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At least 577 records · Page 32

Preliminary Thermal-hydraulics and Selected Safety Analysis for Holos-Quad Reactor Design

This report presents thermal hydraulics and safety analyses of the 10 MWe Holos-Quad micro-reactor design developed by HolosGen LLC. These analyses were executed under steady-state and station black-out (SBO) transient conditions using the System Analysis Module (SAM), with a focus on ensuring peak fuel temperatures do not exceed the safety thresholds of 1,250°C for steady state normal operating conditions and 1,600°C for transient conditions. For steady-state temperature predictions, a coupled 1D fluid-to-3D solid heat conduction methodology was used. The steady-state simulations included a model of the central core region where power peaking occurs to predict the maximum steady-state fuel temperature, and a coarse-mesh full-core model to provide the initial condition for the consequent transient SBO decay heat removal simulation. The maximum fuel temperature predicted from the central core region simulations under the normal operating condition was 1,222°C, which is below the design limit 1,250°C. To simulate a SBO transient, a full core model was necessary because of the non-axial symmetric core design and heat loss from the outer core structures to the environment. To reduce the required computational costs, a novel scaled subassembly approach was used to model the full core domain. With the full core steady-state temperature distribution provided as the initial condition, the passive decay heat removal of the core was simulated with the scaled subassembly full core model under SBO transient conditions. This analysis showed that the peak fuel temperatures of the Holos-Quad core remained below their steady-state values during the transient, therefore much lower than the safety limit 1,600°C. This is mainly due to the large thermal inertia of the graphite matrix, relatively low power density, and the large surface-to-volume-ratio of the core. The preliminary analyses presented in this report confirm the inherent safety of the Holos-Quad micro-reactor thermal-hydraulics design, as the peak fuel temperatures under both normal and off-normal operational conditions remain within design and safety limits.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Chemistry of orange/black soils from core 74001/2

Double core 74001/2 soils were studied. Bulk soils and the 90-150 micrometer and less than 20 micrometer size fractions of 13 soils from the core for both major and trace elements by instrumental neutron activation analysis were analyzed. The orange and black glassy soils in the core 74001/2 are homogeneous in composition and due to their unusual compositional characteristics probably have derived from a single source with the same composition. The top of the core has been reworked to a depth of about 5 cm. The reworked layer is marked by compositional differences in the first 3 cm which reflect the addition (at least 10%) of materials including mare basalt and a CaO-rich component. The maturity profile (I sub s/FeO) clearly indicates reworking depth of about 5 cm. It is likely that at least some of the ferro-magnetically active component (including agglutinitic glass) has been 'inherited' with the added material.

Blanchard, D. P.↗

Environmental and Circular Economy Implications of Solar Energy in a Decarbonized U.S. Grid

This report addresses environmental and circular economy (CE) considerations related to solar technologies via novel analysis of the three Solar Futures core scenarios as well as synthesis of published research. We organize these issues into the three basic life cycle phases of a solar technology: manufacturing, operation (including site selection and construction), and EOL. Related environmental justice issues are also explored. Finally, we recommend research and development (R&D) activities that could help clarify challenges and identify solutions. Because PV deployment is projected to be much larger than CSP deployment, we offer a more detailed analysis of PV-related issues.

14 SOLAR ENERGY↗

Engineering the electronic and strained interface for high activity of PdM core @Pt monolayer electrocatalysts for oxygen reduction reaction

Alloyed nanoparticles with core-shell structures provide a favorable model to modulate interfacial interaction and surface structures at the atomic level, which is important for designing electrocatalysts with high activity and durability. Herein, core-shell structured Pd 3 M@Pt/C nanoparticles with binary PdM alloy cores (M = Fe, Ni, and Co) and a monolayer Pt shell were successfully synthesized with diverse interfaces. Among these, Pd 3 Fe@Pt/C exhibited the best oxygen reduction reaction catalytic performance, roughly 5.4 times more than that of the commercial Pt/C catalyst used as reference. The significantly enhanced activity is attributed to the combined effects of strain engineering, interfacial electron transfer, and improved Pt utilization. Density functional theory simulations and extended X-ray absorption fine structure analysis revealed that engineering the alloy core with moderate lattice mismatch and alloy composition (Pd 3 Fe) optimizes the surface oxygen adsorption energy, thereby rendering excellent electrocatalytic activity. Finally, future researches may use this study as a guide on the construction of highly effective core-shell electrocatalysts for various energy conversions and other applications.

25 ENERGY STORAGE↗

Graphite Degradation Modeling and Analysis

A graphite component in a nuclear reactor core is subjected to variety of stresses and can experience degradation during normal and off-normal operation. Understanding how a graphite component will behave in service is essential to ensuring core structural stability and safe reactor operation. This report summarizes a graphite modeling tool currently under development at Idaho National Laboratory. The model incorporates several of the anticipated stresses during service and includes the effects of oxidation and irradiation prior to turnaround. This tool is intended to be used to help assess the design of graphite components by utilizing design code rules found in Section III, Division 5 of the American Society of Mechanical Engineering Boiler and Pressure Vessel Code. Specifically, the tool uses the methodologies found within the Full and Simplified assessments from Article HHA-3000 to verify that a graphite component has an acceptably low probability of failure.

36 MATERIALS SCIENCE↗

Spatial and temporal baseline of perfluorooctanesulfonic acid retained in sediment core samples from Puget Sound, Washington, USA

Per- and poly-fluorinated alkyl substances (PFAS) are a group of highly persistent synthetic chemicals utilized in many industrial and consumer products, and — most significantly toward introduction to the marine environment — in fire-fighting foams. Recently, PFAS have been linked to adverse health effects, prompting the need to understand transport and fate. Puget Sound’s main basin, hosting densely populated urban and industrial areas, is known to be contaminated with various persistent organic pollutants and provides a central focus for ecosystem monitoring. Perfluorooctanesulfonic acid, a manufactured PFAS and biodegradation product, partitions to marine sediments and thus can be used as a primary indicator toward regulatory efforts. The current study offers a spatial and temporal analysis of the Puget Sound from cores collected adjacent, Tacoma and Seattle, WA, as well as cores from central Hood Canal and Carr Inlet.

Per- and poly-fluorinated alkyl substances (PFAS),↗

Sodium-Cooled Fast Reactor Reference Plant Model

This report details the progress of Idaho National Laboratory (INL) in creating a reference plant multiphysics model for the Advanced Burner Test Reactor (ABTR). This model was developed under Task 13 of the U.S. Nuclear Regulatory Commission project “Development and Modeling Support for Advanced Non-Light Water Reactors,” and is an extension of the reference plant model developed in Task 4b, which was improved upon in the following ways. (1) The discrete ordinates method was used in lieu of the super-homogenization (SPH)-corrected diffusion approximation in order to better capture the anisotropic scattering contribution and the neutron leakage change due to thermal expansion. (2) The novel neutronic spatial discretization approach, termed the ring-heterogeneous (RH) approximation, was conceptualized and introduced to capture the differential expansion of the materials in the core. This new technique proved capable of preserving fission rates and maintaining the eigenvalue within 2.5% and 266 pcm with 9 neutron energy groups, respectively. Separating the different materials in the core enables the differential expansion of materials to be explicitly accounted for, eliminating the need for problem-specific cross-section functionalization techniques. (3) The SAM model for the core and system thermal-hydraulics analysis was updated to include 61 channels instead of just four representative ones. This enables users to obtain improved spatial resolution for sodium temperature and density scalar fields. (4) All the mesh files were created via the Multiphysics Object-Oriented Simulation Environment (MOOSE) Reactor module, eliminating all reliance on external tools for mesh creation. (5) Finally, the fuel axial expansion now leverages the HT9 and UPuZr material properties that have been validated against experimental data. The reference plant model was used to perform a full-core unprotected loss of flow (ULOF) transient calculation, including neutronics, thermal and mechanical feedback mechanisms. Future work will be devoted to further enhancements of the model. Potential improvements to the model include the addition of the control rod driveline expansion feedback and the upgrading of the support plate model so as to explicitly include 3D effects. Additionally, a Nuclear Energy Advanced Modeling and Simulation funded parallel effort has completely automated the creation of the ring-heterogeneous (RH) mesh from the fully heterogeneous (FH) geometry, thus maximizing user friendliness for the sodium fast reactor sodium-cooled fast reactor (SFR) workflow and will be incorporated in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Role of IAC in large space systems thermal analysis

Computer analysis programs to evaluate critical coupling effects that can significantly influence spacecraft system performance are described. These coupling effects arise from the varied parameters of the spacecraft systems, environments, and forcing functions associated with disciplines such as thermal, structures, and controls. Adverse effects can be expected to significantly impact system design aspects such as structural integrity, controllability, and mission performance. One such needed design analysis capability is a software system that can integrate individual discipline computer codes into a highly user-oriented/interactive-graphics-based analysis capability. The integrated analysis capability (IAC) system can be viewed as: a core framework system which serves as an integrating base whereby users can readily add desired analysis modules and as a self-contained interdisciplinary system analysis capability having a specific set of fully integrated multidisciplinary analysis programs that deal with the coupling of thermal, structures, controls, antenna radiation performance, and instrument optical performance disciplines.

Jones, G. K.↗

Multiphysics Nuclear Thermal Rocket Thrust Chamber Analysis

The objective of this effort is t o develop an efficient and accurate thermo-fluid computational methodology to predict environments for hypothetical thrust chamber design and analysis. The current task scope is to perform multidimensional, multiphysics analysis of thrust performance and heat transfer analysis for a hypothetical solid-core, nuclear thermal engine including thrust chamber and nozzle. The multiphysics aspects of the model include: real fluid dynamics, chemical reactivity, turbulent flow, and conjugate heat transfer. The model will be designed to identify thermal, fluid, and hydrogen environments in all flow paths and materials. This model would then be used to perform non- nuclear reproduction of the flow element failures demonstrated in the Rover/NERVA testing, investigate performance of specific configurations and assess potential issues and enhancements. A two-pronged approach will be employed in this effort: a detailed analysis of a multi-channel, flow-element, and global modeling of the entire thrust chamber assembly with a porosity modeling technique. It is expected that the detailed analysis of a single flow element would provide detailed fluid, thermal, and hydrogen environments for stress analysis, while the global thrust chamber assembly analysis would promote understanding of the effects of hydrogen dissociation and heat transfer on thrust performance. These modeling activities will be validated as much as possible by testing performed by other related efforts.

Wang, Ten-See↗

Enabling Dynamic Probabilistic Risk Assessment of Physical Security Using EMRALD and MAAP (Presentation)

The optimization of physical security in nuclear power plants requires sophisticated methodologies that integrate operator actions and plant behavior through advanced simulation tools. Idaho National Laboratory has developed the Modeling and Analysis for Safety and Security using the Dynamic EMRALD Framework (MASS-DEF) methodology, an approach that integrates force-on-force simulations, dynamic probabilistic risk assessment, and thermal hydraulics modeling to enhance security planning while reducing costs. A reduced order model for thermal hydraulic simulations performed by the Modular Accident Analysis Program (MAAP) was developed to evaluate reactor core behavior during attack scenarios. MAAP simulations are computationally intensive and must be run in a secure environment, complicating analysis and validation. By pre-computed scenario outcomes for a small number of modified parameters, the reduced order model significantly decreases the computational cost and enables offsite review of the results.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

Assessment of the Effect of Prototypic High-Burnup Operating Conditions of Fuel Fragmentation, Relocation, and Dispersal Susceptibility

The US nuclear energy industry is investigating strategies that further reduce the cost of energy production by using its existing fleet of nuclear generating stations. Most nuclear power plant operating costs are associated with purchasing fresh fuel assemblies or the efficiency of the reactor core design. Material costs are typically beyond the operator’s control; however, the core design optimizations offer potential operational savings. The core design envelope available to operators is constrained by two primary regulatory criteria: an enrichment limit of 5% 235 U and a burnup limit of 62 GWD/tU. These constraints have resulted in renewed efforts by the nuclear industry to pursue extending the peak rod-average burnup beyond 62 GWd/tU. This effort will likely require additional safety analyses beyond what is currently accepted by the US Nuclear Regulatory Commission. The purpose of this work is to demonstrate a best estimate plus uncertainty pin-by-pin high-burnup loss of coolant accident analysis technique to assess full-core high-burnup fuel fragmentation, relocation, and dispersal (FFRD) and identify approaches for minimizing or potentially mitigating FFRD through core design optimizations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A Coaxial Flow Reactor - A Gaseous Nuclear-Rocket Concept

Aspects of a new gaseous nuclear-rocket concept are considered. The principle of this type of cavity reactor is one of maintaining the separation of two coaxial streams - fuel within propellant - rather than attaining the separation of fuel propellant gases within the cavity. An acceptably low uranium consumption rate is possible if the fuel stream can be maintained at an axial velocity which is 1/100 to 1/1000 of the propellant velocity. An elementary hydrodynamic analysis for laminar incompressible flow indicates that the average uranium core velocity may only reach 2 to 11 times the initial core velocity for a typical reactor. The results of a 1-D criticality analysis for a D2O/(235)U system show that the effect of shrinking the fuel region within the reactor cavity is not serious; the effects of moderator heating and structural materials on critical mass are also shown. Overall rocket performance calculations are made for reasonable operating conditions; these results indicate that thrusts from 25,000 to 50,000 pounds and accelerations from 10(exp -1) to 3 g are possible.

Robert G Ragsdale↗

Detection and Monitoring of Stratigraphic Markers and Temperature Trends at the Greenland Ice Sheet Project 2 Using Passive-Microwave Remote-Sensing Data

Satellite passive-microwave sensors provide a sensitive means of studying ice-sheet surface processes that assists ice-core interpretation and can extend local observations across regional scales. Analysis of special sensor microwave/imager (SSM/I) brightness temperature (TB) data supports ice-core research in two specific ways. First, the summer hoar complex layers used to date the Holocene portion of the Greenland Ice Sheet Project 2 ice core can be defined temporally and spatially by SSM/I 37-GHz vertically (V) and horizontally (H) polarized B ratio (V/H) trends. Second, comparison of automatic weather station temperatures to SSM/I 37-GHz V TB data shows that they are an effective proxy temperature record in this region. Also, the TB data can be correlated with proxy temperature trends from stable-isotope-ratio (delta O-18 and delta-D) profiles from snow pits and this allows the assignment of dates to specific snow depths.

Shuman, C. A.↗

Fast Modular Reactor Nuclear Design Parameters of Fuel Cycle and Power Distributions

This paper describes conceptual design of a 112 MW thermal (50 MW electric) Fast Modular Reactor (FMR) system operating at 7 MPa with inlet/outlet temperatures of 509/800 ºC. The reactor system includes the reactor core, fuel assemblies, fuel rods, reactor internals, reflector, neutron control system, flow control, and structural components. The nuclear design and analysis were conducted to search for a baseline core with a cycle length greater than 8-year and the power peaking factor less than 1.5 by adjusting the fuel assembly, reactor core, and reflector configurations. The neutronics calculations of the baseline core showed that a refueling interval of 9-year is achieved with a total peaking factor of 1.47 and a fuel rod average linear power of 3.6 kW/m.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗