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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,279 records · Page 71
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

A New Layer to the PRA: Operational Performance Risk Assessment

A supervisory control system (SCS) concept developed by Oak Ridge National Laboratory provides automated risk-informed decision-making capabilities with the ultimate goal of significantly boosting plant availability while reducing dependency on human resources for multi-module advanced reactors. The decision-making process integrates online-monitoring systems, and the associated diagnostic and prognostics tools to continuously take into account the component health in an integrated probabilistic/deterministic meta model to capture the system behavior during a select set of postulated anticipated operational occurrence scenarios. The operational performance risk assessment (OPRA) approach is introduced to support SCS decisions probabilistically to minimize/prevent unnecessary trips and challenges to plant safety systems. OPRA identifies and ranks the success paths, combination of the non-safety systems and components, with the real-time failure data using event tree/fault tree method. OPRA does not interfere with the safety systems and adds a buffer zone to the existing probabilistic risk assessment (PRA) domain by minimizing the possible transients. In this work, failure scenarios of feedwater and turbine control valves are analyzed for ALMR PRISM (Power Reactor Innovative Small Module) design at full power. As an alternative to the default shutdown option, five and seven success paths (full and reduced power operation) are defined respectively within the OPRA framework.

Guler Yigitoglu, Askin↗
DOE OSTI2017

OSIRIS User's Manual, v. 1.0

The On-Site Inspection RadioIsotopic Spectroscopy (OSIRIS) system is a gamma-ray spectroscopy instrument for on-site inspections under the Comprehensive Nuclear-Test-Ban Treaty (CTBT). [1] OSIRIS identifies fission-product isotopes with a high-resolution gamma-ray spectrometer. The spectrometer is controlled by a ruggedized notebook computer running data acquisition and data analysis software customized for CTBT on-site inspections. This manual explains OSIRIS preparation and setup for operation; startup procedures, including energy-calibration of the spectrometer; spectral measurements and spectral analyses; and shutdown procedures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Benchmark Progress towards TREAT M8CAL Core Loading to Support Validation of TREAT Operations

Benchmark experiment data is necessary to validate modeling and simulation activities to support Transient Reactor Experiment Test (TREAT) Facility restart and operations. Key measurements of interest include criticality, control rod worth, excess reactivity, and shutdown margin for varying core loading sizes. Benchmark evaluations are being developed according the guidelines provided in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook) that can also support advanced modeling and simulation activities, new experiment design for advanced reactors and accident tolerant fuel, and potential low-enriched uranium conversion of the TREAT reactor. This summary discusses development activities supporting a benchmark evaluation of the TREAT M8CAL core loading, which will be submitted for peer-review and publication in the IRPhEP Handbook with the identifier TREAT-FUND-RESR-002. Detailed models of TREAT were developed and evaluated to prepare benchmark models of the minimum critical mass configuration. Calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Discussion of results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program

Reactivation of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program The Transient Reactor Test (TREAT) Facility is a 19,000 MW reactor located at the Idaho National Laboratory designed to perform transient testing on nuclear fuel experiments and materials. A neutron radiography facility was installed with the first radiograph being produced in 1967 that utilized the indirect (or transfer) method of neutron radiography to examine irradiated nuclear fuels and experiment test loops. The TREAT neutron radiography facility performed approximately 5,000 radiographs by the spring of 1977. The original radiography facility was upgraded in 1974-1975 to correct previous deficiencies and to improve capabilities. In 1986, an upgraded aperture was installed to increase the resolution of images followed by a major reactor upgrade completed in 1988. Overall, the existing neutron radiography system was in good working order. Originally built in 1958, the TREAT Facility was in operation until it was placed in a shutdown status in 1994. Following the Fukishima disaster and seeing a need for enhanced accident tolerant fuels, the United Stated Department of Energy decided to restart the TREAT facility and resume transient operations. In November 2017, the TREAT reactor was successfully restarted and is currently performing operational testing in preparation for initial experiment irradiations and transient testing. This paper reports on efforts to reactivate the TREAT neutron radiography station. A few minor shielding and system improvement upgrades have been completed along with the installation of a dedicated room for computed radiography equipment. Detailed beam characterization work was not performed following the last reactor upgrade. To characterize the neutron beam, gold foil activation measurements were made to determine an average neutron flux and flux profile. Additional multi-foil activation measurements will be made to calculate the energy spectrum. An open beam image will be taken that will provide the information about variations in the beam profile. A series of system qualification shots will also take place to accurately determine the effective image quality and image acquisition parameters. Full operational status of the TREAT neutron radiography system is scheduled for late summer of 2018.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Identification of biomass fractions responsible for wear on biomass processing equipment

Abstract: There is a current problem in the DOE supported Integrated Bio refineries where metal parts, including grinder blades and pneumatic ducting, are wearing out too early in the lifetime of the equipment. Repairing and replacing the equipment leads to unexpected equipment costs and the complete shutdown of the affected production lines.. It is believed that the ash content of the biomass is the likely cause of the wear on the metals. To address this ongoing problem, INL staff are analyzing the wear of metals caused by interactions with biomass . A sandblaster was modified to shoot ground biomass at metal coupons to simulate the wear that is occurring in the bio refineries. This system was able to test several variables including which biomass causes erosive and abrasive wear, impingement angle, particle densities of the biomass, and how the quantity of mass impacts the wear of the metals. Coupon mass loss after blasting was used to evaluate the wear caused by different biomass fractions. It was determined that the bark fraction of biomass caused the most wear and the whitewood fraction caused the least wear. Coupon mass loss in logging residues was found to be positively correlated with ash content, indicating that the ash content in the biomass is responsible for wear. It was also found that hand separated forest residues, sodium, aluminum and iron were positively correlated with wear, while silicon showed no correlation with wear.

09 BIOMASS FUELS↗
DOE OSTI2018

The CMS Event-Builder System for LHC Run 3 (2021-23)

The data acquisition system (DAQ) of the CMS experiment at the CERN Large Hadron Collider (LHC) assembles events of 2MB at a rate of 100 kHz. The event builder collects event fragments from about 750 sources and assembles them into complete events which are then handed to the highlevel trigger (HLT) processes running on O(1000) computers. The aging eventbuilding hardware will be replaced during the long shutdown 2 of the LHC taking place in 2019/20. The future data networks will be based on 100 Gb/s interconnects using Ethernet and Infiniband technologies. More powerful computers may allow to combine the currently separate functionality of the readout and builder units into a single I/O processor handling simultaneously 100 Gb/s of input and output trac. It might be beneficial to preprocess data originating from specific detector parts or regions before handling it to generic HLT processors. Therefore, we will investigate how specialized coprocessors, e.g. GPUs, could be integrated into the event builder. We will present the envisioned changes to the event-builder compared to today’s system. Initial measurements of the performance of the data networks under the event-building trac pattern will be shown. Implications of a folded network architecture for the event building and corresponding changes to the software implementation will be discussed. Corresponding author: remigius.mommsen@cern.c

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2019

System Thermal-Hydraulic Modeling of the Phénix Dissymmetric Test Benchmark

Phénix is a French pool-type sodium-cooled prototype reactor; before the definitive shutdown, occurred in 2009, a final set of experimental tests are carried out in order to increase the knowledge on the operation and the safety aspect of the pool-type liquid metal-cooled reactors. One of the experiments was the Dissymmetric End-of-Life Test which was selected for the validation benchmark activity in the frame of SESAME project. The computer code validation plays a key role in the safety assessment of the innovative nuclear reactors and the Phénix dissymmetric test provides useful experimental data to verify the computer codes capability in the asymmetric thermal-hydraulic behaviour into a pool-type liquid metal-cooled reactor. This paper shows the comparison of the outcomes obtained with six different System Thermal-Hydraulic (STH) codes: RELAP5-3D©, SPECTRA, ATHLET, SAS4A/SASSYS-1, ASTEC-Na and CATHARE. The nodalization scheme of the reactor was individually achieved by the participants; during the development of the thermal-hydraulic model, the pool nodalization methodology had a special attention in order to investigate the capability of the STH codes to reproduce the dissymmetric effects which occurs in each loop and into pools, caused by the azimuthal asymmetry of the boundary conditions. The modelling methodology of the participants is discussed and the main results are compared in this paper to obtain useful guide lines for the future modelling of innovative liquid metal pool-type reactors.

Narcisi, V.↗
DOE OSTI2015

Isotope Ratio Method Analysis of the WWR-SM Reactor

In this paper, we use the Isotope Ratio Method (IRM) to estimate the total energy production of the Uzbekistan Institute of Nuclear Physics WWR-SM research reactor. The isotope ratio method is a technique for verifying the operational history of a nuclear reactor by measuring trace element isotope ratios in non-fuel reactor components. The isotope ratios in these components can then be directly related to cumulative energy production with standard reactor modeling methods. This method has been successfully demonstrated on several shutdown reactors in the past. The WWR-SM is a 10MW pool reactor that has been operating since 1959. In 2013, a Pacific Northwest National Laboratory (PNNL) team successfully removed a surface sample from the aluminum core shroud during a refueling outage. The sample was subjected to Secondary Ionization Mass Spectrometry (SIMS) analysis at PNNL in 2014. Our results show good agreement with the actual operating history of the reactor.

Nuclear archeology, isotope ratio method, nonproli↗
DOE OSTI2018

Advanced Polymer-Modified Cements for Subsurface Applications

Wellbore failure at the cement lining is one of the most common drivers of reservoir intervention during geothermal energy and unconventional oil/gas production. The main sources for cement fracturing are chemical, mechanical, and thermal stresses. As a result, expensive and time-intensive production shutdowns and repairs are required. Intervention costs average $1.5million per wellbore without taking into consideration the economic losses as a result of production stoppage, which can be several million dollars depending on the time frame of plant in non-production mode. To address these problems, we developed a thermally stable polymer-cement composite with self-healing properties while maintaining the required rheological (during pumping) and mechanical properties of typical wellbore cement. In this work we report on the composition and mineralogy, the mechanical properties and self-healing ability of these novel cement composites as compared to standard wellbore cement. The self-healing capability of the composite materials was demonstrated by the consistent permeability reduction of mechanically-induced fractures in the 0.3-0.5mm aperture range. These polymer-cement composite materials could represent a definite solution to wellbore failure, production stoppage and reservoir intervention during geothermal energy production.

Fernandez, Carlos A.↗
DOE OSTI2019

On Xenon Fission Product Poisoning

We all appreciate the special role Xenon-135 plays in reactor transients. Unless there is sufficient reactivity override, on shutdown a conventional reactor cannot immediately restart because of the buildup of strongly absorbing Xenon from Iodine decay and the absence of neutron flux to continue its elimination. We are also aware of how the flux imbalance in a large reactor core gives rise to Xenon induced power precession around the core in unpredictable ways. In addition, with regard to computational methods, we lack readily accessible numerical algorithms enabling reliable characterization of Xenon oscillations with simple models to explain this fundamentally important phenomenon. Nowhere is this more apparent than in the classroom. In this presentation, we first consider a new 12-element comprehensive decay to model Xenon and its corresponding numerical implementation and verification to establish a predictive model of Xenon buildup in a constant flux. We then follow with a reduced self-consistent flux model in an infinite reactor- a model not found in textbooks. This presentation should appeal to those interested in developing high order numerical methods for reactor transients.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Identifying Effective Signals to Predict Deleted and Suspended Accounts on Twitter across Languages

Social networks have an ephemerality to them where accounts and messages are constantly being edited, deleted, or marked as private. This continuous change comes from concerns around privacy, a potential desire for deception, and spam-like behavior. In this study we analyze multiple large datasets of thousands of active and deleted Twitter accounts to produce a series of predictive features for the removal or shutdown of an account. We have selected these accounts from speakers of three languages -- Russian, Spanish, and English to evaluate if speakers of various languages behave differently with regards to deleting accounts. We find that unlike previously used profile and network features, the discourse of deleted vs. active accounts forms the basis for highly accurate account deletion prediction. More precisely, we observed that the presence of a certain set of terms in user tweets leads to a higher likelihood for that user's account deletion. We show that the predictive power of profile, language, affect, and network features is not consistent across speakers of the three evaluated languages.

social media, machine learning, natural language p↗
DOE OSTI2020

Conceptual Design Study of a Thermal Energy Storage System Using Granular Solids for Coal-fired Boilers

<span style="font-size: 12pt;">Presentation for the 2020 International Pittsburgh Coal Conference held virtually from September 8-11, 2020.</span><span style="font-size: 12pt;">Abstract: As electricity from various renewable energy sources is increasingly supplied to the power grid, the grid’s stability will be challenged. Proposed grid-stability solutions will need to address variability of sources, such as solar and wind energy, in order to alleviate problems such as overgeneration and ramping. In regions where increases in solar energy are prominent, such as California, the net power demand curve often referred to as “the Duck Curve,” has caused problems for energy dispatchers, due to over-generation and fast ramping demands. In this project, we investigated a conceptual design for a Thermal Energy Storage System (TESS) integrated into a coal-fired power plant to support grid-scale storage. Implementing a TESS into power plants can allow for an increase in renewable sources, increase grid resilience, and allow conventional coal-fired plants to operate more efficiently. For this project, a TESS using granular solids as the energy-storage media is integrated into a conventional coal-fired power plant. The purpose of the TESS is to store energy from the flue gas during low demand periods and transfer the energy to steam during peak demand periods. The storage allows the stabilization of combustion and related boiler components at low loads and provides the opportunity to avoid some shutdowns. This study uses thermodynamic properties of several candidate thermal storage materials to estimate the size of storage system required for several operating scenarios. </span>

Huckaby, E. David↗
DOE OSTI2021

Interoffice Memorandum RE-04-21 "Advanced Test Reactor Power History Through Cycle 169A-1"

Table 1 lists the Advanced Test Reactor (ATR) N-16 constrained power history data since the Beryllium VI Core Internals Changeout (CIC) Cycle 134A-1 through Cycle 169A-1. The powers tabulated for Cycles 159A-1, 163A-1, and 167A-1 are higher than powers observed during operation, because powers are computed from cycle exposure, which for these cycles includes operation during the low-power “soak” portion of each cycle; whereas cycle length in Effective Full Power Days (EFPD) includes only the high-power “casualty” portion occurring after NF is established. Table 2 lists the accumulated N-16 lobe and total core exposure, as obtained from the ATR Data Acquisition System (RDAS) for Cycles 134A-1 through 169A-1. Table 3 lists the startup and shutdown dates and times, as obtained from logbooks or RDAS, depending on availability. The ATR power history prior to Cycle 134A 1 is presented in references (a) through (d).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Advanced Fuels Campaign Execution Plan

The Advanced Fuels Campaign (AFC) Execution Plan outlines the strategy, mission, scope, near-term and long-term goals, structure, and organization associated with nuclear fuels and materials research, development, and demonstration activities within the Department of Energy’s (DOE) Nuclear Fuel Cycle and Supply Chain (NFCSC) program. NFCSC has been given responsibility to identify and mature advanced fuel technologies for the DOE using a science-based approach, focused on developing a fundamental understanding of nuclear fuels and materials to drive development of integrated nuclear fuel and materials technology. This science-based approach combines theory, experiments, and multiscale modeling and simulation to achieve a predictive understanding of relevant behaviors ranging from fuel fabrication processes (and their resulting fuel microstructures) through fuel/cladding performance under irradiation (in contrast to more empirical, observation-based approaches frequently used in fuel performance modeling and fuel qualification). The traditional scope of AFC includes the evaluation and development of multiple fuel forms to support two fuel cycle options: once-through and full recycle. The word “fuel” is used generically to include conventional fuels, transmutation targets, and any associated cladding or duct materials. The once-through fuel cycle addresses advanced light water reactor fuels with enhanced performance, extended burnup, and reduced waste generation. In fiscal year (FY) 2012, AFC’s scope expanded to include research, development, and demonstration (RD&D) for light water reactor (LWR) fuels with enhanced accident tolerance. Fuel fabrication activities include the development of innovative methods to enhance process efficiencies, reduce waste, and improve control over as-fabricated fuel microstructural properties to achieve desired in-reactor performance. Using modern modeling and simulation approaches, the objective is to predict fresh fuel properties given the feedstock characteristics and fabrication process parameters. The performance-related activities include small-scale, in-reactor, and out-of-reactor phenomenological testing (distinct from, but synergistic with, integral prototypic testing) and extensive, quantitative characterization (focusing on characterization of fuel and cladding materials at the scale of microstructure) both before and after testing. Larger-scale, prototypic experiments are conducted in concert with phenomenological testing to drive a Fuel Development and Qualification program, incorporating a fundamental understanding of fuel behavior performance characteristics. Then, using the tools developed under the productive science-based approach, fuels will be optimized to meet specific performance requirements, thereby minimizing the need to repeatedly perform large-scale, integral experiments over a wide parametric range as a means of experimental exploration. Two significant initiatives are underway within AFC. First, a gap analysis completed in early FY 2019 identified critical irradiation testing needs that are lacking within the national light water reactor (LWR) fuels testbed since the shutdown of the Halden Reactor in 2018. The identified gaps are for instrumented, prototypic testing of LWR fuels, especially under boiling water reactor conditions, ramp conditions, and conditions leading to fuel failure; these needs exist for supporting current LWR fuels and their possible extension to higher burnups, but are especially urgent relative to near-term development and qualification of accident-tolerant fuels. Recommendations that resulted from the Halden Gap Analysis focused on enhancements at Advanced Test Reactor (ATR) and Transient Reactor Test Facility (TREAT) to fill gaps in testing capabilities relative to these needs. Second, a concerted effort to develop and demonstrate a systematic approach to accelerating the development, testing, and qualification of new fuel systems has been initiated. This is highlighted by a test strategy that combines the considerable advances in multiscale, mechanistic fuel modeling of recent years with a MiniFuel separate effects test program in the High Flux Isotope Reactor (HFIR) and a Fission Accelerated Steady-state Testing (FAST) semi-integral accelerated test program in ATR. This approach is being tested/demonstrated using the metallic fuel system, but if successful it is expected to be extensible to multiple fuel types and diverse applications. This document includes an overview of the NFCSC program, a definition of science-based development of nuclear fuels, near-term goals for Advanced LWR fuels (ALFs), and longer-term goals for Advanced Reactor Fuels (ARFs) RD&D. This includes the activities that will be conducted to achieve success toward the grand challenge, as well as the goals and milestones to be achieved over the next few decades of research and development. Long-term goals are based on the DOE Office of Nuclear Energ

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Quantity of 135I Released from the AGR 5/6/7 Experiment

A series of four Advanced Reactor Technologies (ART) experiments have been conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). From 2006 through 2020, these experiments supported the development and qualification of the new U.S. tristructural isotropic (TRISO) particle fuel for Very High Temperature Reactors (VHTR). Each Advanced Gas Reactor (AGR) experiment consisted of multiple fueled capsules, each plumbed for independent temperature control using a mix of helium and neon gases. The gas leaving a capsule was routed to individual Fission Product Monitor (FPM) detectors. For intact fuel particles, the TRISO particle coatings provide a substantial barrier to fission product release. However, particles with failed coatings, whether because of a minute percentage of initially defective particles, those which fail during irradiation, or those designed to fail (DTF) particles, can release fission products to the flowing gas stream. Because reactive fission product elements like iodine and cesium quickly deposit on cooler capsule components and piping structures as the effluent gas leaves the reactor core, only the noble fission gas isotopes of Kr and Xe tend to reach FPM detectors. The FPM system utilizes High Purity Germanium (HPGe) detectors coupled with a thallium activated sodium iodide NaI(Tl) scintillator. The HPGe detector provides individual isotopic information, while the NaI(Tl) scintillator is used as a gross count rate meter. During irradiation, the 135mXe concentration reaching the FPM detectors is from both direct fission and by decay of the accumulated 135I. About 2.5 hours after irradiation (ten 15.3 minute 135mXe half lives) the directly produced 135mXe has decayed and only the longer lived 135I remains as a source. Decay systematics dictate that 135mXe will be in secular equilibrium with its 135I parent, such that its production rate very nearly equals the decay rate of the parent, and its concentration in the flowing gas stream will appear to decay with the parent half life. This equilibrium condition enables the determination of the amount of 135I released from the fuel particles by measurement of the 135mXe at the FPM following reactor shutdown. In this paper, the 135I released will be reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

IMPROVING INTERMEDIATE TEMPERATURE PERFORMANCE OF NI-YSZ CERMET ANODES FOR SOLID OXIDE FUEL CELLS BY INFILTRATION OF NICKEL NANOPARTICLES AND MIXED IONIC ELECTRONIC CONDUCTOR

Solid oxide fuel cells (SOFCs) are one of the most efficient and environment-friendly devices for electricity generation. One critical challenge of SOFC commercialization is high cell operating temperatures (800°C-1000°C), which lead to high material costs, high performance degradation rates, long start-up and shutdown times, and limited portable applications. Intermediate temperature (600°C-800°C) operation of SOFCs is limited by sluggish electrode reaction kinetics. The objective of this research is to improve intermediate temperature performance of commercially available Ni-YSZ cermet anode supported SOFCs by liquid infiltration of the anode. One effective method to improve kinetics of electrochemical reactions at the anode is to increase the density of reaction sites, which are known as the triple phase boundaries (TPBs). The porous Ni-YSZ cermet anodes were liquid infiltrated with Ni nanoparticles, leading to a four-fold increase in TPB density in the anode. The improved electrochemical performance of the infiltrated cells compared to the uninfiltrated cells highlights the effectiveness of anode infiltration in facilitating improved anode electrochemical reaction kinetics. However, the post-electrochemical testing characterization revealed that Ni nanoparticles were not stable due to Ni coarsening and were mostly isolated indicating that not all of the additional TPBs were fully utilized in electrochemical reactions due to the lack of an electronic pathway between the Ni nanoparticles. In order to improve microstructural stability of the infiltrated Ni nanoparticles, and to fully utilize the added TPBs, co-infiltration of Ni with a mixed ionic and electronic conductor (MIEC) was carried out. Two MIEC materials are chosen based on their chemical stability and conductivity in the anode operating environments; Gd 0.1 Ce 0.9 O 2-δ (GDC), a predominantly an ionic conductor, and La 0.6 Sr 0.3 Ni 0.15 Cr 0.85 O 3-δ (LSNC), a predominantly electronic conductor, and cells were successfully co-infiltrated to form Ni-GDC and Ni-LSNC nanostructures with the MIEC phases connecting the Ni nanoparticles. Stability tests demonstrated that both MIECs inhibited Ni nanoparticle coarsening. Electrochemical studies showed that Ni-GDC is the most effective for improved anode kinetics. A long-term (120 hours) electrochemical test indicated that infiltration of Ni-GDC into Ni-YSZ cermet anode effectively improves overall cell performance at intermediate temperatures and maintains the performance gain for a long period of time.

Lu, Yanchen↗
DOE OSTI2021

Photo Doppler Velocimetry and Gamma/Neutron Yield Measurements of Godiva-IV Critical Assembly

The purpose of this experiment was to collect data for a dynamic benchmark of the Godiva-IV burst assembly. Data were taken using two different measurement systems. The gamma & neutron flux was measured using a calibrated, high-dynamic-range scintillator detector, the MHD-240. The measurement will be used to determine the total fission yield and alpha throughout the burst and shutdown sequence of Godiva-IV. The Photon Doppler Velocimetry (PDV) measurement detected the surface motion of the burst assembly as a function of time. The combined data will be used in a multi-physics simulation of the Godiva-IV burst assembly.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

SMART-COM – Scalable Multi-Agent Adaptive Resolution Tools for Collaborative Outage Management

The purpose of this grant was to conduct scientific research and prototype applications to support NPP outage staff in their adaptive decision-making in efficient scheduling and resource allocation while preventing violation of safety technical specifications. The project contributed to scientific knowledge and engineering methods in (1) user interface design, (2) scheduling optimization and risk estimation, and (2) natural language processing that would benefit the nuclear power plants in minimizing schedule overruns and even unexpected shutdowns. The research team conducted site visits at a test reactor facility and an operating nuclear power plant to gather necessary information and inputs for research and development of a software application to support NPP staff in executing their outages. The final software application consisted of three modules. First, the natural language processing module supports interactive processing of technical documentation to build a database for outage staff to query non-permissible actions on system components. This module can alleviate outage staff from reviewing extensive documentation and minimize violation of technical specifications, especially in time-sensitive situations. Second, the schedule optimization module schedules outage activities and compute risk indices that outperform existing software and current practice. This module can reduce completion time of an outage that typically have too many activities for human to optimize based on current practice that does not apply the latest operations research. Finally, the visualization module presents progress and risk information of the overall outage and individual activities, as well as enabling access to the natural language processing and schedule optimization modules. This module can provide outage staff with situation awareness that are necessary to make risk-informed decisions in response to unexpected events during the execution of an outage.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗