Search NASA⌕ Search

SEARCH · Search NASA

Results for “Reactor Physics Method”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,423 records · Page 79
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Development of Thermal Tomography Imaging System for In-Service Nondestructive Evaluation of Additively Manufactured Components

Additive manufacturing (AM) is an emerging method for cost-efficient fabrication of nuclear reactor parts. AM of metallic structures for nuclear energy applications is currently based on laser powder bed fusion (LPBF) process, which can introduce internal material flaws, such as pores and anisotropy. Integrity of AM structures needs to be evaluated nondestructively because material flaws could lead to premature failures due to exposure to high temperature, radiation and corrosive environment in a nuclear reactor. Thermal tomography (TT) provides a capability for non-destructive evaluation of sub-surface defects in arbitrary size structures. We investigate TT of AM stainless steel 316L specimen with imprinted internal porosity defects using relatively low-cost, small form factor infrared (IR) camera based on uncooled micro bolometer detector. Sparse coding related K-means singular value decomposition (SVD) machine learning, image processing algorithms are developed to improve quality of TT images through removal of Additive white Gaussian noise without blurring the images. Following initial qualification of an AM component for deployment in a nuclear reactor, a compact TT can also be used for in-service nondestructive evaluation (NDE). With capability to perform in-service NDE of the AM component lifecycle, TT data can be used for development of a component digital twin. MIT research reactor (MITR) has been identified as a representative environment to perform investigation of feasibility of in-service NDE during shutdown. Preliminary study of radiation level in the MITR viewport, with geometric parameters potentially compatible with compact TT system, has been conducted.

36 MATERIALS SCIENCE↗
DOE OSTI2024

GaN-based W-band receiver chip development for fusion plasma diagnostics

Millimeter-wave diagnostics have proven effective on various magnetic fusion devices worldwide, yet the formidable challenges posed by the harsh environments of future burning plasma devices, characterized by extreme temperatures, pressures, and radiation levels, remain a significant hurdle. To address these challenges, the utilization of wide bandgap Gallium Nitride (GaN)-based millimeter-wave diagnostics is a most promising solution for fusion reactor safety monitoring and control. A noteworthy W-band GaN-based system-on-chip receiver has been the demonstrated by employing HRL T3 40 nm GaN technology. This receiver chip, compactly designed with dimensions of 3 × 5 mm 2 , incorporates essential components such as the 75–110 GHz RF Low-Noise Amplifier (LNA), mixer, Intermediate Frequency (IF) amplifier, and Local Oscillator (LO) chain. This receiver chip will be packaged as a millimeter-wave receiver module and applied on the DIII-D National Fusion Facility, for fusion plasma edge shape monitoring for operational safety and dangerous disruption prediction. The laboratory measurement results have demonstrated suitable performance. Furthermore, this advancement is pivotal for accurate analysis of plasma behavior in the extreme conditions of burning plasma devices, driving progress in fusion research and technology.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

The Covariance of PFNS Results from the Chi-Nu Experiment

The prompt fission neutron spectrum (PFNS) from neutron-induced fission is a fundamental quantity for the behavior of nuclear reactors, and has been measured many times on a wide variety of nuclei and covering different ranges of incident and emitted neutron energies. However, results from past measurements are frequently called into question in modern nuclear data evaluations because of a lack of thorough experimental documentation and incomplete uncertainty analyses. The Chi-Nu experiment at Los Alamos National Laboratory was designed to produce high-precision measurements of the PFNS of major actinides over a wide range of incident and emitted neutron energies, and with the documentation and covariance analysis required to ensure that the results of this experiment maintain their impact long into the future, thereby avoiding this pitfall of past measurements. In this work we describe the Chi-Nu experiment along with summaries of the treatment of and methods developed to address two important components of the analysis of Chi-Nu data: random-coincidence backgrounds and MCNP simulations. Furthermore, we describe the first results for correlations not just between all data points collected on a single target nucleus, but also between all data points from separate Chi-Nu measurements on 235 U and 239 Pu. These correlations are important for accurately calculating ratios of the PFNS from one actinide to another, which are rare and can be informative for nuclear data evaluation efforts.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Bubble Mass Transport Measurement in the Large-Scale Test Loops at Oak Ridge National Laboratory

Molten salts are complex fluids that incorporate multi-phase behavior depending on the chemistry and the physical properties of the entrained components. These components include the carrier salt, the actinide fuel, and fission, activation, and corrosion products, the concentrations of which depends on the burnup history of the salt. Radionuclide transport from the salt into the cover gas / off-gas system depends on volatility as predicted by thermochemistry, but data from the Molten Salt Reactor Experiment (MSRE) conducted in the late 1960s suggest that bubble formation and transport are also important. Anomalously high fractions of noble metals were found in the off-gas system and were attributed to transportation with parent salt aerosols and their association with rising noble gas bubbles. The prediction of such phenomena requires coupled neutronic, thermal hydraulic, and chemical equilibrium calculations, the framework of which is being developed within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE). However, separate effects tests and models of experiments using tools such as SAM and Thermochimica can guide model developers through important processes. Conversely, model development guides the choice of experiments and systems to provide data that are relevant for validation. This report describes several experiments that tracked gas transport in molten salts, ranging from small-scale systems to large-scale loops. Gas transportation in a molten salt, LiCl-KCl, has been studied using the shadowgraph technique. Sensors such as residual gas analysis, Raman spectroscopy, and laser-induced breakdown spectroscopy have been tested for off-gas measurements. These data were used to interpret how the gases move through the upstream salt / cover gas and the interface between them. Differential pressure measurements were able to detect individual gas bubbles as they popped at the liquid–gas interface. Salt aerosols were collected on a cascade impactor. Their formation was also observed directly via the shadowgraph method, and most of these aerosols were launched ballistically into the cover gas. Fine mists could also be observed. Convection currents through the salt were visualized and can be used to calculate the thermophysical properties of the salt itself. The apparatuses described in this report and in a previous work (McFarlane et al. 2023) have been commissioned and are available for use in making further measurements of salt/surrogate fission product behavior. Novel approaches using neutron imaging are planned for the study of fluoride salts, which cannot be contained in quartz, so shadowgraph visualization is not available. Bubble transport in convective flow and in a slow-moving salt column are planned. The mobile laser-induced breakdown spectroscopy (LIBS) system is available for several applications, including iodine capture in a molten hydroxide scrubber, H2 transport though molten salts to complement Raman analysis, and online tracking of salt aerosol generation, transport, and deposition. This report summarizes the findings from FY24 and the plans for FY25. The work completes milestone M2AT-24OR0702013 of the DOE-NE Advanced Reactor Technology, Molten Salt Reactor Campaign, DOE-NE-5.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

SULI Intern Final Report: Computationally Investigating Hydrogen Thermo-Diffusion in Yttrium Hydride Using Multiscale Methods

The renaissance of nuclear energy has arrived, heralding an age of abundant inexpensive clean energy, and renewed space exploration. In nuclear-powered spacecraft and microreactors, safety and size are of utmost importance. Yttrium Hydride (YHx) is being researched for its utility as a neutron moderator in nuclear reactors; the hydrogen in YHx slows down neutrons, enabling a continuous nuclear reaction in the reactor. This has the benefit of allowing reactors to be more safe, compact, and efficient. The goal of this effort is to computationally predict the coefficient of temperature-dependent hydrogen diffusion within YHx, the Soret coefficient. This parameter is essential for determining the safe operating modes of YHx moderators. Zirconium Hydride (ZrHx) is used in the Training, Research, Isotopes, General Atomics (TRIGA) reactor, is the reference material for these calculations. In this work, nanoscale atomic modeling in the Vienna Ab initio Simulation Package (VASP) is combined with the mesoscale finite element phase-field module in the Multiphysics Object-Oriented Simulation Environment (MOOSE); this culminates in a new multiscale computational method to simulate Soret diffusion of hydrogen in YHx. This data is useful for predicting experimental outcomes. This workflow involves convergence testing followed by static, Nudged Elastic Band (NEB), Quasi-Harmonic Approximation (QHA), and Molecular Dynamics (MD) calculations - linked with phase field simulation. NEB simulates hydrogen migration, while QHA and MD predict temperature-dependent properties. The static calculations align with literature, and preliminary NEB and QHA calculations yield accurate results. Once the atomic calculations are complete, we will incorporate Electron Backscatter Diffraction (EBSD) images and VASP-generated parameters into the phase field module to simulate intra- and intergranular transport of hydrogen in ZrHx and YHx. Future research will extend our approach to fuel-moderator materials systems such as Uranium-Yttrium Hydride (U-YHx). This work contributes to the development of advanced nuclear energy solutions for space travel.

36 - MATERIALS SCIENCE↗
DOE OSTI2018

System Studies for Global Nuclear Assurance & Security: 3S Risk Analysis for Small Modular Reactors (Volume I)—Technical Evaluation of Safety Safeguards & Security

Coupling interests in small modular reactors (SMR) as efficient and effective method to meet increasing energy demands with a growing aversion to cost and schedule overruns traditionally associated with the current fleet of commercial nuclear power plants (NPP), SMRs are attractive because they offer a significant relative cost reduction to current-generation nuclear reactors—increasing their appeal around the globe. Sandia's Global Nuclear Assurance and Security (GNAS) research perspective reframes the discussion around the "complex risk" of SMRs to address interdependencies between safety, safeguards, and security. This systems study provides technically rigorous analysis of the safety, safeguards, and security risks of SMR technologies. The aim of this research is three-fold. The first aim is to provide analytical evidence to support safety, safeguards, and security claims related to SMRs (Study Report Volume I). Second, this study aims to introduce a systems-theoretic approach for exploring interdependencies between the technical evaluations (Study Report Volume II). The third aim is to demonstrate Sandia's capability for timely, rigorous, and technical analysis to support emerging complex GNAS mission objectives.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Microfluidic, One-Batch Synthesis of Pd Nanocrystals on N-Doped Carbon in Surfactant-Free Deep Eutectic Solvents for Formic Acid Electrochemical Oxidation

One of the grand challenges that impedes practical applications of nanomaterials is the lack of robust manufacturing methods that are scalable, cheap, and environmentally friendly. In this work, we address this challenge by developing a microfluidic approach that produces surfactant-free Pd nanocrystals (NCs) uniformly loaded on N-doped porous carbon in a one-batch process. The deep eutectic solvent (DES) prepared from choline chloride and ethylene glycol was employed as a novel synthesis solvent, and its extended hydrogen networks and abundant ionic species effectively stabilize Pd facets and confine nanocrystal sizes without using surfactants. The microreactors provide faster heat exchange and more uniform mass transport, which in combination with DES produced Pd NCs with better-defined shape and predominately exposed Pd (100) facet. Furthermore, we describe that the N-doped functional groups in porous carbon direct dense and uniform heterogeneous growth of Pd NCs in a one-batch process, thereby eliminating a separate catalyst deposition step that is often involved in conventional synthesis. The Pd NCs in the one-batch-produced Pd/C catalysts exhibited a size distribution of similar to ~13 ± 3.5 nm and a high ESCA of 46.0 m 2 /g and delivered 362 mA/mg for formic acid electrochemical oxidation with improved stability, demonstrating the unique potentials of microfluidic reactors and DES for the controllable and scalable synthesis of electrocatalyst materials for practical applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Bare, Complex, and Potassium-Filled HEU Oralloy Annuli Benchmark Experiments

Hundreds of critical experiments were constructed from Oak Ridge Alloy (oralloy) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purpose of these experiments included the evaluation of storage, casting, and handling limits and providing data for verification of calculation methods and cross sections for nuclear criticality safety applications. Recent contributions to the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) include bare, complex, and potassium-filled annuli comprising multiple stacked units of highly enriched uranium (HEU), ~93.2 wt.% 235U, annuli, discs, and parallelepipeds. Prior contributions to the ICSBEP Handbook include bare (unreflected and unmoderated) cylinders; configurations of cylinders and annuli reflected with polyethylene, beryllium, or thin-graphite reflectors; GROTESQUE; and ORSphere. The newest contributions include a bare HEU 11-7 (11 inch outer diameter and 7 inch inner diameter) annulus capped at the top and bottom; bare HEU 13-7, 15-7, and 15-9 annuli; bare HEU complex annuli consisting of a 15-9 annulus containing either an off-center 7-inch-diameter cylinder, 5 × 5 inch parallelepiped, or split parallelepiped; and an HEU 13-7 annulus containing either empty stainless steel cans or potassium-metal-filled cans. The purpose of this latter pair of annulus configurations was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs. A total of nine new configurations were evaluated as acceptable benchmark experiment measurements. Additional ORCEF oralloy experiments have also been identified for future benchmark evaluation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Ru-Sn/AC for the Aqueous-Phase Reduction of Succinic Acid to 1,4-Butanediol under Continuous Process Conditions

Succinic acid is a biomass-derived platform chemical that can be catalytically converted in the aqueous phase to 1,4-butanediol (BDO), a prevalent building block used in the polymer and chemical industries. Despite significant interest, limited work has been reported regarding sustained catalyst performance and stability under continuous aqueous-phase process conditions. As such, this work examines Ru-Sn on activated carbon (AC) for the aqueous-phase conversion of succinic acid to BDO under batch and flow reactor conditions. Initially, powder Ru-Sn catalysts were screened to determine the most effective bimetallic ratio and provide a comparison to other monometallic (Pd, Pt, Ru) and bimetallic (Pt-Sn, Pd-Re) catalysts. Batch reactor tests determined that a ~1:1 metal weight ratio of Ru to Sn was effective for producing BDO in high yields, with complete conversion resulting in 82% molar yield. Characterization of the fresh Ru-Sn catalyst suggests that the sequential loading method results in Ru sites that are colocated and surface-enriched with Sn. Postbatch reaction characterization confirmed stable Ru-Sn material properties; however, upon a transition to continuous conditions, significant Ru-Sn/AC deactivation occurred due to stainless steel leaching of Ni that resulted in Ru-Sn metal crystallite restructuring to form discrete Ni-Sn sites. Computational modeling confirmed favorable energetics for Ru-Sn segregation and Ni-Sn formation at submonolayer Sn incorporation. To address stainless steel leaching, reactor walls were treated with an inert silica coating by chemical vapor deposition. With leaching reduced, stable Ru-Sn/AC performance was observed that resulted in a molar yield of 71% BDO and 15% tetrahydrofuran for 96 h of time on stream. Postreaction catalyst characterization confirmed low levels of Ni and Cr deposition, although early-stage islanding of Ni-Sn will likely be problematic for industrially relevant time scales (i.e., thousands of hours). Overall, these results (i) demonstrate the performance of Ru-Sn/AC for aqueous phase succinic acid reduction, (ii) provide insight into the Ru-Sn bimetallic structure and deactivation in the presence of leached Ni, and (iii) underscore the importance of compatible reactor metallurgy and durable catalysts.

09 BIOMASS FUELS↗
DOE OSTI2016

MELCOR/CONTAIN LMR Implementation Report - FY16 Progress.

This report describes the progress of the CONTAIN - LMR sodium physics and chemistry models to be implemented in MELCOR 2.1. In the past three years , the implementation included the addition of sodium equations of state and sodium properties from two different sources. The first source is based on the previous work done by Idaho National Laboratory by modifying MELCOR to include liquid lithium equation of state as a working fluid to model the nuclear fusion safety research. The second source uses properties generated for the SIMMER code. The implemented modeling has been tested and results are reported in this document. In addition, the CONTAIN - LMR code was derived from an early version of the CONTAIN code, and many physical models that were developed since this early version of CONTAIN are not available in this early code version. Therefore, CONTAIN 2 has been updated with the sodium models in CONTAIN - LMR as CONTAIN2 - LMR, which may be used to provide code-to-code comparison with CONTAIN - LMR and MELCOR when the sodium chemistry models from CONTAIN - LMR have been completed. Both the spray fire and pool fire chemistry routines from CONTAIN - LMR have been integrated into MELCOR 2.1, and debugging and testing are in progress. Because MELCOR only models the equation of state for liquid and gas phases of the coolant, a modeling gap still exists when dealing with experiments or accident conditions that take place when the ambient temperature is below the freezing point of sodium. An alternative method is under investigation to overcome this gap . We are no longer working on the separate branch from the main branch of MELCOR 2.1 since the major modeling of MELCOR 2.1 has been completed. At the current stage, the newly implemented sodium chemistry models will be a part of the main MELCOR release version (MELCOR 2.2). This report will discuss the accomplishments and issues relating to the implementation. Also, we will report on the planned completion of all remaining tasks in the upcoming FY2017, including the atmospheric chemistry model and sodium - concrete interaction model implementation .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Thermal Model Heat Rate Predictions of the AGR-5/6/7 Experiment

Several fuel and material irradiation experiments have been planned for the U.S. Department of Energy Advanced Gas Reactor Fuel Development and Qualification Program, which supports the development and qualification of tristructural isotropic (TRISO) coated particle fuel for use in high-temperature gas-cooled reactors. The goals of these experiments are to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support development of fuel performance models and codes, and provide irradiated fuel and materials for post-irradiation examination and safety testing. Originally planned and named as separate fuel experiments, but subsequently combined into a single test train. AGR-5/6/7 will test low-enriched uranium oxycarbide (UCO – a heterogeneous mixture of uranium oxide and uranium carbide) TRISO fuel. The AGR-5/6 portion of the experiment will provide data to support qualification of the selected reference fuel design, while the AGR-7 portion will serve as a margin test, irradiating the fuel beyond normal operating conditions. During irradiation, the temperature in each capsule is controlled by varying the helium / neon gas mixture flowing through the capsule until the desired thermocouple temperature is reached. Further adjustments can be made by changing the lobe power and the neutron filter. Three neutron filters are available to use as the experiment burns up the fissionable fuel. A thermal finite element model has been created for the five capsules comprising the AGR-5/6/7 experiment. Heat rates from a physics analysis are imported into the model along with fast neutron fluence. Graphite shrinkage due to the fast neutron fluence is incorporated into the model. Gas gaps change as a function of fast neutron fluence. Radiation heat transfer is a major contributor in this model. This is a large model with more than 1 million finite element brick elements. More than 150 different parts are modeled in the finite element model and communicate with each other from a heat transfer sense. More than 50 thermocouples are used in the experiment and are compared to actual measurements. The experiment is composed of five separate stainless steel capsules all welded together. This paper presents a method used to predict the gas mixture in the event that all of the thermocouples fail in a particular capsule. Curve fitting of heat rates from previous cycles and projecting them into the future is the basis of the method.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Surface normal emissivity measurements in uncoated and oxidized B-coated plasma facing component materials in the long wavelength infrared spectrum

Accurate surface emissivity measurements are integral to the fidelity of the infrared thermography diagnostic evaluation in fusion reactors. The emissivities of ATJ TM graphite, Sigrafine® (R6510) graphite, and TZM alloy were measured as a function of temperature in accordance with the Contact Thermometer Method in the ASTM E1933 standard. Samples were heated resistively in a high-vacuum chamber, and the surface temperature was monitored using a surface thermocouple and a Telops long-wavelength infrared camera. The surfaces of Sigrafine® graphite and TZM alloy samples were coated with 10 nm and 20 nm layers of oxidized boron and were also measured in the uncoated condition. The thickness was assessed from measurement by a quartz crystal microbalance (QCM). ATJ TM graphite was uncoated. The measured emissivity of ATJ TM Graphite ranged from 0.82 to 0.83, uncoated Sigrafine® graphite ranged from 0.76 to 0.98, and uncoated TZM alloy ranged from 0.11 to 0.13. Oxidized boron coatings increased the emissivity of the TZM alloy to 0.18–0.23 but decreased the emissivity of Sigrafine® graphite to 0.56–0.66 for a 10 nm coating and to 0.65–0.74 for a 20 nm coating. In contrast to the expected blackbody radiance, the emissivity of uncoated Sigrafine® graphite and TZM alloy did not monotonically increase with temperature from 100 to 500 ℃.

Emissivity↗
DOE OSTI2024

Studies of the outer-off-midplane lower hybrid wave launch scenario for plasma start-up on the TST-2 spherical tokamak

Abstract Establishment of an efficient central solenoid (CS) free tokamak plasma start-up method may lead to an economical fusion reactor. CS-free start-up using lower hybrid (LH) waves has been studied on the TST-2 spherical tokamak. Plasma current of about a quarter of CS-driven discharges has been obtained fully non-inductively using the outer-midplane and top LH launchers. Recently, an outer-off-midplane LH launcher was developed to achieve higher plasma current by optimizing for core absorption and minimal fast electron losses. Using the (outer-)off-midplane launcher, fully non-inductive plasma current start-up up to about 8 kA was achieved. Coupled ray-tracing and Fokker–Planck simulation was performed on equilibria reconstructed with an extended MHD model. It was found that the experimentally observed plasma current was in reasonable agreement with the numerical simulation. The simulation predicted appreciable orbit losses for the off-midplane launcher driven discharge at the present parameters, which was consistent with the experimentally observed x-ray radiation characteristics. The simulation showed that the current density was saturated for the present off-midplane launcher discharges and higher density and higher LH power was necessary to achieve higher plasma current.

Physics↗
DOE OSTI2018

Encapsulation of Defective Fuel Rods for Transport and Dry Storage - 18190

Removing of the whole inventory of fissile and non-fissile material out of the spent fuel pool is performed when the power plant is about to start its last operation phases or may occur during the extended operating life cycle due to plant storage space limitation. Removal of defective spent fuel is then definitely required for emptying the pool and may in addition also be performed in the course of plant operation for reducing pool contamination. Depending on the general used fuel management solution chosen for post-storage pool operations, specific management of a defective fuel assembly (DFA) and/or a defective fuel rod (DFR) should be implemented accordingly to the regulation requirements to be followed. For meeting these DFA and/or DFR management needs, Framatome together with New AREVA offer wide portfolio of solutions with flexibility and adaptability advantages: - from simple to advanced technologies; - for several types of fuel assemblies and/or defective rods; - for suitable and compatible cask designs; - for compatible solution with both used fuel final management strategies (closed and open fuel cycles). For instance, a technology is jointly proposed and developed by New AREVA and Framatome for implementing dry storage of DFR. It consists in: - encapsulating each DFR by loading the fuel rod in a respective capsule, drying the fuel rod using a process which allows to physically verify with high accuracy the dryness inside the DFR, and seal welding the capsule - assembling the capsules in a capsule canister (a skeleton with external dimensions similar to a respective fuel assembly) and - transferring the capsule canister along with intact spent fuel assemblies in an associated dry transport and storage (dual purpose) cask. Such a project is currently being implemented in Belgium. After successfully obtaining the cask transport and storage license extensions for including the encapsulated DFR content, the first encapsulation campaign has been started in 2017. The purpose of the paper is to present the industrial solutions proposed by Framatome and New AREVA together, for management of DFR, notably the one related to dry storage of encapsulated DFR implemented in Belgium and the perspectives opened by this successful project. The paper presents: - the available solutions for management of DFA and/or DFR depending on applicable requirements - the drying method, welding of a capsule and associated cask as the state of art technology for dry storage of DFR and/or DFA; - lessons learned from project management and encapsulation campaign performed in Belgium; - perspectives for management of DFR and/or DFA. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Survey of Cyber Risk Analysis Techniques for Use in the Nuclear Industry

Using traditional probabilistic risk analysis methods for severe accident safety risk management on non-digital systems, structures, and components at nuclear power plants is well-established. In contrast, cyber risk analysis of digital assets is still an immature field with unproven techniques due, in part, to the continuously changing threat environment and the challenge of digital assets failing in unexpected ways. As the nuclear fleet continues to adopt digital instrumentation and control systems, it is increasingly important to have effective and efficient cyber risk analysis techniques to support risk management decisions, such as risk elimination by system redesign or risk mitigation by implementation of prioritized security controls. To understand the state of the art in cyber risk analysis for future research, we surveyed 36 publications across ten application domains. We describe our survey methodology and rate each technique based upon scope, adoptability, and repeatability. In this work, we examine the unique constraints of the nuclear industry and outline the strengths and weaknesses of using the cyber risk analysis techniques in the industry, highlighting gaps with current techniques. We also discuss challenges and potential research directions for advancing the science for both existing and new advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Resonance Self-shielding Impact on Neutron Spectrum Determination for Missouri S and T Reactor

Neutron Activation Analysis is a useful technique for characterizing the neutron flux at nuclear facilities. It involves irradiating samples to produce unstable isotopes, which enables qualitative and quantitative analyses of unknown samples. In the process, samples are bombarded with neutrons; new elements then become radioactive isotopes due to absorption of neutrons. These emit different amounts of gamma energy during the decay process, and gamma rays are counted using HPGe. Each reaction product has a known half-life and cross-section, and by measuring the amount of gamma radiation energy emitted from a radioactive isotope, it is possible to identify the isotope and activity. However, the neutron spectrum is not precise or representative of the actual neutron environment because of the self-shielding effect. This occurs with large samples that have large cross-sections or thicknesses in the resonance range of the neutron flux spectrum. Several methods have been used to mitigate this problem, such as thinning or diluting alloys samples, calculating the self-shielding factor for each material, and generating a corrected library for the unfolding procedure code. The Missouri University of Science and Technology Reactor (MSTR) is a pool-type research reactor that operates up to 200 kW. It uses light-water for both moderation and natural convective heat removal. The reactor core consists of 15 fuel elements positioned in a 9 x 6 aluminum grid plate, with four elements used for reactor power control. The fuel element is 19.75% low enriched uranium in 18 curved fuel plates. The reactor has seen various configurations starting with configuration designation 101W and subsequent reconfiguration increases the numeric designation by 1. The current configuration, 120W, was implemented in 2010. The new configuration provides more in-core locations for activation experiments and uses more fuel elements than previous ones. One of the irradiation locations in the MSTR core is the source holder. It is desired to improve the documentation of neutronic characteristics of the MSTR which are cardinal to experiments supported by the facility. One of these characteristics is the neutron spectrum at various locations in the reactor. Successful efforts yielded the neutron spectrum at a known location through the unfolding of the activity of irradiated foils. The purpose of this current study is to investigate the impact of resonance self-shielding of some of the activated foils on the determined spectrum.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Electrochemical Control for Corrosion in Molten Chlorides During CSP Plant Operation

The Liquid Pathway of the Concentrating Solar Power Generation 3 (CSP Gen3) program proposed low-cost molten chloride salt for energy storage. However, online corrosion control was identified as a remining major risk of the Liquid Pathway approach. This project addressed that risk. Electrochemical solutions for corrosion mitigation during CSP plant operation were investigated and their feasibility and scalability were evaluated. The leading cause of corrosion in molten chloride salt systems was identified as corrosive impurities that form within the salt upon exposure to trace amounts of air and moisture. Leveraging electrochemistry, reduction/oxidation reactions can be employed to remove these corrosive impurities. In Phase 1 of this project, a bench-scale batch electrochemical reactor was designed, fabricated, and used to assess the kinetics and thermodynamics of electrochemical salt purification. In Phase 2, a laboratory-scale flow reactor was designed, fabricated, and used to assess the efficacy of the electrochemical method under flowing conditions. Results show that under proposed operating conditions for the Liquid Pathway Gen3 Pilot Plant, the electrochemical method is significantly more effective at removing impurities than alternative chemical and thermal methods, and that the electrochemical method produces less harmful byproducts. A key advance made in the course of this project was the development of a 2-electrode method for electrochemical purification that is more scalable than previously developed 3 electrode methods. This novel method is based on Magnesium (Mg) electrowinning. A provisional patent based on this invention has been submitted (USPTO Application No. 63/480,355). Additional key advances made during this project include assessment of the effect of dissimilar alloys on corrosion, kinetic and thermodynamic evaluation of thermolysis reactions of impurities within the molten salt, characterization of byproducts of purification reactions, and generation of IP focused on isolating value-added products using molten salt-based electrochemistry that could be deployed to valorize the process (USPTO Application No. 63/478,806). Ultimately, this project represented a step toward feasibility of Liquid Pathway Gen3 CSP. The method developed under this project could significantly reduce capital expenses and operating costs and increase plant profitability by enabling use of less expensive alloys, decreasing maintenance, and increasing plant longevity. Key focus areas for follow-on work have been identified as 1) evaluation of the efficacy of the electrochemical method under turbulent conditions in a larger flow system, such as the FASTR loop, 2) development of methods for removal of purification byproducts, 3) modeling pilot and industrial scale performance of electrochemical salt purification during plant operation and 4) further assessment of the effect of impurities on salt vapor phase.

14 SOLAR ENERGY↗
DOE OSTI2018

Numerical analysis of experimental studies of methane hydrate dissociation induced by depressurization in a sandy porous medium

Methane Hydrates (MHs) are a promising energy source abundantly available in nature. Understanding the complex processes of MH formation and dissociation is critical for the development of safe and efficient technologies for energy recovery. Many laboratory and numerical studies have investigated these processes using synthesized MH-bearing sediments. A near-universal issue encountered in these studies is the spatial heterogeneous hydrate distribution in the testing apparatus. In the absence of direct observations (e.g. using X-ray computed tomography) coupled with real time production data, the common assumption made in almost all numerical studies is a homogeneous distribution of the various phases. In an earlier study (Yin et al., 2018) that involved the numerical description of a set of experiments on MH-formation in sandy medium using the excess water method, we showed that spatially heterogeneous phase distribution is inevitable and significant. In the present study, we use as a starting point the results and observations at the end of the MH formation and seek to numerically reproduce the laboratory experiments of depressurization-induced dissociation of the spatially-heterogeneous MH distribution. This numerical study faithfully reproduces the geometry of the laboratory apparatus, the initial and boundary conditions of the system, and the parameters of the dissociation stimulus, capturing accurately all stages of the experimental process. Using inverse modelling (history-matching) that minimized deviations between the experimental observations and numerical predictions, we determined the values of all the important flow, thermal, and kinetic parameters that control the system behaviour, which yielded simulation results that were in excellent agreement with the measurements of key monitored variables, i.e. pressure, temperature, cumulative production of gas and water over time. We determined that at the onset of depressurization (when the pressure drop – the driving force of dissociation – is at its maximum), the rate of MH dissociation approaches that of an equilibrium reaction and is limited by the heat transfer from the system surroundings. Furthermore, as the effect of depressurization declines over time, the dissociation reaction becomes kinetically limited despite significant heat inflows from the boundaries, which lead to localized temperature increases in the reactor.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗