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At least 1,279 records · Page 71

Radically Engineered Modular Air Separation System with Tailored Oxygen Sorbents

The commercial energy sector relies heavily on fossil fuel conversion, and in the process releases a significant amount of CO 2 . A promising technology to utilize fossil fuels with relatively affordable CO 2 capture is gasification. However, it requires a pure oxygen stream. The current state of the art method to produce oxygen is cryogenic air separation, which supercools air to a liquid, and then using distillation columns to separate the components. While this method has been thoroughly studied, it only has a 25% efficiency from a second law standpoint and therefore requires a significant amount of energy (and associated emissions) for oxygen production. This, then, lowers the incentive for carbon capture within a plant, so alternative methods need to be investigated. One potential method, chemical looping air separation (CLAS), is a promising method to replace state of the art oxygen generation technologies. CLAS utilizes a cyclic redox scheme with an oxygen sorbent to create pure oxygen streams. This approach typically utilizes a dual reactor scheme where the oxygen deficient sorbent enters the first reactor and is subjected to high oxygen partial pressures to re-oxidize the sorbent. Then the sorbent is sent to the reducing reactor, where it is subjected to low oxygen partial pressure (steam or vacuum) to releases oxygen. The overarching objective of this project was to discover the principles for rational design and optimization of oxygen sorbents and process design to ensure the process is a viable replacement for cryogenic air separation, especially in the context of modular gasification systems. This was done through development, characterization, testing, and analyses of (a) high temperature mixed composite oxides; (b) low temperature doped perovskite oxides (A1 x A2 1-x B1 y B2 1-y O 3 ); (c) scale up synthesis and testing of the optimized sorbent particles; (d) process design and analyses of the CLAS system in the context of modular gasification applications.

01 COAL, LIGNITE, AND PEAT↗

Improved Fracture Models for Relocation Modeling

The BISON fuel performance code is being developed to provide a modern tool that has the flexibility to analyze a wide variety of fuel forms and to model conditions and phenomena that could not be represented in legacy tools. There are a number of motivations for this, including providing support for development of advanced fuel with improved accident tolerance for existing light water (LWR) reactors, improved understanding of mechanisms in fuel designs in current use in a wider variety of conditions, and facilitating the development of fuel for advanced reactor designs. To accomplish these goals, it is clear that BISON must rely on models of fuel behavior that are based on fundamental physical behavior, rather than on empirical correlations that represent that behavior in a simplified fashion. BISON still does employ many empirical models that were originally developed for other fuel performance codes, but efforts are underway to replace these with models that are more physically based. Radial relocation in LWR fuel is an example of a phenomenon that is currently represented by an empirical model, but which is ripe for replacement by physically based models. During normal operation, ceramic LWR fuel experiences significant fracturing that is driven by spatially nonuniform volumetric expansion. This occurs due to the significant thermal gradients within the fuel that occur in fresh and irradiated fuel, as well as nonuniform swelling due to fission products that occurs over longer-term irradiation. Fracture and fragmentation of fuel allows the outer radius of the fuel pellet to expand due to the loss of mechanical constraint and outward radial migration of fragments. This radial expansion has a significant effect on the fuel system response because it decreases the size of the gap between the fuel and cladding. This decreases the thermal resistance across that gap, which leads to decreased fuel centerline temperatures. This report documents recent work to improve the ability of BISON to model radial relocation using the extended finite element method (XFEM). These enhancements include the ability to include cohesive zone models, improved code architecture for propagating cracks based on fracture integrals, and improved handling of fuel and cladding interfaces with XFEM. These capabilities are demonstrated on a simulation of radial relocation including a surface interaction model that enforces residual opening.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Protecting and Defending against Autonomous Control Systems and Digital Twin Cyber Attacks: Response Strategy for Hyperparameter attacks of Digital Twin Machine Learning Models in Nuclear Power Plants (Final)

Navigating through the complex tapestry of technological advancements, "Response Strategy for Hyperparameter attacks of Digital Twin Machine Learning Model in Nuclear Power Plants" stands at the intersection of cybersecurity and nuclear power plant operations, embarking on a journey through the intricacies of securing digital twins against malicious cyber activities. As nuclear power plants progressively integrate digital twin technology and machine learning models to optimize operations and ensure system reliability, they inadvertently expose themselves to a new spectrum of vulnerabilities, notably in the realm of hyperparameter attacks. Hyperparameters, integral in machine learning model tuning and optimal performance of digital twins, have emerged as a target for adversaries aiming to destabilize the predictive capabilities and therefore, the operational accuracy of these digital entities within critical infrastructures like nuclear plants. This paper, therefore, meticulously threads the needle through the development of a robust response strategy, poised to shield these digital reflections against calculated hyperparameter manipulations, ensuring that the digital twin can effectively and securely function as a reliable proxy for its physical counterpart. The ensuing sections delve into the orchestrated maelstrom of multi-rate time-changing intelligent coordinated hyperparameter attacks and the implementation of event-triggered predictive control, laying down a structured, predictive, and responsive framework that safeguards the nexus where the digital and physical realms of nuclear power plants coalesce. The operational integrity of digital twins in nuclear power plants depends critically on the security of machine learning hyperparameters. This study makes two different contributions. First, a decision-based idea known as a multi-rate time changing intelligent coordinated hyperparameter attack is put forth. In this attack, many hyperparameters are repeatedly changed using both random and intelligent optimal techniques by the attacker. These assaults introduce varied rates at different attack steps, compromise various amounts of hyperparameters, and improve stealth and flexibility. Second, a technique is developed for event triggered predictive control to rapidly respond to potential hyperparameter attacks. This control integrates a sliding window framework, retaining a history of previous data points and employing linear regression to predict the next data point from the current dataset. The control gain K is determined using the Lyapunov-Krasovskii method, and subsequently, an action is developed. Finally, the outcome of the simulation demonstrates the viability of the proposed method for defending nuclear power plant digital twins from hyperparameter attacks.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

THERMAL SAFETY MARGIN CALCULATION OF THE MP-2 EXPERIMENT IN THE ADVANCED TEST REACTOR

The Mini-Plate 2 (MP-2) irradiation test is a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). The experiment is a drop-in test where small aluminum-clad fuel plate samples (mini plates) are cooled directly by the ATR Primary Coolant System (PCS) water. The MP-2 fuel plate experiment will be irradiated in several different irradiation locations of the ATR. This fueled experiment contains aluminum-clad fuel mini plates consisting of monolithic U-Mo. Four different types of fuel plates with fuel meat thickness and cladding are part of the MP-2 test. A thermal analysis has been performed on the MP-2 experiment. A method for calculating Departure from Nucleate Boiling Ratio (DNBR) and Flow Instability Ratio (FIR) during a reactivity transient using the commercial finite element and heat transfer code ABAQUS is discussed. At the start of an ATR cycle the heat generation rate of the fueled experiment is high and the heat rate multiplier from the outer shim control cylinders is low, while the reverse is true at the end of the ATR cycle. Thermal analyses at 10 day increments during the cycle calculate the DNBR and FIR during a reactivity transient. This technique calculates DNBR for the fuel plate surfaces and FIR for all water components for each finite element surface and node at various times during the ATR cycle. Heat rates vary with time during the transient calculation that are provided by a detailed physics analysis. Oxide growth on the fuel plates is also incorporated Results from the transient calculations are displayed with the ABAQUS post processor. By calculating these parameters at each location in the finite element model, conservatism is replaced with accuracy. This allows for a greater margin for the thermal hydraulic safety parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Innovative test method for the estimation of the foaming tendency of substrates for biogas plants

Graphical abstract: Display Omitted - Highlights: • Foaming in biogas plants depends on the interactions between substrate and digestate. • Foaming tests enable the evaluation of substrate foaming tendency in biogas plants. • Leipzig foam tester enables foaming tests of substrates prior to use. - Abstract: Excessive foaming in anaerobic digestion occurs at many biogas plants and can cause problems including plugged gas pipes. Unfortunately, the majority of biogas plant operators are unable to identify the causes of foaming in their biogas reactor. The occurrence of foaming is often related to the chemical composition of substrates fed to the reactor. The consistency of the digestate itself is also a crucial part of the foam formation process. Thus, no specific recommendations concerning substrates can be given in order to prevent foam formation in biogas plants. The safest way to avoid foaming is to test the foaming tendency of substrates on-site. A possible solution is offered by an innovative foaming test. With the help of this tool, biogas plant operators can evaluate the foaming disposition of new substrates prior to use in order to adjust the composition of substrate mixes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

High-resolution Gamma Spectroscopy Measurements of Pressurized Water Reactor Spent Nuclear Fuel Rods

International nuclear safeguards use nondestructive assay techniques to verify declared characteristics of spent nuclear fuel (SNF) assemblies. The declared characteristics usually include the initial enrichment, assembly average burnup, and irradiation history. The primary nondestructive assay tool used by the International Atomic Energy Agency has been the fork detector irradiated fuel measurement system, which uses gross gamma and neutron correlations; however, there have been several research publications within the last decade on the use of high-resolution gamma spectroscopy for the determination of SNF characteristics. References [1]–[6] are published journal articles on this subject from 2006–2016. This report describes measurements of 25 individual SNF rods using high-resolution gamma spectroscopy. Details of the measurement geometry, measurement settings, and rod characteristics are documented. These spectra are intended for the validation of simulation codes that calculate SNF inventory and source term characteristics. The spectrum archive is available to evaluate future spectrum analysis methods. The archive contains full axial scan spectra of each SNF rod. This allows more detailed comparison between simulation and nondestructive analysis than is possible on assemblies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Short-baseline electron antineutrino disappearance study by using neutrino sources from {sup 13}C + {sup 9}Be reaction

To investigate the existence of sterile neutrino, we propose a new neutrino production method using {sup 13}C beams and a {sup 9}Be target for short-baseline electron antineutrino (ν-bar {sub e} ) disappearance study. The production of secondary unstable isotopes which can emit neutrinos from the {sup 13}C + {sup 9}Be reaction is calculated with three different nucleus-nucleus (AA) reaction models. Different isotope yields are obtained using these models, but the results of the neutrino flux are found to have unanimous similarities. This feature gives an opportunity to study neutrino oscillation through shape analysis. In this work, expected neutrino flux and event rates are discussed in detail through intensive simulation of the light ion collision reaction and the neutrino flux from the beta decay of unstable isotopes followed by this collision. Together with the reactor and accelerator anomalies, the present proposed ν-bar {sub e} source is shown to be a practically alternative test of the existence of the Δ m {sup 2} ∼ 1 eV{sup 2} scale sterile neutrino.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Application of Shift Ex-Core Calculations to a Detailed MSR Model

This report documents the collaborative research conducted between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) on applying the Monte Carlo neutronics code Shift to calculate the radiation dose rates for a detailed molten salt reactor (MSR) model based on the Molten Salt Breeder Reactor (MSBR). In this work scope, the ANL and ORNL teams worked together and applied Shift to calculate the radiological environments of this detailed MSR facility modeled with explicit geometries. The radiological conditions within the MSR facility were modeled for when the reactor is at two different operation modes: normal full power and drained state. The FW-CADIS hybrid method in Shift was applied successfully to calculate the ex-core neutron and gamma dose rates for the MSBR at full power. Dose rate maps showed that in the current MSBR numerical model, potential pathways exist for neutrons and gammas to stream through the 8-foot concrete shield to reach the top of the MSBR reactor cell. Neutron and gamma dose rates within the drain cell were also calculated for the MSBR at the drained state by integrating the source terms obtained from an ORIGEN-S depletion calculation into the Shift simulation. The results indicate that in the drained state, the delayed gammas from the depleted fuel salt are the main contributors to the dose rates, which suggests that a steel liner in the tank model must be added for calculating the dose rates in the drain cell.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Thermonuclear neutron emission from a sheared-flow stabilized Z-pinch

We report the fusion Z-pinch experiment (FuZE) is a sheared-flow stabilized Z-pinch designed to study the effects of flow stabilization on deuterium plasmas with densities and temperatures high enough to drive nuclear fusion reactions. Results from FuZE show high pinch currents and neutron emission durations thousands of times longer than instability growth times. While these results are consistent with thermonuclear neutron emission, energetically resolved neutron measurements are a stronger constraint on the origin of the fusion production. This stems from the strong anisotropy in energy created in beam-target fusion, compared to the relatively isotropic emission in thermonuclear fusion. In dense Z-pinch plasmas, a potential and undesirable cause of beam-target fusion reactions is the presence of fast-growing, “sausage” instabilities. This work introduces a new method for characterizing beam instabilities by recording individual neutron interactions in plastic scintillator detectors positioned at two different angles around the device chamber. Histograms of the pulse-integral spectra from the two locations are compared using detailed Monte Carlo simulations. These models infer the deuteron beam energy based on differences in the measured neutron spectra at the two angles, thereby discriminating beam-target from thermonuclear production. An analysis of neutron emission profiles from FuZE precludes the presence of deuteron beams with energies greater than 4.65 keV with a statistical uncertainty of 4.15 keV and a systematic uncertainty of 0.53 keV. This analysis demonstrates that axial, beam-target fusion reactions are not the dominant source of neutron emission from FuZE. These data are promising for scaling FuZE up to fusion reactor conditions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Development of a Discrepancy Checker for the Digital Twin in a Supervisory Control System for a Thermal Energy Delivery System

Defined as a virtual representation of a physical object, process, or service, and used to support real-world decision-making, a digital twin (DT) can be utilized to combine classical and novel frameworks in sensors, state predictions, and multi-input/multi-output systems, and to enable optimal autonomous operations. However, a DT’s usefulness largely depends on its ability to adequately mirror the state of its physical counterpart, and this adequacy should be reflected by the level of uncertainty in the underlying simulation models when estimating and predicting quantities of interest (QOIs). Moreover, simulation models in a DT may involve multiple fidelities of representations—ranging from physics-based models to data-driven ones—but classical uncertainty quantification (UQ) methods struggle to handle numerous uncertainty sources, nor are they designed for real-time applications. This work presents a UQ-based discrepancy checking and diagnosis tool for a DT-based supervisory control system applied to a thermal energy delivery system (TEDS) at Idaho National Laboratory. The discrepancy checker was developed using metadata from an automated DT development process, and these metadata included different combinations of physical model forms and model parameters, training data and hyperparameters for surrogate models, and design parameters for supervisory control systems. Next, correlations between the uncertainty results and the metadata were established and then applied to the DT operations. The discrepancy checker evaluates the discrepancies between model predictions from virtual and sensor measurements and backtraces them to the corresponding major sources of uncertainty. The discrepancy checker showed reasonable performance in detecting discrepancies and diagnosing sources of uncertainty in testing scenarios.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

Theoretical Calculation and Simulation Studies of Axisymmetric Plasma and Halo Current Analysis in EAST

A new technology for developing fusion energy is to use hydrogen isotopes i.e., deuterium (D) and tritium (T) which is a combine effort for building up of International Thermonuclear Reactor (ITER) named as Tokamak, which will come into operation in 2020. Handsome amount of work has already been done by many researchers contingent with plasma shape, halo current and plasma equilibrium properties by numerical techniques. Halo current calculation is the most important work in fusion reactors. In this paper, we have calculated halo current and plasma cross-section energy. The research covers calculation of plasma energy and determines the force on the vessel by magnetic field transformation method, which is the essential part of calculating halo current. Further, we have developed VDE up/down halo current models. These models give total magnetic field, asymmetric forces, conducting points and poloidal halo current. Confined plasma by magnetic field is considered to be an abundant and durable source of energy for future aspects. In this paper, plasma magnetic field of each cross section was calculated by using energy integral. Transformation method has been successfully applied which provides the connection between major sideways forces acting on the plasma. Theoretical investigation and analyses of halo current have successfully been done. Two observations were analyzed, large ratio of halo current during VDE down and cancellation of specific forces at conducting point. The results obtained will be useful for equalizing the simulation and experimental results. The research will be extended to developed new models for future fusion research. (authors)

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

PROTEUS-SN User Manual

PROTEUS-SN is a three-dimensional, highly scalable, high-fidelity neutron transport code developed at Argonne National Laboratory. The code is applicable to all spectrum reactor transport calculations, particularly those in which a high degree of fidelity is needed either to represent spatial detail or to resolve solution gradients. PROTEUS-SN solves the second order formulation of the transport equation using the continuous Galerkin finite element method in space, the discrete ordinates approximation in angle, and the multigroup approximation in energy. PROTEUS-SN’s parallel methodology permits the efficient decomposition of the problem by both space and angle, permitting large problems to run efficiently on hundreds of thousands of cores. PROTEUS-SN can also be used in serial or on smaller compute clusters (10’s to 100’s of cores) for smaller homogenized problems, although it is generally more computationally expensive than traditional homogenized methodology codes. PROTEUS-SN has been used to model partially homogenized systems, where regions of interest are represented explicitly and other regions are homogenized to reduce the problem size and required computational resources. PROTEUS-SN solves forward and adjoint eigenvalue problems and permits both neutron upscattering and downscattering. An adiabatic kinetics option has recently been included for performing simple time-dependent calculations in addition to standard steady state calculations. PROTEUS-SN handles void and reflective boundary conditions. Multigroup cross sections can be generated externally using the MC2-3 fast reactor multigroup cross section generation code or internally using the cross section application programming interface (API) which can treat the subgroup or resonance table libraries. PROTEUS-SN is written in Fortran 90 and also includes C preprocessor definitions. The code links against the PETSc, METIS, HDF5, and MPICH libraries. It optionally links against the MOAB library and is a part of the SHARP multi-physics suite for coupled multi-physics analysis of nuclear reactors. This user manual describes how to set up a neutron transport simulation with the PROTEUS-SN code. A companion methodology manual describes the theory and algorithms within PROTEUS-SN.

97 MATHEMATICS AND COMPUTING↗

Fabrication of neutron absorbing metal hydride entrained ceramic matrix shield composites

With significant improvement in High Temperature Superconductors (HTS), several projects are adopting HTS technology for fusion power systems. Compact HTS tokamaks offer potential advantages including lower plant costs, enhanced plasma control, and ultimately lower cost of electricity. However, as compact reactors have a reduced radial build to accommodate shielding, HTS degradation due to radiation damage or heating is a significant and potentially design limiting issue. Shielding must mitigate threats to the superconducting coils: neutron cascade damage, heat deposition and potentially organic insulator damage due x-rays. Unfortunately, there are currently no hi-performance shielding materials to enable the potential performance enhancement offered by HTS. In this work, we present a manufacturing method to fabricate a new class of composite shields that are high performance, high operating temperature, and simultaneously neutron absorbing and neutron moderating. The composite design consists of an entrained metal-hydride phase within a radiation stable MgO ceramic host matrix. We discuss the fabrication, characterization, and thermophysical performance data for a series of down-selected composite materials inspired by future fusion core designs and their operational performance metrics. To our knowledge these materials represent the first ceramic composite shield materials containing significant metal hydrides.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Bridging the simulator gap: Measuring motivational bias in digital nuclear power plant environments

Using digital NPP simulator facilities for research in Human Reliability Analysis (HRA) and nuclear power research is increasingly popular. For this work, we propose a method for characterizing and quantifying the gap between data collected in a simulator and data that reflect NPP operations. Using novice operators, we demonstrate how to manipulate and measure the impact of the simulator environment on operator actions. A set of biases are proposed to characterize factors introduced by the simulator environment. There are two categories of simulator bias: Environmental Biases (physical differences between the simulator and the control room), and Motivational Biases (cognitive differences between training in a simulator and operating a NPP). This study examines Motivational Bias. A preliminary causal model of Motivational Biases is introduced and tested in a demonstration experiment using 30 student operators. Data from 41 simulator sessions are analyzed. Data include crew characteristics, operator surveys, and time to recognize and diagnose the accident in the scenario. Quantitative models of the Motivational Biases using Structural Equation Modeling (SEM) are proposed. With these models, we estimate how the effects of the scenario conditions are mediated by simulator bias, and we demonstrate how to quantify the strength of these effects.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Performance Improvements of the Griffin Solvers in FY24

The Griffin code is a MOOSE-based reactor physics application jointly developed by Idaho National Laboratory and Argonne National Laboratory under the Department of Energy Office of Nuclear Energy Nuclear Energy Advanced Modeling and Simulation Program. This fiscal year, we have made significant efforts to improve the performance of transport solver options and cross-section generation for the efficient use of Griffin in advanced reactor applications. For the HFEM-PN solver, the residual evaluations of HFEM kernels were optimized by utilizing the pre- computed averaged cross sections for individual elements. Numerical integration involving the evaluation of basis functions at quadrature points was bypassed by facilitating precomputed element mass matrices for response matrices. Red-black iterations were improved by introducing a new generalized minimum residual based solver. The memory usage of response matrix storage was significantly reduced by applying basis function rotations on interfaces and calculating volumetric odd-parity moments on the fly. Additionally, the adjoint flux and transient calculation capabilities of the HFEM-PN solver were successfully implemented and verified using the TWIGL benchmark problem. For the DFEM-SN solver, memory footprint and computation time were significantly reduced by not treating angular flux vectors as the MOOSE nonlinear system vectors. Specifically for IQS, scalar adjoint weighting was introduced to further eliminate angular adjoint flux storage in the MOOSE auxiliary system. It was demonstrated through the three-dimensional Advanced Burner Test Reactor core problem that the memory usage for transient calculations with the IQS method was reduced by over 7.5× compared to before the optimizations. For the self-shielding application programming interface, a new double-heterogeneity treatment method, named the Bell Function-Based Analytic Two-Region Slowing Down Method, was developed to efficiently flux-volume homogenize TRISO particles with the matrix. Additionally, optimizations were made to hyper- fine group (HFG) slowing down calculations by pretabulating collision probability coefficients and grouping isotopes, significantly reducing the computational time for calculating scattering sources per HFG. Lastly, the pin power reconstruction module was extended to account for temporal behavior in a microreactor analysis problem, specifically for a control drum transient. Verification tests for each of these improvements demonstrated significant performance enhancements and memory reduction.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of Analysis Methods that Integrate Numeric and Textual Equipment Reliability Data

Within the Light Water Reactor Sustainability (LWRS) program, the Risk-Informed Systems Analysis (RISA) Pathway is performing collaborative research on the development and deployment of technologies designed to assist operating nuclear power plants (NPPs) to reduce operating costs improve plant reliability and availability. One of the RISA research areas is focusing on the development of methods and tools designed to optimize plant operations (e.g., maintenance/replacement schedules, optimal maintenance postures for plant structures, systems, and components [SSCs]) in a manner that is more cost effective than current approaches and makes better use of available SSC health data. The Risk-Informed Asset Management (RIAM) project targets this research area by creating a direct bridge between component equipment reliability (ER) data and system engineer decision making regarding maintenance activity scheduling and component aging management. In this respect, one challenge that NPP system engineers are facing is that the amount of ER data being continuously generated is not only extremely large in size, but it comes in different forms: textual (e.g., condition or maintenance reports) and numeric (e.g., generated by monitoring systems). All these data elements provide them with valuable insights and information regarding: 1) the discovery of anomalous behaviors or degradation trends, 2) the identification of the possible causes behind such behaviors/trends, and 3) the prediction of their direct consequences. However, several challenges have proved to be roadblocks to this process. While some of these challenges are technical in nature (i.e., data are often distributed over several physical servers/databases), others are conceptual in nature: data elements come in different formats (e.g., numeric or textual), and measured values have different scales (e.g., vibration spectra and oil temperature). The activities performed by the RIAM project during FY23 directly tackles the need to simultaneously integrate the analysis of ER data in all its forms, numeric and textual. Note that such task has never been performed before due to the complexity of the systems under consideration but, most importantly, because of the technical challenges behind the harmonization of ER data formats and the lack of adequate computational methods to analyze them. Our approach borrows ideas and concepts from the medical field where integration of several data sources is vital to assist medical practitioners to perform correct diagnosis and indicate optimal treatments. In our view a NPP asset is equivalent to a patient in a medical context. The main difference is the complexity of a human body is a magnitude more complex when compared to typical assets commonly present in NPPs (e.g., centrifugal pumps, or motor operated valves). This simplifies our first requirement when analyzing heterogenous ER data formats: to put data into “context”. Context is here intended as the additional piece of information that is needed by ER data analysis tools to understand what these data elements are referring to, i.e., which king of knowledge they are generating. In our context, this knowledge can be translated into models that capture the form and functional architecture of assets/systems, their dependencies, and how they interact. These models actually emulate the knowledge that that NPP system engineers possess about assets and systems; this is their key of success when analyzing ER data, their challenge is ability to handle large amount of data. Here, we employ model-based system engineering (MBSE) models of systems and assets to represent and capture their architecture and functional, i.e. cause-effect, relations. Then, ER data elements are processed by identifying first of all which elements of the developed MBSE elements they are referring to. For numeric ER data this task is fairly easy since it is possible to precisely pinpoint what MBSE elements the corresponding sensor are observing (e.g., bearing temperature of a centrifugal pump). Task is much harder for textual data since the information contained in issue or maintenance reports needs to “be understood” by a computational tool. Here we called this process as “knowledge extraction”. Once again, we borrow the experience in the medical field where methods to extract knowledge from textual data have been developed in the past decade. The missing element for us is the availability of a complete dictionary of NPP related concepts (in addition to the MBSE models presented earlier) that can put “text into context”. In FY23, such dictionary has been developed along with all the computational elements required for knowledge extraction. Lastly, once numeric and textual ER data elements have been processed and “understood”, then the last step is the discovery of possible cause-effect relations among them. This is performed by observing if a logical connection through the MBSE models exists, and if the

97 MATHEMATICS AND COMPUTING↗

Electrochemical Salt Wasteform Development: A Review of Salt Treatment and Immobilization Options

Electrochemical reprocessing, also referred to as pyroprocessing, is a technique for recycling actinides from used nuclear fuel (UNF) to produce fuel for future reactors. Here, UNF is dissolved in a molten salt (e.g., LiCl-KCl eutectic) within an electrorefiner. After UNF dissolution, fission products are released into the electrolyte salt and converted to chlorides. This paper discusses wasteform options for processing the base electrolyte salt with the fission product salts as well as just the rare-earth fission products (as RECl 3 , REOCl, or REO x ) with the intent of finding optimal methods for reducing total waste salt volumes or partitioning the salt for alternate wasteform options. Furthermore, two of the more detailed partitioning options discussed herein include halide removal from the salt (dehalogenation), which accounts for more than half of the salt on a molar basis, and RE fission product removal for wasteforms with high-RE loadings. Wasteform properties are compared with emphasis on wasteform volume starting from a given amount of (1) total salt cations or (2) RE cations. Comparisons are also made of wasteform chemical durabilities, with the data available from like testing methods. A main conclusion from this work is the justification of subsequent salt processing after electrorefiner operations for achieving significant wasteform volume reduction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Revealing Fine Structure in the Antineutrino Spectra from a Nuclear Reactor

Here, we calculate the Inverse Beta Decay (IBD) antineutrino spectrum generated by nuclear reactors using the summation method to understand deviations from the smooth Huber-Mueller model due to the decay of individual fission products, showing that plotting the ratio of two adjacent spectra points can effectively reveal these deviations. We obtain that for binning energies of 0.1 MeV or lower, abrupt changes in the spectra due to the jagged nature of the individual antineutrino spectra could be observed for highly precise experiments. Surprisingly, our calculations also reveal a peak-like feature in the adjacent points ratio plot at 4.5 MeV even with a 0.25 MeV binning interval, which we find is present in the IBD spectrum published by Daya Bay in 2016. We show that this 4.5 MeV feature is caused by the contributions of just four fission products, 95 Y, 98,101 Nb and 102 T c . This would be the first evidence of the decay of a few fission products in the IBD antineutrino spectrum from a nuclear reactor. This result is supported by applying the same numerical technique to measured aggregate electron spectra.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗