Advanced Reactor Research Articles

Modeling, analysis, and optimization of complex nuclear processes and facilities via computational methods: The HALEU process case study

Improving and adapting industrial systems to timely meet changing programmatic and market demands is an important goal to achieve, including when operating and maintaining complex nuclear processes and facilities. However, changes to these complex systems are costly, particularly when they are already in place and bounded to stringent requirements and constraints such as when handling radioactive material and contaminated equipment. These conditions often exist when treating spent nuclear fuel remotely within shielded nuclear radiation chambers, commonly referred as hot cells, to condition nuclear material and/or fabricate products for utilization in other nuclear enterprises such as in the manufacture of advanced nuclear fuel. The illustrative case considered here is the production of high assay low enriched uranium (HALEU) products supporting the deployment of advanced nuclear reactors. For the HALEU program, resources invested were and are being systematically analyzed so that these investments are maximized in a facility that is nearly 60 years old. A methodology that has effectively enabled optimized and improvements in the Spent Fuel Treatment (SFT) program, and consequently the HALEU program, involves discrete event simulation as addressed in this article. The quantification of multiple productivity metrics, including material processing rates, cycle times, bottlenecks, number of material transfers as well as equipment, workstation, and material handling utilization, has resulted in a myriad of diverse discoveries and data-informed decisions regarding process layout and constituent, labor levels and schedules, selection of new process units, storage needs, and other critical process configurations. This article describes such a computational capability being applied for decision-making, illustrates its application to an actual process and program, provides illustrative results, and argues how computational methods for the modeling, analysis, and optimization of complex processes and facilities does lead to informed decisions derived from data and not only from intuition.

Preliminary conceptual design with neutronics and thermal-hydraulics assessments of FuSTAR

The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction.

Hydrogen and deuterium permeation in Hastelloy N

Hastelloy N was chosen as the fluoride salt-contacting structural material for the Molten Salt Reactor Experiment due to its excellent compatibility with the fuel salt FLiBe. FLiBe is currently investigated for several advanced fusion and fission reactor concepts where tritium generation in the FLiBe is anticipated. Knowledge of hydrogen transport properties through Hastelloy N is important to understand how tritium would permeate through this material and result in an unintentional release. In this study, the hydrogen and deuterium permeability, diffusivity, and solubility were measured from 500 to 700 °C at a primary-side pressure of 10 kPa in a well-characterized sample of Hastelloy N. The prepared polycrystalline Hastelloy N had C and O impurities present on the surface. These impurities were investigated using Auger Emission Spectroscopy and Ar depth profiling. The adventitious C was removed upon the first Ar sputter cycle whereas O persisted deeper into the sample. For permeation experiments, applied deuterium pressures ranged from 13 Pa to 100 kPa and deuterium transport remained in the diffusion-limited regime (J ∝ P0.5) throughout the pressure range examined. Two methods are employed to measure the effective hydrogen and deuterium diffusivity: rise and decline. The decline method produced improved statistical model fits for calculating the effective diffusion coefficient compared to the rise method. The resultant transport properties compared well to published values for other nickel alloys.

Analysis of the Application of Nickel Superalloys for Advanced Reactor Plants

Analysis of the Application of Nickel Superalloys for Advanced Reactor Plants

Evaluation of uncertainty in antineutrino spectra normalization calculations for advanced nuclear reactor monitoring

Antineutrino detection systems have been envisioned as an important aspect of safeguarding the next generation of nuclear reactors, especially considering designs utilizing exotic fuel cycles. Deployment of antineutrino detection systems for safeguarding applications is hindered by the uncertainties associated with the calculations required for antineutrino spectra measurements and predictions. The focus of this work is to assess the impact of system components on antineutrino spectra normalization uncertainties and their significance in reactor power monitoring sensitivity. The dominant limitation in antineutrino detection calculations is typically the uncertainty associated with a cosmogenic background. This limitation becomes more pronounced when signals are weak, although the issue is mitigated in larger reactors due to their stronger source strength. Additionally, antineutrino emission uncertainties vary with the isotopic composition of the reactor fuel. Unconventional fuel cycles, featuring less common fissioning isotopes, such as Pu-240, introduce larger antineutrino yield uncertainties. The findings from this study suggest that future research on safeguard-targeted antineutrino detection should prioritize background mitigation, particularly when background simulation is necessary. Advanced nuclear reactor designs have a major influence on background understanding and successful system implementation.

Risk-informed design of advanced nuclear reactor: A target unavailability approach

Incorporating specified safety and production targets during the design phase of an engineered facility can effectively support reductions in construction and operation costs of nuclear power plants. In addition, this supports efforts toward making nuclear power technology more competitive and profitable than other electric power production forms by incorporating operational and design features into the design phase that directly support the increased likelihood of meeting such originating targets. Unavailability is the key parameter to establish plant-level capabilities that cascade down to function, system, and component levels to ensure that all sources of unavailability are addressed in the facility's design phases. In this article, two methods are presented to propose unavailability targets for nuclear reactor systems to risk-inform and optimize the design features, from an availability perspective: the first method is a simple method, and the second is a non-dominated sorting genetic algorithm III (NSGA-III) multi-objective optimization. These iterative methods are demonstrated in this paper for a hypothetical facility with a design maturity such that there is knowing the main components to comply with the engineering and security functions of the systems. With an assumed mature design, it is implied that generic data to execute quantitative reliability, and availability analyses are available at the systems levels that are sufficient to support a design-specific probabilistic risk assessment (PRA). The proposed methods modify the probabilities of basic events within the fault tree structures that model system and component unavailability sources that will be required for the facility’s operating license. The methods provided are applied to visualize the characteristics of the problem and the quantitative results. This work forms the basis for future work in which design alternatives affecting unavailabilities can be estimated and compared to their associated unavailability targets at different levels of granularity.

An Approach for Integration of User Requirement and Anthropometry Data in The Process Design of Reactor Main Control Room

The construction of a Nuclear Power Plant (NPP) using Small Modular Reactor (SMR) technology is an interesting scheme to support Net-Zero Carbon Emission. The SMR design is an advanced generation reactor with high safety and utilization features, especially the electricity needed and industry. Its modular size can also be applied to remote areas with lower construction costs compared to other types of power plants. Considering the geographical location and territory of Indonesia which is an archipelagic country, this type of reactor is suitable for application in Indonesia. To ensure safety and increase mastery of technology, it is necessary to create a simulator to support this program. Nonetheless, specific regulations govern human-machine interactions (HMI) which is covering the nuclear reactor simulators in Indonesia is not yet available. The research carried out is a review of the regulations that have been implemented in other countries, then provides a choice of operator condition designs, which are adjusted to the average size of Indonesian by considering anthropometric aspects and ergonomic aspects.

Angular distribution uncertainty influence in a large sodium-cooled fast reactor with mixed-oxide fuel

Nuclear data uncertainties are considered a main limiting factor in proving safety of advanced reactors. A reasonable amount of available information should be applied to sensitivity and uncertainty analyses for this, and this work considers the influence of usually omitted angular distribution uncertainties. A sensitivity and uncertainty analysis of MOX3600 is conducted using Serpent. The angular distribution influence is assessed by applying covariances of the state-of-the-art libraries. The sensitivity analysis shows that quantifying the influence of higher-order Legendre moment uncertainties on the eigenvalue and void reactivity is of minor importance for MOX3600. The uncertainty analysis shows the P1-coefficients for 238U and 16O introduce uncertainties to the eigenvalue similar to 56Fe and should be accounted for. A library comparison reveals TENDL-2021 and ENDF/B-VIII.0 introduce uncertainties of a few hundred pcm to the eigenvalue, and the data should be reassessed, while the other libraries give an uncertainty of several tens of pcm.

Advanced Catalyst Design and Reactor Configuration Upgrade in Electrochemical Carbon Dioxide Conversion.

Electrochemical carbon dioxide reduction reaction (CO2 RR) driven by renewable energy shows great promise in mitigating and potentially reversing the devastating effects of anthropogenic climate change and environmental degradation. The simultaneous synthesis of energy-dense chemicals can meet global energy demand while decoupling emissions from economic growth. However, the development of CO2 RR technology faces challenges in catalyst discovery and device optimization that hinder their industrial implementation. In this contribution, a comprehensive overview of the current state of CO2 RR research is provided, starting with the background and motivation for this technology, followed by the fundamentals and evaluated metrics. Then the underlying design principles of electrocatalysts are discussed, emphasizing their structure-performance correlations and advanced electrochemical assembly cells that can increase CO2 RR selectivity and throughput. Finally, the review looks to the future and identifies opportunities for innovation in mechanism discovery, material screening strategies, and device assemblies to move toward a carbon-neutral society.

Parameter study for critical heat flux under external reactor vessel cooling condition based on a mechanism model

External reactor vessel cooling has been adopted an important method to implement In Vessel Retention severe accident mitigation strategy for many advanced passive pressurized water reactors. Critical heat flux is one of the important criteria to evaluate the success or failure of this strategy. The uncertainty of engineering factors in practical design and operation conditions may make influence on critical heat flux more or less. In this paper, a parameter study for critical heat flux was carried out based on a new critical heat flux mechanism model. This mechanism model is developed based on the mechanism of liquid film dryout, and the predicted results of this model are in good agreement with the experimental results. Five important parameters, namely, coolant subcooling degree, geometry of lower head, system pressure, flow rate and heating power shape are analyzed based on this mechanism model. The results of parameter study show that the decrease of inlet coolant temperature and lower head radius are beneficial to the increase of critical heat flux. And the critical heat flux increases with the increase of system pressure. While the change of flow rate and heating power shape have little effect on critical heat flux. Under the condition of same circumference, the critical heat flux of spherical lower head is higher than that of elliptical lower head in the middle and low inclination angle region, while the trend is opposite in high inclination angle region, but not significant.

Testing of fiber optic based sensors for advanced reactors in the Texas A&M University TRIGA reactor

An Optical Fiber Based Gamma Thermometer has been developed and tested at the Texas A&M University reactor as well as the Ohio State Research Reactor. This sensor technology provides temperature measurement of a thermal mass undergoing gamma heating, from which the gamma flux along the length of the sensor may be calculated. The gamma flux data from an array of OFBGTs could allow the determination of local power density within a nuclear reactor core in three dimensions. The design of the sensor makes use of optical fiber as a truly distributed temperature sensor with a spatial resolution less than 1 mm and temperature accuracy of ± 1 °C. The testing program showed that the system provided the expected temperature data for each location. Overall, this sensor technology shows promise for current Light Water Reactors as well as advanced nuclear reactors by providing an unparalleled level of precision in reactor power measurement.

Kinetic modeling of IG-110 oxidation in inert atmosphere with low oxygen concentration for innovative high-temperature gas-cooled reactor applications

ABSTRACT The oxidation behavior of IG-110, a graphite core component, was investigated at temperatures ranging from 400 to 1000°C in a 10 ppm Ar/O2 flow to simulate the oxidation process between the graphite core component and helium coolant with low O2 concentrations employed in advanced High-Temperature Gas-cooled Reactors (HTGRs). The results reveal that IG-110 undergoes significant mass loss at temperatures above 700°C, resulting in total mass changes of −1.5%, −5.3%, and −9.0% at 700, 800, and 1000°C, respectively, during a 10-hour oxidation period. No significant mass loss is observed below 600°C. To understand the oxidation mechanism of IG-110 under low O2 concentrations, we propose a kinetic model as the current chemical kinetic-controlled model does not fully explain the oxidation behavior observed in our research. Our analysis shows lower estimated reaction rates compared to studies at higher O2 concentrations; the activation energy values exhibit good agreement. The proposed kinetic model sheds light on the oxidation mechanism of IG-110 under 10 ppm Ar/O2 flow. This study provides new insights into the oxidation behavior of graphite core components in HTGRs and highlights the importance of controlling the O2 concentration in the helium coolant to prevent severe degradation of SiC-matrix fuel compacts.

Development of an interatomic potential for L12 precipitates in Fe–Ni–Al alloys

Fe–Ni-based alloys have been considered as candidate structural materials for advanced nuclear reactors, and some of their excellent properties are particularly associated with L12 precipitates dispersed in the alloys. For a better understanding of the irradiation response, classical molecular dynamics is an essential tool to simulate the irradiation-induced defect formation and subsequent microstructural evolution. We here develop an interatomic potential for L12(Ni,Fe)3Al precipitates in the ternary Fe–Ni–Al system. The fitting parameters are tuned to describe the lattice and elastic constants of NiFe solid solutions and defect formation/vacancy migration energies in the fcc matrix. Furthermore, the lattice properties and defect energies of L12(Ni,Fe)3Al are considered for the fitting. We demonstrate the developed potential reproduces the lattice and defect properties with reasonable accuracy by comparison with density functional theory calculations, existing empirical potentials, and experimental data. This potential will be useful to investigate the defect formation and evolution of (Ni,Fe)3Al precipitates in the alloys under irradiation, which would provide critical insights into the materials design strategy to achieve enhanced properties.

Simulation of SN transport equation for hexagonal-z reactor using the COMSOL Multiphysics software

Detailed neutron transport simulation has undergone much investigation in advanced reactor engineering, where the adaptability of complex geometrics and the simplicity of computational implementation are of great significance. In this paper, a detailed neutron transport module (DNTM) based on the COMSOL Multiphysics platform is developed for hexagonal-z reactor conditions. The multi-group, multidimensional neutron transport SN equation is discretized using the least-squares finite element method and is embedded in COMSOL using the weak form partial differential equation interface without the necessity for additional coding. The proposed DNTM’s feasibility and accuracy are demonstrated through the analysis of three neutron transport benchmarks, encompassing various energy groups, dimensions, and scales. All the numerical results show good agreement with the reference values, indicating that the proposed DNTM in COMSOL can be used to solve the neutron transport process in the hexagonal-z reactor accurately and reliably. This approach may contribute a simple and practical high-fidelity neutronics solving tool and an effective direction to realize multi-physics tight coupling simulations.

Visual Style Elements in Human-System Interface Design for Nuclear Power Operations: Does Style Affect Performance and Preference?

As the world increasingly adopts renewable and sustainable energy systems, transitionary solutions include nuclear power, which currently provides 20% of the United States' electricity and is the largest single source of carbon-free electricity generation. Advanced reactors are a critical component of a carbon-free mixed energy portfolio that require careful design of first-of-a-kind control rooms. The application of Human Factors Engineering (HFE) is essential for scientific and iterative testing of novel human-system interface (HSI) concepts to ensure effective, efficient, and safe plant operations. Microworlds are simulators that use simplified physics models and control systems to distill nuclear power operations into essential functions. HFE scientists used the Rancor Microworld Simulator to obtain preference and performance metrics for novel and traditional static HSI design styles. Participants comprised advanced reactor company employees and nuclear industry consultants. A mixture of quantitative and qualitative data was captured. There was a preference for the basic graphical style that included high contrast and traditional color scheme elements. No single HSI design outperformed the others, and the participants did not perform better using their preferred HSI style. This experiment is the first in a series of HFE testing for HSIs in advanced reactor control room development. Clear user preferences emerged for elements within static displays. The cutting-edge neumorphic style was the least preferred. Future directions include tests of dynamic displays. HFE is used in evaluating and designing HSI devices that will improve the efficiency and safety of advanced nuclear power operations.

SCALE depletion capabilities for molten salt reactors and other liquid-fueled systems

Nuclear reactor systems that use fuel dissolved in a liquid have the potential for enhanced safety characteristics, improved fuel-cycle outcomes, and more efficient isotope-production configurations. In these reactor systems, the fueled liquid may simultaneously undergo irradiation, physical and chemical removal processes, and fueling. The modeling and simulation of this transmutation and decay with material additions and removals is an ongoing research area. An accurate simulation tool is critical to the reactor and fuel-cycle design, reactor deployment, and source-term characterization for these advanced reactor systems. The work described herein involved implementing, testing, and applying the capability to perform reactor physics simulations within the Oak Ridge National Laboratory-developed SCALE suite for nuclear systems analyses and design, leveraging much of its pedigree in quality-assurance and reactor-analysis capabilities. The functionalities to simulate irradiation with material feeds and removals had been added in ORIGEN, and the TRITON reactor physics sequence was extended to calculate the total removed material and track external nonirradiated mixtures to estimate separate processing or waste streams. Results from these capabilities align with analytical expectations obtained from ORIGEN for simplified test cases and with expectations for a molten salt reactor application. This implementation, available with the SCALE 6.3 release, provides for a more efficient and accurate material accountability methodology, allowing for the characterization, design, and analysis of the complete isotopic material inventory of advanced liquid-fueled systems for a variety of applications.

Cyber threat assessment of machine learning driven autonomous control systems of nuclear power plants

Advanced cyber-attacks against critical infrastructure and the energy sector are becoming more common. With the invention of autonomous control systems (ACS) within advanced nuclear reactor designs, system designers, reactor operators, and regulators must consider cybersecurity during the design and operational phases. This article provides a cyber threat assessment of machine learning (ML)-based digital twinning (DT) technologies in the context of advanced reactor ACS. A cyber–physical testbed was created to emulate nuclear reactor digital instrumentation and controls (I&C) and act as a basis for the ACS. The ACS was designed as two plant-level DTs predicting reactor malfunctions and determining control actions and two component-level DTs responsible for classifying component states and forecasting component inputs and outputs (I/O). Two duplicate ACS designs– one using a traditional ML framework and one using an automated ML (AutoML) framework– were created and tested against cyber-attacks on training data, real-time process data, and ML model architectures to determine their respective qualitative cyber-risk in terms of likelihood and impact. Both frameworks showed similar cyber-resilience against training, real-time, and ML architecture attacks, proving that neither is inherently more secure. Recommended safeguard and security measures are posed to system designers, reactor operators, and regulators to maintain the cybersecurity of ML-based DT technologies such as ACS, prompting a holistic view of shared responsibility for maintaining cyber-secure ML-based systems.

A Whole-Core Transient Thermal Hydraulic Model for Fluoride Salt-Cooled Reactors

Abstract A thermal hydraulic model is developed for a solid pin-fueled fluoride-salt-cooled small modular advanced high temperature reactor (SmAHTR). This preconceptual SmAHTR was developed by the Oak Ridge National Laboratory (ORNL). For the fuel assembly configuration investigated in this study, the fuel and non-fuel pins are arranged in a hexagonal layout. The molten FLiBe salt coolant flows parallel to the bank of pins. A finite volume model is developed and used to compute temperatures in the solid regions (fuel and non-fuel pins, and the graphite reflectors) in the core. The temperature, flow, and pressure profiles for the coolant flowing through the pin bundles in the core are calculated using the conventional subchannel methodology. Pertinent closure relations are used to compute the hydraulic losses, momentum, and energy exchange between adjacent subchannels, and heat transfer between the solid and fluid regions. The resulting model can perform both steady-state and transient computations across the entire core. This fully implicit model also includes an adaptive time-stepping algorithm for automatic time-step adjustment. A preliminary code-to-code comparison demonstrates good agreement between the present subchannel-based model and a computational fluid dynamics (CFD)-based model for a transient case in which the core inlet flowrate varies with time. Following the code-to-code comparison, the thermal hydraulic model is used to analyze the protected loss of heat sink (P-LOHS) accident scenario. Key results such as the transient evolution of peak fuel, non-fuel, and coolant temperatures, and 3-D core temperature distribution are presented and discussed.

A proposal for advanced supplementary technologies and a hybrid system with gas-cooled fast reactor concept ALLEGRO

A coupling of advanced reactors with high-capacity energy storage and other innovative non–nuclear energy technologies is a way to increase the flexibility and utilisation factor of the future nuclear systems and to diversify its energy outputs. In this paper, the basic layout of a hybrid system for a gas-cooled fast reactor is proposed. The system is composed of a high-temperature thermal energy storage unit coupled with a supercritical CO2 conversion cycle and both low-temperature and high-temperature hydrogen production and utilization. As a reference GFR reactor, the ALLEGRO concept is considered. The paper describes state-of-the-art and overviews potentially applicable technologies. Selection and justification of concepts of the individual parts of the hybrid system including operational conditions is also given. The R&D needs necessary for the potential implementation of the system are identified and the activities ongoing in the Czech Republic and especially in Centrum Výzkumu Řež are also described.

Study on creep characteristics and the isochronous stress–strain curve of Ni-Cr-Mo superalloy

ABSTRACT C276 superalloy is considered as a potential structural material for advanced nuclear reactor with good mechanical properties and corrosion resistance. High-temperature creep behaviour of C276 alloy was investigated in the temperature range of 650°C–700°C and at stresses of 140–430 MPa. A linear relationship was fitted between stress and minimum creep rate in the logarithmic coordinate system. The rupture time is analysed for life prediction in terms of isotherm extrapolation method, Monkman–Grant relation, and Larson–Miller parameter method, respectively. The isochronous stress–strain curves as a means of representing stress–strain–time relations under creep conditions were established by the parameter method. The fracture surface morphology of ruptured specimens was characterised by a scanning electron microscope to elucidate the failure mechanism.