of applications based on both propagating-wave and standing-wave methodologies. Hence, opportunities exist for transitioning these SHM technologies into RWSS monitoring. However, there exist certain special operational conditions specific to RWSS such as: radiation field, caustic environments, marine environments, and chemical, mechanical and thermal stressors. In order to address the high discharge of used nuclear fuel (UNF) and the limited space in the storage pools the U.S. the Department of Energy (DOE) has adopted a 'Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste' (January 2013). This strategy endorses the key principles that underpin the Blue Ribbon Commission's on America's Nuclear Future recommendations to develop a sustainable program for deploying an integrated system capable of transporting, storing, and disposing of UNF and high-level radioactive waste from civilian nuclear power generation, defense, national security, and other activities. This will require research to develop monitoring, diagnosis, and prognosis tools that can aid to establish a strong technical basis for extended storage and transportation of UNF. Monitoring of such structures is critical for assuring the safety and security of the nation's spent nuclear fuel until a national policy for closure of the nuclear fuel cycle is defined and implemented. In addition, such tools can provide invaluable and timely information for verification of the predicted mechanical performance of RWSS (e.g. concrete or steel barriers) during off-normal occurrence and accident events such as the tsunami and earthquake event that affected Fukushima Daiichi nuclear power plant. The ability to verify the conditions, health, and degradation behavior of RWSS over time by applying nondestructive testing (NDT) as well as development of nondestructive evaluation (NDE) tools for new degradation processes will become challenging. The paper discusses some of the challenges associated to verification and diagnosis for RWSS and identifies SHM technologies which are more readily available for transitioning into RWSS applications. Fundamental research objectives that should be considered for the transition of SHM technologies (e.g., radiation hardened piezoelectric materials) for RWSS applications are discussed. The paper ends with summary, conclusions, and suggestions for further work. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS2014

Application of Fault Management Theory to the Quantitative Selection of a Launch Vehicle Abort Trigger Suite

The theory of System Health Management (SHM) and of its operational subset Fault Management (FM) states that FM is implemented as a "meta" control loop, known as an FM Control Loop (FMCL). The FMCL detects that all or part of a system is now failed, or in the future will fail (that is, cannot be controlled within acceptable limits to achieve its objectives), and takes a control action (a response) to return the system to a controllable state. In terms of control theory, the effectiveness of each FMCL is estimated based on its ability to correctly estimate the system state, and on the speed of its response to the current or impending failure effects. This paper describes how this theory has been successfully applied on the National Aeronautics and Space Administration's (NASA) Space Launch System (SLS) Program to quantitatively estimate the effectiveness of proposed abort triggers so as to select the most effective suite to protect the astronauts from catastrophic failure of the SLS. The premise behind this process is to be able to quantitatively provide the value versus risk trade‐off for any given abort trigger, allowing decision makers to make more informed decisions. All current and planned crewed launch vehicles have some form of vehicle health management system integrated with an emergency launch abort system to ensure crew safety. While the design can vary, the underlying principle is the same: detect imminent catastrophic vehicle failure, initiate launch abort, and extract the crew to safety. Abort triggers are the detection mechanisms that identify that a catastrophic launch vehicle failure is occurring or is imminent and cause the initiation of a notification to the crew vehicle that the escape system must be activated. While ensuring that the abort triggers provide this function, designers must also ensure that the abort triggers do not signal that a catastrophic failure is imminent when in fact the launch vehicle can successfully achieve orbit. That is, the abort triggers must have low false negative rates to be sure that real crew‐threatening failures are detected, and also low false positive rates to ensure that the crew does not abort from non‐crew‐threatening launch vehicle behaviors. The analysis process described in this paper is a compilation of over six years of lessons learned and refinements from experiences developing abort triggers for NASA's Constellation Program (Ares I Project) and the SLS Program, as well as the simultaneous development of SHM/FM theory. The paper will describe the abort analysis concepts and process, developed in conjunction with SLS Safety and Mission Assurance (S&MA) to define a common set of mission phase, failure scenario, and Loss of Mission Environment (LOME) combinations upon which the SLS Loss of Mission (LOM) Probabilistic Risk Assessment (PRA) models are built. This abort analysis also requires strong coordination with the Multi‐Purpose Crew Vehicle (MPCV) and SLS Structures and Environments (STE) to formulate a series of abortability tables that encapsulate explosion dynamics over the ascent mission phase. The design and assessment of abort conditions and triggers to estimate their Loss of Crew (LOC) Benefits also requires in‐depth integration with other groups, including Avionics, Guidance, Navigation and Control(GN&C), the Crew Office, Mission Operations, and Ground Systems. The outputs of this analysis are a critical input to SLS S&MA's LOC PRA models. The process described here may well be the first full quantitative application of SHM/FM theory to the selection of a sensor suite for any aerospace system.

Lo, Yunnhon↗

NASA NTRS2008

Visualization Component of Vehicle Health Decision Support System

The visualization front-end of a Decision Support System (DSS) also includes an analysis engine linked to vehicle telemetry, and a database of learned models for known behaviors. Because the display is graphical rather than text-based, the summarization it provides has a greater information density on one screen for evaluation by a flight controller.This tool provides a system-level visualization of the state of a vehicle, and drill-down capability for more details and interfaces to separate analysis algorithms and sensor data streams. The system-level view is a 3D rendering of the vehicle, with sensors represented as icons, tied to appropriate positions within the vehicle body and colored to indicate sensor state (e.g., normal, warning, anomalous state, etc.). The sensor data is received via an Information Sharing Protocol (ISP) client that connects to an external server for real-time telemetry. Users can interactively pan, zoom, and rotate this 3D view, as well as select sensors for a detail plot of the associated time series data. Subsets of the plotted data can be selected and sent to an external analysis engine to either search for a similar time series in an historical database, or to detect anomalous events. The system overview and plotting capabilities are completely general in that they can be applied to any vehicle instrumented with a collection of sensors. This visualization component can interface with the ISP for data streams used by NASA s Mission Control Center at Johnson Space Center. In addition, it can connect to, and display results from, separate analysis engine components that identify anomalies or that search for past instances of similar behavior. This software supports NASA's Software, Intelligent Systems, and Modeling element in the Exploration Systems Research and Technology Program by augmenting the capability of human flight controllers to make correct decisions, thus increasing safety and reliability. It was designed specifically as a tool for NASA's flight controllers to monitor the International Space Station and a future Crew Exploration Vehicle.

Jacob, Joseph↗

DOE OSTI2022

Regulatory Treatment of Non-Core Sources of Radioactivity for Advanced Reactor Designs

The recent resurgence in advance (non-light water) reactor development has been paralleled by the development of risk-informed performance-based (RIPB) licensing pathways. Specifically, the creation of the RIPB Licensing Modernization Project (LMP) approach and subsequent endorsement by the U.S. Nuclear Regulatory Commission (NRC) now provides advanced reactor vendors with a defined RIPB method to develop an affirmative safety case for licensing. In addition, the Technology Inclusive Content of Applications Project (TICAP) has published guidance on developing a license application based on the LMP approach. To support the utilization of risk information as part of advanced reactor design and licensing efforts, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Joint Committee on Nuclear Risk Management (JCNRM) has developed a probabilistic risk assessment (PRA) standard for advanced reactors. The standard, which was formerly approved by the American National Standards Institute (ANSI) in 2021 and recently endorsed by the NRC in trial use Regulatory Guide (RG) 1.247, is an integral standard, covering from initiating events to offsite consequence. A major feature of the standard is that it permits the inclusion of any source of radioactivity material at the site within the plant PRA. Therefore, non-core sources of radioactivity, such as fuel storage, fuel processing, and purification systems, can be included within a single comprehensive plant PRA. For those advanced reactor vendors utilizing a RIPB licensing approach, there is an opportunity to include the non-core sources of radioactivity within the RIPB framework for licensing decision-making, such as the categorization of events, classification of structures, systems, and components (SSCs), and evaluation of the adequacy of defense-in-depth (DID). For advanced reactor designs that contain multiple non-core sources of radioactivity, or for monolithic plant sites that include associated fuel facilities, this approach could potentially simplify licensing applications through the use of a single, uniform, and consistent decision-making framework across all radioactive sources at the site. In addition, a RIPB approach could provide additional insights regarding plant behavior, flexibility regarding licensing decision-making, and potentially allow the use of risk information as part of the plant oversight process. Risk-informing these aspects of advanced reactor licensing would also be consistent with the NRC’s risk policy statement. However, there is diverse regulation and guidance regarding the licensing of non-core sources of radioactivity and generally limited experience using RIPB approaches for the evaluation as part of licensing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

Identifying the speciation of salt-based actinides in the presence of contaminants [Slides]

Molten salt reactors (MSR) present advantages over light water reactors, such as higher safety and energy efficiency, convenient waste processing, and the ability to use more abundant thorium instead of uranium as the fuel source. However, due to the high temperatures associated with these reactors the reactor container corrosion product can have a high influence on the molten salt system, and they are considered as a part of the fuel salt system. It is essential to have a comprehensive understanding of the chemical reactions that are occurring in the molten salt in the presence of contaminants such as nickel, manganese, chromium chlorides and oxides. Because they can change the local structure of these salts and the local structures of these salts govern the thermophysical properties of the molten salts, which would determine the safety and operational parameters of the reactor. This study is focused on identifying f-element materials speciation resulting from reaction with corrosion/degradation products in a molten salt environment to understand the flow of MSR. In this work, f-element chlorides are mixed with alkali and alkali earth metals and corrosion products such as transition metal chlorides are introduced systematically, inside a glovebox. Then these are heated to around 700 ? and slowly cooled to room temperature. Afterwards, these are analyzed using different characterization techniques such as powder X-ray diffraction, UV-vis and Raman spectroscopy and solid-state NMR. The initial work was conducted with lanthanide chlorides as a surrogate for actinides and the acquired data strongly indicates that in the presence of corrosion products new phase formation/ change in coordination environments occurs. This work has been presented at multiple conference presentations. The proposed work would allow to extend this work to actinides (depleted uranium and thorium) and this would assist to identify the speciation of actinide chloride in the presence of corrosion products. Also, this work would open an opportunity to compare the coordination behavior of lanthanides with the actinides.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗

DOE OSTI2012

Numerical Analysis of JNES Seismic Tests on Degraded Combined Piping System

Nuclear power plant safety under seismic conditions is an important consideration. The piping systems may have some defects caused by fatigue, stress corrosion cracking, etc., in aged plants. These cracks may not only affect the seismic response but also grow and break through causing loss of coolant. Therefore, an evaluation method needs to be developed to predict crack growth behavior under seismic excitation. This paper describes efforts conducted to analyze and better understand a series of degraded pipe tests under seismic loading that was conducted by Japan Nuclear Energy Safety Organization (JNES). A special 'cracked-pipe element' (CPE) concept, where the element represented the global moment-rotation response due to the crack, was developed. This approach was developed to significantly simplify the dynamic finite element analysis in fracture mechanics fields. In this paper, model validation was conducted by comparisons with a series of pipe tests with circumferential through-wall and surface cracks under different excitation conditions. These analyses showed that reasonably accurate predictions could be made using the abaqus connector element to model the complete transition of a circumferential surface crack to a through-wall crack under cyclic dynamic loading. The JNES primary loop recirculation piping test was analyzed in detail. This combined-component test had three crack locations and multiple applied simulated seismic block loadings. Comparisons were also made between the ABAQUS finite element (FE) analyses results to the measured displacements in the experiment. Good agreement was obtained, and it was confirmed that the simplified modeling is applicable to a seismic analysis for a cracked pipe on the basis of fracture mechanics. Pipe system leakage did occur in the JNES tests. The analytical predictions using the CPE approach did not predict leakage, suggesting that cyclic ductile tearing with large-scale plasticity was not the crack growth mode for the acceleration excitations considered here. Hence, the leakage was caused by low-cycle fatigue with small-scale yielding. The procedure used to make predictions of low-cycle fatigue crack growth with small-scale yielding was based on the Dowling {Delta}J procedure, which is an extension of linear-elastic fatigue crack growth methodology into the nonlinear plasticity region. The predicted moments from the CPE approach were used using a cycle-by-cycle crack growth procedure. The predictions compare quite well with the experimental measurements.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2010

Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

NASA NTRS2008

Reliability and Confidence Interval Analysis of a CMC Turbine Stator Vane

High temperature ceramic matrix composites (CMC) are being explored as viable candidate materials for hot section gas turbine components. These advanced composites can potentially lead to reduced weight, enable higher operating temperatures requiring less cooling and thus leading to increased engine efficiencies. However, these materials are brittle and show degradation with time at high operating temperatures due to creep as well as cyclic mechanical and thermal loads. In addition, these materials are heterogeneous in their make-up and various factors affect their properties in a specific design environment. Most of these advanced composites involve two- and three-dimensional fiber architectures and require a complex multi-step high temperature processing. Since there are uncertainties associated with each of these in addition to the variability in the constituent material properties, the observed behavior of composite materials exhibits scatter. Traditional material failure analyses employing a deterministic approach, where failure is assumed to occur when some allowable stress level or equivalent stress is exceeded, are not adequate for brittle material component design. Such phenomenological failure theories are reasonably successful when applied to ductile materials such as metals. Analysis of failure in structural components is governed by the observed scatter in strength, stiffness and loading conditions. In such situations, statistical design approaches must be used. Accounting for these phenomena requires a change in philosophy on the design engineer s part that leads to a reduced focus on the use of safety factors in favor of reliability analyses. The reliability approach demands that the design engineer must tolerate a finite risk of unacceptable performance. This risk of unacceptable performance is identified as a component's probability of failure (or alternatively, component reliability). The primary concern of the engineer is minimizing this risk in an economical manner. The methods to accurately determine the service life of an engine component with associated variability have become increasingly difficult. This results, in part, from the complex missions which are now routinely considered during the design process. These missions include large variations of multi-axial stresses and temperatures experienced by critical engine parts. There is a need for a convenient design tool that can accommodate various loading conditions induced by engine operating environments, and material data with their associated uncertainties to estimate the minimum predicted life of a structural component. A probabilistic composite micromechanics technique in combination with woven composite micromechanics, structural analysis and Fast Probability Integration (FPI) techniques has been used to evaluate the maximum stress and its probabilistic distribution in a CMC turbine stator vane. Furthermore, input variables causing scatter are identified and ranked based upon their sensitivity magnitude. Since the measured data for the ceramic matrix composite properties is very limited, obtaining a probabilistic distribution with their corresponding parameters is difficult. In case of limited data, confidence bounds are essential to quantify the uncertainty associated with the distribution. Usually 90 and 95% confidence intervals are computed for material properties. Failure properties are then computed with the confidence bounds. Best estimates and the confidence bounds on the best estimate of the cumulative probability function for R-S (strength - stress) are plotted. The methodologies and the results from these analyses will be discussed in the presentation.

Murthy, Pappu L. N.↗

DOE OSTI2024

Multiphysics Demonstration of Temperature-Driven Assembly Bowing in SFRs using MOOSE-Based Codes

Core bowing is an important passive safety mechanism in liquid metal cooled fast reactors. When the core restraint system is properly designed, temperature and flux gradients influence assemblies in the core to bow into less reactive configurations during accident scenarios, resulting in negative reactivity feedback. Prediction of core bowing involves complex interplay of radiation transport, impacts of fluid flow and heat transfer on duct temperature, and mechanical responses to the induced temperature and flux gradients. Under the U.S. Department of Energy Office of Nuclear Energy’s Advanced Modeling and Simulation (NEAMS) Program [1], an integrated multiphysics approach is being developed to model the core bowing phenomena in liquid metal-cooled fast reactors with the Multiphysics Object Oriented Simulation Environment (MOOSE) [2]. In this methodology, the MOOSE-based reactor physics code Griffin [3] will solve the neutron transport equation and determine the power distribution. With the detailed power distribution from Griffin, the subchannel analysis codes MOOSE-Subchannel [4] and Pronghorn [5] are utilized to calculate the assembly temperature distribution. MOOSE’s Solid Mechanics [6] and Contact [7] Modules are leveraged to calculate the thermal expansion and duct bowing displacement with the duct wall temperature from thermal hydraulics calculation. In this work, an initial one-way coupling demonstration of the integrated multiphysics approach has been performed on a seven-assembly problem based on the sodium-cooled fast reactor ABR-1000 design [8]. The neutronics calculation with Griffin is not yet involved in the current simulation. MOOSE-Subchannel and Pronghorn evaluate fluid and solid temperature based on a fixed power distribution. In addition, one-way coupling is utilized in this coupled calculation, via Pronghorn passing the duct temperature data to the MOOSE Solid Mechanics calculation. An assessment of the Solid Mechanics module was performed in parallel to verify duct bowing behavior with duct-to-duct contact phenomenon [9]. The displacement from MOOSE Solid Mechanics is not yet transferred back and utilized in the Pronghorn and MOOSE-Subchannel calculation. This model will be available on the National Reactor Innovation Center (NRIC) Virtual Test Bed (VTB) repository [10]. Future stages of this work will involve solving problems of increasing complexity as well as adding more physics (e.g. reactor physics) to the integrated workflow to reach the end goal of modeling the core bowing phenomenon with an integrated multiphysics workflow.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS2004

Physics in NASA Exploration

The primary focus of the workshop was NASA's new concentration on sending crewed missions to the Moon by 2020, and then on to Mars and beyond. Several speakers, including JPL s Fred O Callaghan and NASA's Mark Lee, broached the problem that there is now a serious reduction of capability to perform experiments in the ISS, or to fly significant mass in microgravity by other means. By 2010, the shuttle fleet will be discontinued and Russian craft will provide the only access to the ISS. O Callaghan stated that the Fundamental Physics budget is being reduced by 70%. LTMPF and LCAP are slated for termination. However, ground-based experiments are continuing to be funded at present, and it will be possible to compete for $80-90 million in new money from the Human Research Initiative (HRI). The new program thrust is for exploration, not fundamental physics. Fundamental, we were told by Lee, does not ring well in Washington these days. Investigators were advised to consider how their work can benefit missions to the Moon and Mars. Work such as that regarding atomic clocks is looked upon with favor, for example, because it is considered important to navigation and planetary GPS. Mark Lee stressed that physicists must convey to NASA senior management that they are able and willing to contribute to the new exploration research programs. The new mentality must be we deliver products, not do research. This program needs to be able to say that it is doing at least 50% exploration-related research. JPL s Ulf Israelsson discussed the implications to OBPR, which will deliver methods and technology to assure human health and performance in extraterrestrial settings. The enterprise will provide advanced life-support systems and technology that are reliable, capable, simpler, less massive, smaller, and energy-efficient, and it may offer other necessary expertise in areas such as low-gravity behavior. Like Dr. Lee, he stated that the focus must be on products, not research. While there is not yet a formal direction, he said, LTMPF and PARCS ISS flight projects are slated to terminate in October 2004. All flight investigations are being returned to ground programs and phased out by the end of FY07. Physics ground programs are intact for now, but to survive we must shift about 50% of research to supporting exploration. Basic research programs in other disciplines are being cancelled. Product lines will support human health, safety and life-support, including countermeasures against radiation and other hazards, as well as advances in time-keeping, navigation and communications technologies. Israelsson said that the new Fundamental Physics for Exploration Roadmap points to how fundamental physics research can and does support exploration. JPL will use the roadmap to argue for support for fundamental physics research under several codes. Nicholas Bigelow of the University of Rochester encouraged attendees not to become discouraged, but rather to embrace the opportunities presented by NASA's new direction.

O'Callaghan, Fred↗