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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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DOE OSTI2016

Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

This report outlines the thermodynamics of a supercritical carbon dioxide (sCO 2 ) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO 2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO 2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO 2 > Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO 2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository

This report discusses an initial criticality consequence analysis approach of commercial spent fuel canisters in a repository that will be developed to support a performance assessment (PA) of a repository with criticality events. A repository PA with criticality events may be needed to support direct disposal of currently loaded dual-purpose canisters. The ultimate goal is to develop a criticality consequence analysis model that can interface with any repository PA model. The criticality consequence analysis approach describes multi-physics coupling between neutronics, thermal hydraulics, and mechanics analysis codes. This report also discusses the sensitivity and uncertainty analysis approach for quantifying conservatism in the criticality consequence outcomes, as well as a defensible multi-physics validation approach to support any future licensing effort. This criticality consequence analysis approach will be refined in the future to support specific regulatory frameworks for a specific repository PA if needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Differential Privacy in Grid Kitchen: Implementation & Software Documentation

Sharing of power grid feeder models faces significant challenges due to the potential risk of exposing sensitive operational information. Traditional anonymization techniques have shown notable limitations in other sensitive domains, as evidenced by documented re-identification attacks that combine supposedly anonymized datasets with auxiliary information, raising concerns that similar vulnerabilities could affect power grid data. Consequently, there is a pressing need for a more rigorous privacy protection strategy that not only delivers formal mathematical guarantees but also preserves the analytical value of the shared models. To address this challenge, we have enhanced the Grid Kitchen framework by implementing differential privacy mechanisms within the distribution model dehydration pipeline. This implementation carefully calibrates and applies noise to sensitive attributes in feeder models according to configurable privacy levels—low, moderate, and high—each offering different balances between data utility and privacy protection. Our approach uses established noise functions (Gaussian for continuous data and Discrete Laplace for integer values) with parameters carefully calibrated so that the impact of individual data points is effectively masked in the final output. The integration leverages our Noise Catalog, which we developed to categorize feeder model properties by component type, data type, and sensitivity. This catalog guides the application of appropriate noise functions and privacy parameters ($\varepsilon$ and $\delta$) to each attribute, ensuring consistent privacy protection across the model while maintaining its structural integrity and analytical usefulness. This implementation also includes evaluation tools that allow model owners to assess the impact of privacy-preserving transformations before sharing data with external parties. This report provides documentation for the differential privacy capabilities added to the Grid Kitchen project. It includes a primer on differential privacy concepts and their importance in modern data sharing, details the architecture of our implementation, explains the privacy modes and parameter configurations, and offers practical guidance on using the code for applying differential privacy to grid feeder models. Through examples and code snippets, we demonstrate the effective application of these privacy-enhancing technologies, enabling utility operators and researchers to confidently share grid data while protecting sensitive information.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2022

Nuclide Inventory Benchmark for BWR Spent Nuclear Fuel: Challenges in Evaluation of Modeling Data Assumptions and Uncertainties

This work discusses challenges and approaches to uncertainty analyses associated with the development of a nuclide inventory benchmark for fuel irradiated in a boiling water reactor. The benchmark under consideration is being developed based on experimental data from the SFCOMPO international database. The focus herein is on how to address missing data in fuel design and operating conditions that are important for adequately simulating the time-dependent changes in fuel during irradiation in the reactor. The effects of modeling assumptions and uncertainties in modeling parameters on the calculated nuclide inventory were analyzed and quantified through computational models developed using capabilities in the SCALE code system. Particular attention was given to the impact of the power history and water coolant density on the calculated nuclide inventory, as well as to the effect of geometry modeling considerations not usually addressed in a nuclide inventory benchmark. These considerations include gap closure, channel bow, and channel corner radius, which do not usually apply to regular reactor operation but are relevant for assessing impacts of potential anomalous operating scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Numerical Simulation and Experimental Comparison of System Analysis Module 1D Mixing Model for Cold Shock Transients in the Gallium Thermal-Hydraulic Mixing Facility

Abstract Liquid metals are being investigated as coolants in many advanced reactor designs because of their high thermal conductivity and effectiveness at high temperatures. However, they often pose challenges to reactor operation and safety because of the complex thermal mixing and stratification in the plenum of pool-type reactor designs. The advanced system analysis code System Analysis Module (SAM) currently under development at Argonne National Laboratory aims to develop and implement thermal mixing models to accurately capture these complex thermal fluid behaviors. In this study, the SAM thermal mixing model was compared against experimental data from the Gallium Thermal-Hydraulic Experiment facility, a scaled liquid metal test facility that uses gallium as a surrogate fluid to investigate the stratification and thermal mixing of low-Prandtl-number fluids in the upper plenum of a liquid metal-cooled reactor. Two cold shock transient cases were used: one with stable stratified flow (Ri = 32) and one with stronger thermal mixing (Ri = 0.5). The resultant temperatures were then compared with the experimental temperatures over the entire plenum to assess the ability of the mixing models to capture the thermal behavior and to better correspond mixing parameters to various flow scenarios. Generally, the zero-dimensional mixing model was more capable of capturing the bulk temperature of the component modeled assuming that an accurate mass flow rate was provided, but it was inherently unable to capture thermal gradients in space. The one-dimensional mixing model was capable of capturing that the thermal gradients provided accurate selection of the mixing coefficients. Further, the temperature at the outlet junction was compared over time for each of the mixing models with the recorded experimental temperature. The implemented mixing models demonstrated the ability to effectively capture the overall thermal behavior for stronger mixing scenarios but struggled with more stably stratified flows. It was found that a system analysis code's covering of the entire range of different operating conditions still remains a challenging task, and it is suggested that further model and closure improvements are necessary to accurately capture complex thermal mixing and stratification phenomena.

stratification↗
DOE OSTI2016

Inspection of Dry Storage Casks Using PGAA and NAA Techniques

In this effort, an assessment of a chloride detection method is performed for inspecting the surfaces of dry cask storage system (DCSS) canisters, with focus on the implementation of a non-destructive nuclear analysis and methodology. Some DCSS canisters (especially those located in coastal environments) will be exposed to marine environments, which can cause atmospheric stress corrosion cracking (SCC). Chloride deposits on canister surfaces provide the corrosive environment necessary for atmospheric SCC with the residual stresses and materials properties conditions. The most susceptible regions on the canister surface are the weld areas, or where cold work has been performed, and may provide the necessary driving force for crack growth. The heat-affected zones (HAZs) near welds are also potentially susceptible to atmospheric SCC due to sensitization of the material caused by chromium depletion during the welding process. Information collected from the field and from laboratory studies has not been able to rule out the possibility of atmospheric chloride induced stress corrosion cracking (CISCC), although no occurrences of atmospheric SCC in DCSS canisters have been detected to-date. A number of researchers have demonstrated that localized corrosion can take place under such situations on materials commonly used for interim storage containers, including 304SS (Cook et al. 2010; Shirai et al. 2011) and 316SS (Tani et al. 2009). In this summary, the preliminary results of a neutron activation analysis (NAA) and prompt gamma activation analysis (PGAA) are performed using MCNP6, a highly sophisticated, multi-particle Monte Carlo simulation code (with the employment of the high performance computing system at the Idaho National Laboratory), and confirm the feasibility of using PGAA and NAA as an environmental inspection methodology for detection and monitoring of small quantities of chlorine on canister surfaces. The specific signature of gamma rays due reactions of sea salt constituents are documented and summarized in this paper. The environmental constituent for passive oxide film breaching is chlorine, which can cause pitting initiation and incubation time that may lead to cracking in stainless steel. Therefore, an assessment of chloride presence at a canister surface is essential before moving to other nondestructive analysis (NDA) methods for detection of pitting and cracking. Two nonlinear acoustic techniques will also be used in our integrated research project (IRP) research study that proves to be useful in the detection of SCC: nonlinear resonant ultrasound spectroscopy (NRUS), and time revisited elastic nonlinear diagnostic (TREND). These two techniques are complementary to marine atmospheric detection techniques. In this presentation only PGAA and NAA are described in detail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Real Time - Optimal Power Flow Based Distributed Energy Resources Management System (DERMS) (CRADA Number CRD-20-16909 Final Report)

The integration of behind-the-meter distributed energy resources (DERs) into distribution systems brings transformative changes to power systems. This requires operators and planners to find solutions to modernize electric grids and to effectively manage DERs for grid services. NREL developed novel DER management algorithms (referred to as Real-Time Optimal Power Flow, RT-OPF) through U.S. Department of Energy (DOE)-funded efforts, including Advanced Research Projects Agency–Energy (ARPA-E) Network Optimized Distributed Energy Systems (NODES) funding. This cutting-edge control technology aims to modernize distribution systems with large amounts of DER integration, which will help utilities solve issues brought by renewable integration and build resilient and renewable-based electric grids nationwide. Utilidata worked with NREL to investigate the commercialization opportunity of this RT-OPF-based distributed energy resource management system (DERMS). In this project, NREL performs the technology transfer of the RT-OPF to Utilidata to help them fully understand the RT-OPF solution, to identify potential engineering hurdles, and to assess the expected commercial value of various RT-OPF use cases and deployment. The technology transfer work includes two major tasks. First, NREL performs an in-depth knowledge transfer of the entire RT-OPF solution to Utilidata to help them gain an extensive and detailed understanding of the complete RT-OPF solution. In this task, NREL provides exhaustive information (e.g., documentation, code packages, laboratory and field trial data, performance results) while conducting in-depth training sessions to provide a thorough explanation of the entire solution. NREL hosts meetings to present different topics related to the RT-OPF solution, and question-and-answer sessions are included in each meeting to better explain the RT-OPF-related work. Second, the RT-OPF simulations are performed in a laboratory environment. The main objectives of this task are to walk through with Utilidata engineers how to set up a simulation of RT-OPF, identifying each RT-OPF code block/component in operation, learning how these components interact with each other, and eventually running RT-OPF simulations under various system conditions. This task helps Utilidata engineers understand performance limitations and constraints while also quantifying the commercial value of the RT-OPF for different use cases. Based on these two tasks, Utilidata engineers should be able to define and prioritize the next steps of RT-OPF implementation with an eye toward commercial success and scalability of the solution. The next steps are expected to be part of a new project following the conclusion of this project.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2024

Review of Life Cycle Cost Analysis Tools

Life cycle costing (LCC) is a vital aspect of decision-making in building retrofitting. It allows stakeholders to thoroughly assess all expenses involved throughout the lifespan of these retrofits, covering initial costs; ongoing operation, maintenance, and repair costs; and disposal costs of the system. The Federal Energy Management Program of the US Department of Energy has established clear guidelines for conducting LCC analysis, particularly for investments aimed at energy and water conservation, as well as renewable energy projects. These guidelines are detailed in the Code of Federal Regulations, 10 CFR 436, Subpart A, which outlines the methodology and procedure for LCC analysis (Department of Energy 1990). The National Institute of Standards and Technology (NIST) has developed the NIST Handbook 135 (Joshua Kneifel 2022) to standardize the process of LCC analysis within the building industry. Whereas the handbook is tailored to the needs of federal agencies, its principles and methodologies can be valuable for other organizations and industries interested in conducting life cycle costing analyses for their building projects.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2024

Method for Spatiotemporal Solar Power Profile Estimation for a Proposed U.S.–Caribbean–South America Super Grid under Hurricanes

Solar photovoltaic (PV) generation technology stands out as a scalable and cost-effective solution to enable the transition toward decarbonization. However, PV solar output, beyond the daily solar irradiance variability and unavailability during nights, is very sensitive to weather events like hurricanes. Hurricanes nucleate massive amounts of clouds around their centers, shading hundreds of kilometers in their path, reducing PV power output. This research proposes a spatiotemporal method, implemented in MATLAB R2023b coding, to estimate the shading effect of hurricanes over a wide distribution of PV solar plants connected to a high-voltage power infrastructure called the U.S.–Caribbean–South America super grid. The complete interconnection of the U.S., the Caribbean, and South America results in the lowest power valley levels, i.e., an overall percentual reduction in PV power output caused by hurricane shading. The simulations assess the impact of hurricanes in 10 synthetic trajectories spanning from Texas to Florida. The Caribbean would also experience lower power valleys with expanded interconnectivity schemes. The U.S.–Caribbean–South America super grid reduces Caribbean variability from 37.8% to 8.9% in the case study. The proposed spatiotemporal method for PV power profile estimation is a valuable tool for future solar power generation expansion, transmission planning, and system design considering the impact of hurricanes.

14 SOLAR ENERGY↗
DOE OSTI2013

Dynamic Event Tree Analysis Through RAVEN

Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics is not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also know as Dynamic Probabilistic Risk Assessment (D-PRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used D-PRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). In the past two years, the Idaho National Laboratory (INL) has developed its own tool to perform Dynamic PRA: RAVEN (Reactor Analysis and Virtual control ENvironment). RAVEN has been designed in a high modular and pluggable way in order to enable easy integration of different programming languages (i.e., C++, Python) and coupling with other application including the ones based on the MOOSE framework, developed by INL as well. RAVEN performs two main tasks: 1) control logic driver for the new Thermo-Hydraulic code RELAP-7 and 2) post-processing tool. In the first task, RAVEN acts as a deterministic controller in which the set of control logic laws (user defined) monitors the RELAP-7 simulation and controls the activation of specific systems. Moreover, RAVEN also models stochastic events, such as components failures, and performs uncertainty quantification. Such stochastic modeling is employed by using both MC and DET algorithms. In the second task, RAVEN processes the large amount of data generated by RELAP-7 using data-mining based algorithms. This paper focuses on the first task and shows how it is possible to perform the analysis of dynamic stochastic systems using the newly developed RAVEN DET capability. As an example, the Dynamic PRA analysis, using Dynamic Event Tree, of a simplified pressurized water reactor for a Station Black-Out scenario is presented.

42 ENGINEERING↗
DOE OSTI2024

Eligibility Assessment of Items in the TA-60 Rack Assembly and Alignment Complex Legacy Storage Yard

The U.S. Department of Energy, National Nuclear Security Administration, Los Alamos Field Office (NA-LA), documented and evaluated seven structures in the Technical Area 60 (TA-60) Rack Assembly and Alignment Complex (RAAC) legacy storage yard at Los Alamos National Laboratory (LANL or Laboratory) for listing in the National Register of Historic Places (National Register). This documentation and evaluation was conducted in compliance with Section 106 of the National Historic Preservation Act of 1966, as amended; the Code of Federal Regulations (36 CFR 800); the Programmatic Agreement among the U.S. Department of Energy, National Nuclear Security Administration, Los Alamos Field Office, the New Mexico State Historic Preservation Office and the Advisory Council on Historic Preservation Concerning Management of the Historic Properties at Los Alamos National Laboratory, Los Alamos New Mexico (PA)1; and A Plan for the Management of the Cultural Heritage at Los Alamos National Laboratory, New Mexico (LANL’s Cultural Resources Management Plan (2017). NA-LA makes the following National Register eligibility determinations: la cuna; the rack transporter (jeep, rack trailer, and steering dolly); the Mexia diagnostic rack; the Mexia device canister; the Mexia device mounting stand; and the Mexia target stand are eligible for listing in the National Register, and the second steering dolly is not eligible for listing in the National Register. This National Register evaluation was completed because LANL proposes to develop a consolidated waste facility in TA-60 next to the former RAAC. Across 3.56 acres, LANL proposes to construct an 8,000-square-foot, pre-engineered waste storage building; a 1,500-square-foot office/warehouse building; and 28,500 square feet of covered storage. The development will also include access control features and fencing, parking spaces, and utilities. Additionally, the Laboratory will continue to use the southern part of TA-60-0017 and reuse TA-60-0086 and TA-60-0324 in the development. The consolidated waste facility will operate as a central accumulation area—storing universal waste, mixed low-level (radioactive) waste, hazardous chemicals, and New Mexico special waste—until the waste can be shipped off site.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2013

In-Situ Testing and Performance Assessment of a Redesigned WIPP Panel Closure - 13192

There are two primary regulatory requirements for Panel Closures at the Waste Isolation Pilot Plant (WIPP), the nation's only deep geologic repository for defense related Transuranic (TRU) and Mixed TRU waste. The Federal requirement is through 40 CFR 191 and 194, promulgated by the U.S. Environmental Protection Agency (EPA). The state requirement is regulated through the authority of the Secretary of the New Mexico Environment Department (NMED) under the New Mexico Hazardous Waste Act (HWA), New Mexico Statutes Annotated (NMSA) 1978, chap. 74-4-1 through 74-4-14, in accordance with the New Mexico Hazardous Waste Management Regulations (HWMR), 20.4.1 New Mexico Annotated Code (NMAC). The state regulations are implemented for the operational period of waste emplacement plus 30 years whereas the federal requirements are implemented from the operational period through 10,000 years. The 10,000 year federal requirement is related to the adequate representation of the panel closures in determining long-term performance of the repository. In Condition 1 of the Final Certification Rulemaking for 40 CFR Part 194, the EPA required a specific design for the panel closure system. The U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) has requested, through the Planned Change Request (PCR) process, that the EPA modify Condition 1 via its rulemaking process. The DOE has also requested, through the Permit Modification Request (PMR) process, that the NMED modify the approved panel closure system specified in Permit Attachment G1. The WIPP facility is carved out of a bedded salt formation 655 meters below the surface of southeast New Mexico. Condition 1 of the Final Certification Rulemaking specifies that the waste panels be closed using Option D which is a combination of a Salado mass concrete (SMC) monolith and an isolation/explosion block wall. The Option D design was also accepted as the panel closure of choice by the NMED. After twelve years of waste handling operations and a greater understanding of the waste and the behavior of the underground salt formation, the DOE has established a revised panel closure design. This revised design meets both the short-term NMED Permit requirements for the operational period, and also the Federal requirements for long-term repository performance. This new design is simpler, easier to construct and has less of an adverse impact on waste disposal operations than the originally approved Option D design. The Panel Closure Redesign is based on: (1) the results of in-situ constructability testing performed to determine run-of-mine salt reconsolidation parameters and how the characteristics of the bedded salt formation affect these parameters and, (2) the results of air flow analysis of the new design to determine that the limit for the migration of Volatile Organic Compounds (VOCs) will be met at the compliance point. Waste panel closures comprise a repository feature that has been represented in WIPP performance assessment (PA) since the original Compliance Certification Application of 1996. Panel closures are included in WIPP PA models principally because they are a part of the disposal system, not because they play a substantive role in inhibiting the release of radionuclides to the outside environment. The 1998 rulemaking that certified WIPP to receive transuranic waste placed conditions on the panel closure design to be implemented in the repository. The revised panel closure design, termed the Run-of-Mine (ROM) Panel Closure System (ROMPCS), is comprised of 30.48 meters of ROM salt with barriers at each end. The ROM salt is generated from ongoing mining operations at the WIPP and may be compacted and/or moistened as it is emplaced in a panel entry. The barriers consist of bulkheads, similar to those currently used in the panels as room closures. A WIPP performance assessment has been completed that incorporates the ROMPCS design into the representation of the repository, and compares repository performance to that achieved with the approved Option D design. Several key physical processes and rock mechanics principles are incorporated into the performance assessment. First, creep closure of the salt rock surrounding a panel entry results in consolidation of the ROM salt emplaced in the entry. Eventually, the ROM salt comprising the ROMPCS will approach a condition similar to intact salt. As the ROM salt reaches higher fractional densities during consolidation, back stress will be imposed on the surrounding rock mass leading to eventual healing of the disturbed rock zone above and below the panel closure. Healing of the disturbed rock zone above and below the ROMPCS reduces the porosity and permeability in those areas. Analysis of the new design demonstrates that: (1) the WIPP continues to meet regulatory compliance requirements when the ROMPCS design is implemented instead of Option D, and (2) there is no impact on the short-term effectiveness of the panel closure to limit the concentration of VOCs at the WIPP site boundary to a fraction of the health-based exposure limits (HBLs) during the operational period. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Deterministic modelling of the SPERT IV reactor transients using the multi-physics capability of WIMS

The ANSWERS WIMS reactor physics code is being developed for whole core multi-physics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding. An integrated thermal hydraulic solver permits temperature and density variations to feedback to the neutronics calculation. This capability can be applied to both steady state and time dependent transient problems. Nuclear reactor design and safety case development requires assessment of a range of reactor transients, to inform both normal operation limitations and accident scenario analysis. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver for the simulation of reactor transients in whole core models. To support the validation of the multi-physics capability of WIMS, this capability has been applied to the reactivity insertion transient experiments performed in the plate fuelled SPERT-IV reactor. This study employs WIMS using the whole core solver MERLIN, which calculates the time-dependent flux distribution and magnitude, coupled to ARTHUR, which solves for the time-dependent thermal hydraulics solution, and provides thermal feedback via the cross sections generated by GEOM, which performs resonance shielding calculations and generates cross section data for the plate geometry. This spatial kinetics model, with dynamic cross section generation, allows for variations in the temperature and neutron flux profile with time.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Characterizing peak electricity demand for U.S. households: an assessment of end-use loads and demand factors

Understanding household peak electricity demand is critical to evaluate the technical need for electrical infrastructure upgrades. This study characterizes peak loads for existing and new equipment using metered data from a convenience sample of 11,940 U.S. dwellings from four sources, including 911 from two sources with end-use metering. After standardized data cleaning and labeling, we derived descriptive statistics for key metrics, such as maximum demand and demand factors, and developed predictive models relating 60- to 15-min demand for the National Electrical Code (NEC). Mean 15-min maximum demand was 9.7 kW (median 9.0 kW; IQR 7.0–11.5 kW, 95% CI 9.6–9.8 kW), indicating spare capacity in 98% of homes with hypothetical 100 A panels. Maximum demand increased with floor area and number of high-demand loads. Dwelling maximum demand was driven by higher-power, longer-duration heating appliances and vehicle charging, while most user-operated appliances contributed little. Demand factors are used to account for how most devices contribute less than their rated power to maximum demand. Existing load mean demand factors (28%; median 10%; IQR 0–58%; CI 28–29%) were higher than those for new loads (21%; median 7%; IQR 0–35%; CI 20–21%), because new loads changed the timing and magnitude of maximum demand. New high-demand loads had higher than average demand factors (40–60%). Whole dwelling demand factors support the NEC's 40% assumption, but they challenge its conservative 100% treatment of new HVAC. We propose a data-driven 50% demand factor for new equipment, which would align with metered data, improve affordability, and modernize electrical codes.

Appliances↗
DOE OSTI2017

Wall-resolved spectral cascade-transport turbulence model

A spectral cascade-transport model has been developed and applied to turbulent channel flows (Reτ= 550, 950, and 2000 based on friction velocity, uτ ; or ReδΜ= 8,500; 14,800 and 31,000, based on the mean velocity and channel half-width). This model is an extension of a spectral model previously developed for homogeneous single and two-phase decay of isotropic turbulence and uniform shear flows; and a spectral turbulence model for wall-bounded flows without resolving the boundary layer. Data from direct numerical simulation (DNS) of turbulent channel flow was used to help develop this model and to assess its performance in the 1D direction across the channel width. The resultant spectral model is capable of predicting the mean velocity, turbulent kinetic energy and energy spectrum distributions for single-phase wall-bounded flows all the way to the wall, where the model source terms have been developed to account for the wall influence. We implemented the model into the 3D multiphase CFD code NPHASE-CMFD and the latest results are within reasonable error of the 1D predictions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Implementation of FAIR principles in the IPCC: the WGI AR6 Atlas repository

The Sixth Assessment Report (AR6) of the Intergovernmental Panel on Climate Change (IPCC) has adopted the FAIR Guiding Principles. We present the Atlas chapter of Working Group I (WGI) as a test case. We describe the application of the FAIR principles in the Atlas, the challenges faced during its implementation, and those that remain for the future. We introduce the open source repository resulting from this process, including coding (e.g., annotated Jupyter notebooks), data provenance, and some aggregated datasets used in some figures in the Atlas chapter and its interactive companion (the Interactive Atlas), open to scrutiny by the scientific community and the general public. We describe the informal pilot review conducted on this repository to gather recommendations that led to significant improvements. Finally, a working example illustrates the re-use of the repository resources to produce customized regional information, extending the Interactive Atlas products and running the code interactively in a web browser using Jupyter notebooks.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2015

Progress in preparing scenarios for operation of the International Thermonuclear Experimental Reactor

The development of operating scenarios is one of the key issues in the research for ITER which aims to achieve a fusion gain (Q) of ∼10, while producing 500 MW of fusion power for ≥300 s. The ITER Research plan proposes a success oriented schedule starting in hydrogen and helium, to be followed by a nuclear operation phase with a rapid development towards Q ∼ 10 in deuterium/tritium. The Integrated Operation Scenarios Topical Group of the International Tokamak Physics Activity initiates joint activities among worldwide institutions and experiments to prepare ITER operation. Plasma formation studies report robust plasma breakdown in devices with metal walls over a wide range of conditions, while other experiments use an inclined EC launch angle at plasma formation to mimic the conditions in ITER. Simulations of the plasma burn-through predict that at least 4 MW of Electron Cyclotron heating (EC) assist would be required in ITER. For H-modes at q{sub 95} ∼ 3, many experiments have demonstrated operation with scaled parameters for the ITER baseline scenario at n{sub e}/n{sub GW} ∼ 0.85. Most experiments, however, obtain stable discharges at H{sub 98(y,2)} ∼ 1.0 only for β{sub N} = 2.0–2.2. For the rampup in ITER, early X-point formation is recommended, allowing auxiliary heating to reduce the flux consumption. A range of plasma inductance (l{sub i}(3)) can be obtained from 0.65 to 1.0, with the lowest values obtained in H-mode operation. For the rampdown, the plasma should stay diverted maintaining H-mode together with a reduction of the elongation from 1.85 to 1.4. Simulations show that the proposed rampup and rampdown schemes developed since 2007 are compatible with the present ITER design for the poloidal field coils. At 13–15 MA and densities down to n{sub e}/n{sub GW} ∼ 0.5, long pulse operation (>1000 s) in ITER is possible at Q ∼ 5, useful to provide neutron fluence for Test Blanket Module assessments. ITER scenario preparation in hydrogen and helium requires high input power (>50 MW). H-mode operation in helium may be possible at input powers above 35 MW at a toroidal field of 2.65 T, for studying H-modes and ELM mitigation. In hydrogen, H-mode operation is expected to be marginal, even at 2.65 T with 60 MW of input power. Simulation code benchmark studies using hybrid and steady state scenario parameters have proved to be a very challenging and lengthy task of testing suites of codes, consisting of tens of sophisticated modules. Nevertheless, the general basis of the modelling appears sound, with substantial consistency among codes developed by different groups. For a hybrid scenario at 12 MA, the code simulations give a range for Q = 6.5–8.3, using 30 MW neutral beam injection and 20 MW ICRH. For non-inductive operation at 7–9 MA, the simulation results show more variation. At high edge pedestal pressure (T{sub ped} ∼ 7 keV), the codes predict Q = 3.3–3.8 using 33 MW NB, 20 MW EC, and 20 MW ion cyclotron to demonstrate the feasibility of steady-state operation with the day-1 heating systems in ITER. Simulations using a lower edge pedestal temperature (∼3 keV) but improved core confinement obtain Q = 5–6.5, when ECCD is concentrated at mid-radius and ∼20 MW off-axis current drive (ECCD or LHCD) is added. Several issues remain to be studied, including plasmas with dominant electron heating, mitigation of transient heat loads integrated in scenario demonstrations and (burn) control simulations in ITER scenarios.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗