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At least 1,747 records · Page 97

Aerosol Formation from High-Pressure Sprays for Supporting the Safety Analysis for the Hanford Waste Treatment and Immobilization Plant - 13183

The Waste Treatment and Immobilization Plant (WTP) at Hanford is being designed and built to pretreat and vitrify waste currently stored in underground tanks at Hanford. One of the postulated events in the hazard analysis for the WTP is a breach in process piping that produces a pressurized spray with small droplets that can be transported into ventilation systems. Literature correlations are currently used for estimating the generation rate and size distribution of aerosol droplets in postulated releases. These correlations, however, are based on results obtained from small engineered nozzles using Newtonian liquids that do not contain slurry particles and thus do not represent the fluids and breaches in the WTP. A test program was developed to measure the generation rate, and the release fraction which is the ratio of generation rate to spray flow rate, of droplets suspended in a test chamber and droplet size distribution from prototypic sprays. A novel test method was developed to allow measurement of sprays from small to large breaches and also includes the effect of aerosol generation from splatter when the spray impacts on walls. Results show that the release fraction decreases with increasing orifice area, though with a weaker dependence on orifice area than the currently-used correlation. A comparison of water sprays to slurry sprays with 8 to 20 wt% gibbsite or boehmite particles shows that the presence of slurry particles depresses the release fraction compared to water for droplets above 10 μm and increases the release fraction below this droplet size. (authors)

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Model for Correlating Real-Time Survey Results to Contaminant Concentrations - 12183

The U.S. Environmental Protection Agency (EPA) Superfund program is developing a new Counts Per Minute (CPM) calculator to correlate real-time survey results, which are often expressed as counts per minute, to contaminant concentrations that are more typically provided in risk assessments or for cleanup levels, usually expressed in pCi/g or pCi/m{sup 2}. Currently there is no EPA guidance for Superfund sites on correlating count per minute field survey readings back to risk, dose, or other ARAR based concentrations. The CPM calculator is a web-based model that estimates a gamma detector response for a given level of contamination. The intent of the CPM calculator is to facilitate more real-time measurements within a Superfund response framework. The draft of the CPM calculator is still undergoing internal EPA review. This will be followed by external peer review. It is expected that the CPM calculator will at least be in peer review by the time of WM2012 and possibly finalized at that time. The CPM calculator should facilitate greater use of real-time measurement at Superfund sites. The CPM calculator may also standardize the process of converting lab data to real time measurements. It will thus lessen the amount of lab sampling that is needed for site characterization and confirmation surveys, but it will not remove the need for sampling. (authors)

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Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 – 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 – 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology’s MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

U.S. EPA Superfund Counts Per Minute (CPM) Electronic Calculator - 17305

The U.S. Environmental Protection Agency (EPA) Superfund program is developing a new Counts Per Minute (CPM) calculator to correlate real-time survey results, which are often expressed as counts per minute, to contaminant concentrations that are more typically provided in risk assessments or for cleanup levels, usually expressed in pCi/g or pCi/m{sup 2}. Currently there is no EPA guidance for Superfund sites on correlating count per minute field survey readings back to risk, dose, or other Applicable or Relevant and Appropriate Requirements (ARAR) based concentrations. The CPM calculator is a web-based model that estimates a gamma detector response for a given level of contamination. The intent of the CPM calculator is to facilitate more real-time measurements within a Superfund response framework. (authors)

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Current status of the DOE/NNSA nuclear criticality safety program hands-on criticality safety training courses

In 2011 the US Department of Energy/National Nuclear Security Administration (DOE/NNSA) Nuclear Criticality Safety Program (NCSP) developed and piloted a 2-week nuclear criticality safety (NCS) practitioner course to support training and qualification of new NCS staff. The course was developed in accordance with the American National Standard Institute/American Nuclear Society (ANSI/ANS) standard for NCS training and qualifications (ANSI/ANS-8.26-2007). In 2013, an NCS manager's course was developed for process supervisors, managers, regulators, and other professionals with NCS-related responsibilities. These courses consist of classroom education, facility training, and hands-on subcritical and critical experiments training. Each course is currently offered twice per year. The 2-week practitioner course offers a week of classroom training, with practical workshops and exercises focused on teaching students how to perform an NCS evaluation. The second week of training involves hands-on critical and subcritical experiments and measurements. The first week is offered in Las Vegas, Nevada, at the DOE Nevada Field Office or the National Atomic Testing Museum. Depending on the student's clearance level, the second week is offered at Sandia National Laboratory (SNL) (uncleared and L-cleared students) or at the National Criticality Experiments Research Center (NCERC) (Q-cleared students). The 1-week manager's course is offered at SNL or NCERC, depending on clearance or interest, and includes classroom and hands-on critical and subcritical experiments and measurements. This paper provides an overview and status report for the DOE/NNSA NCSP training courses in NCS and provides information about future course offerings. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

International Low Level Waste Management Practices and Technical Bases for Very Low Level Waste Classification - 19415

Different countries around the globe use varying schemes for the classification of radioactive waste. Many countries use adaptations of the International Atomic Energy Agency (IAEA) waste classification scheme. The United States uses the Class A, B, and C waste classification scheme that is loosely based on the IAEA concepts. Classification may be based on concentration limits or dose rate limits or a combination of concentration and dose rate limits. In some countries the existence of very low-level waste categories allows nuclear power plant waste to be disposed of in hazardous waste or shallow land burial sites separate from low and intermediate level waste disposal facilities. In the United States, the very low-level waste (VLLW) classification does not exist. Nuclear power plants use alternative disposal procedures under 10CFR20.2002 and other licensed processes in agreement states in order to dispose of very low activity waste in hazardous waste disposal sites. [1] Disposal of VLLW in hazardous waste or separate VLLW facilities is generally more cost effective than disposing of waste in radioactive waste sites and is protective of the health and safety of the public. This can become an important factor when large volumes of very low activity waste are generated during decommissioning of nuclear power plants. The Electric Power Research Institute (EPRI) Radiation Safety (RS) Program has developed technical bases for the establishment of a very low-level waste category. EPRI studied how VLLW waste is defined, managed, and disposed of in various countries. This included the study of disposal site design, characteristics, and how these sites protect the health and safety of the environment and the public from low levels of radioactivity. EPRI conducted a cost benefit analysis to understand the economic impact of disposing VLLW in hazardous waste disposal sites versus low-level waste disposal sites. Finally, EPRI calculated potential radionuclide limits that may be used to define very low-level waste. The purpose of this work was to provide an example of how one might approach defining limits for VLLW. The same or similar approaches could be applied to derive radionuclide limits for different dose limit objectives. This paper presents the results of research that explored the waste management and disposal practices at six countries around the world with additional focus and details on the implementation of the very low-level waste category. (authors)

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Risk Model for EM Decision Support Toolsets

The Government Office of Accountability (GAO) has published several reports identifying the need for the Department of Energy (DOE) Office of Environmental Management (EM) to address mounting costs for DoE's cleanup program. DOEEM could greatly benefit from independent decision tool-sets/models that would allow them to evaluate options and inform business decision at the enterprise level considering site-specific life cycle costs and system plans. Program goals: Develop a tool-set that enables EM to independently evaluate alternatives, assess outcomes from different contracting strategies, and inform critical decisions for the enterprise. Project goals: Adaption of a Risk Model for integration with complimentary tool-sets for project level decision making. Operational Events: Discrete event model that evaluates operational variables (e.g. capacity, throughput, maintenance constraints) and identify bottlenecks. Identifying and Bounding Project Risk: Identify risks that impact confidence in meeting goals (e.g. production). Life cycle Cost: Evaluate impacts of staffing levels, inventory, and capital investments on life cycle costs. Methods and approach: Monte Carlo Analysis is being executed to generate results: Input derives from Risk Register Data; Tied to Projected and Target Schedules; Incorporates float duration within the model. Assumes associated risk mitigation actions being completed within a timeline of five fiscal years. Metrics include: Confidence in Meeting Production Goal; Confidence in Safety Standards; Confidence in Continuous Operation; Other Metrics can be added as appropriate regarding specific site needs. The adapted risk model can be used as a stand alone decision tool or can be integrated complementary tool-sets (i.e. process and cost models) for project-specific decisions. These support tool-sets can then be integrated with others for site- and complex- level evaluations. Future work includes designing an adaptable and modular framework that would allow integration of multiple tool-sets for holistic and/or targeted evaluation of alternative strategies for decision making that could lead to risk and cost reduction across the enterprise.

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A review of human-system interface design issues observed during analog-to-digital and digital-to-digital migrations in U.S. nuclear power plants - 312

The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is developing a scientific basis through targeted research and development (R and D) to support the U.S. nuclear power plant (NPP) fleet in extending their existing licensing period and ensuring their long-term reliability, productivity, safety, and security. Over the last several years, human factors engineering (HFE) professionals at the Idaho National Laboratory (INL) have supported the LWRS Advanced Instrumentation, Information, and Control (II and C) System Technologies pathway across several U.S. commercial NPPs in analog-to-digital migrations (i.e., turbine control systems) and digital-to-digital migrations (i.e., Safety Parameter Display System). These efforts have included in-depth human factors evaluation of proposed human-system interface (HSI) design concepts against established U.S. Nuclear Regulatory Commission (NRC) design guidelines from NUREG-0700, Rev 2 to inform subsequent HSI design prior to transitioning into Verification and Validation. This paper discusses some of the overarching design issues observed from these past HFE evaluations. In addition, this work presents some observed challenges such as common tradeoffs utilities are likely to face when introducing new HSI technologies into NPP hybrid control rooms. The primary purpose of this work is to distill these observed design issues into general HSI design guidance that industry can use in early stages of HSI design. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Numerical Analysis of Flow Instability in Once-through Steam Generator

Two-phase flow instability is a critical area to consider in the design of steam generators. Oscillation of the flow rate and the system pressure caused by heating may cause several problems and is unfavorable in an extreme case because it causes a burnout of the heat transfer surface. A large amplitude variation in the heat transfer tube wall temperature generally occurs under density wave oscillation (DWO) conditions, and a continuous cycling tube wall temperature may lead to damage to the tube, causing thermal fatigue problems. Therefore, instability of the flow should be avoided, and a two-phase flow system should be designed to have a sufficient margin to avoid this. A computer program was developed for predicting a two-phase flow instability in a steam generator heated by liquid sodium, and the results of numerical studies for several parameters are presented in this paper.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

The DOE Criticality safety support group - A retrospective perspective - 14580

The Department of Energy (DOE) Nuclear Criticality Safety Support Group (CSSG) came into being as a response to the DNFSB recommendation 1997-2 'Criticality Safety' which dealt with the continuation of criticality safety at defense nuclear facilities in the Department of Energy Enterprise. The DNFSB was concerned over the lack of capability management of practical experience pertinent to avoiding a criticality accident in non-reactor environments. One of the specific recommendations of 1997-2 was to 'Identify a core group of criticality experts experienced in the theoretical and experimental aspects of neutron chain reactions to advise on the above steps and assist in resolving future technical issues'. The CSSG, a group of 10 recognized experts in criticality safety, was chartered in late 1997 to address the recommendation. Members of the CSSG are drawn from DOE employees and contractor staff to provide advice and technical support to help meet the criticality safety needs of DOE missions, including stockpile stewardship, materials stabilization, transportation, storage, facilities life cycle (design through decommissioning), and waste disposal. The CSSG is an integral part of the DOE Nuclear Criticality Safety Program (NCSP) developed to maintain and enhance the operational and technical criticality safety expertise and capability within the Department of Energy Enterprise. This paper outlines the history, purpose and continuing contribution of the CSSG as well as providing an understanding of the interfaces between the DOE CSSG, the DOE Criticality Safety Coordinating Team (CSCT), the ANS Nuclear Criticality Safety Division and the EFCOG Criticality Safety Subgroup. (authors)

42 ENGINEERING↗

Multi-Physics Coupled Seismic Safety Analysis of Molten-Salt Reactors

Pool-type Molten Salt Reactors (MSRs) are a promising advanced nuclear reactor concept relying on passive physical behavior to provide increased safety characteristics. The high heat capacity, high boiling point, unpressurized liquid salt contains the nuclear fuel and as it passes through the vessel achieves criticality and produces nuclear power. The salt then carries away the heat produced along with the delayed neutron precursors towards heat exchangers and primary pumps before entering the core. This forms a highly coupled multiphysics problem between neutronics and fluid flow, which makes modeling these reactors challenging. The safety/licensing case heavily relies on modeling and simulation to prove their safety, and significant effort has been expanded by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to develop the appropriate simulation tools. Their safety/licensing will require a thorough analysis of their behavior during earthquake-based transients. Earthquakes are a transient condition commonly experienced in many regions, and are safely survived by dozens of reactors every year. Nonetheless, significant conservatisms were introduced with limited predictive modeling and simulation available. The first-of-a-kind capability developed by this research will allow for engineers to analyze ranges of design-basis and beyond-design basis accident scenarios to assess the integrity of the reactor vessel and surrounding system, possibly allowing for limiting the costly conservatisms in the design. This projects uses MASTODON, Griffin, and Pronghorn in the NEAMS ecosystem to model the coupled multiphysics problem posed by an earthquake in a molten salt reactor. It marches through a number of number of coupling schemes, from uncoupled simulations to tightly two-way coupled simulations. The appropriate level of coupling for these simulations will then be determined based on the accuracy in the quantities of interest (QoI) and computational effort involved in each scheme. The QoIs chosen for this paper include the mass flow rate across the heat exchanger (for Pronghorn simulations), power output of the reactor (for Griffin simulations), and the maximum Von Mises stress (for MASTODON simulations). This summary paper reports on the current progress of this project, with 2D tightly coupled multiphysics results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Interfacial Pressure Improves Calendar Aging of Lithium Metal Anodes

Lithium metal is a very attractive anode material because its theoretical specific capacity is approximately 10 times higher than conventional graphite anodes. Despite great promise, Li anodes suffer from capacity fade due to instabilities with the electrolyte as well as stranding of active Li. We have previously shown that applied interfacial pressure improves Li anode cycling because the pressure reduces the propensity for Li isolation and enables easier reconnection. Many researchers have also shown that calendar aging can lead to Li capacity loss and this has been attributed to either electrolyte decomposition with concurrent Li corrosion or to the formation of stranded Li. Our prior research focused on calendar aging during cycling suggests the mechanism for calendar aging is largely related to stranding of Li during rest and reconnection of the stranded Li upon further cycling, evidenced by similar average Coulombic efficiencies and Li loss in cells with and without rest. Because our calendar aging studies suggest Li stranding as a major cause of Coulombic efficiency drops and our Li cycling studies suggest this can be mitigated partially through applied interfacial pressure, we hypothesized that applied pressure would improve calendar aging by reducing stranded Li and enabling reconnection. We systematically varied applied pressure (0-1000 kPa) on Li metal anodes during cycling tests with and without intermittent calendar aging periods. Though the Coulombic efficiency decreases during aging periods, the lost capacity is recovered during subsequent cycles, as shown though average Coulombic efficiency and cumulative Li capacity loss analysis. We find that application of pressure partially mitigates calendar aging, in accordance with our hypothesis that calendar aging is caused by Li standing and can be mitigated to some degree with interfacial pressure. This is further supported by our results showing that the average Coulombic efficiency and cumulative Li capacity losses are similar over 50 cycles for cells that were continuously cycled and cells with periodic calendar aging periods. This result indicates that the losses during aging are reversible, which is consistent with Li stranding and reconnection. We show that pressure is one mitigation technique that helps reduce Li calendar aging in this study, but our finding that calendar aging is primarily governed by the stranding and reconnection of dead Li has wider implications. This research suggests that other mitigations which have been shown to prevent dead Li formation or encourage reconnection during cycling would also likely be successful for the purpose of improving calendar aging. The authors were supported by a Laboratory Directed Research and Development (LDRD) program. This work was performed, in part, at the Center for Integrated Nanotechnologies, an Office of Science User Facility operated for the U.S. Department of Energy (DOE) Office of Science. Sandia National Laboratories is a multi-mission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC (NTESS), a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA) under contract DE-NA0003525. This written work is authored by an employee of NTESS. The employee, not NTESS, owns the right, title and interest in and to the written work and is responsible for its contents. This work was authored in part by the National Renewable Energy Laboratory, operated by Alliance for Sustainable Energy, LLC, for the U.S. Department of Energy (DOE) under Contract No. DE-AC36-08GO28308. Any subjective views or opinions that might be expressed in the written work do not necessarily represent the views of the U.S. Government.

applied pressure↗

Design and Fabrication of a Temperature-Controlled Fueled Molten Salt Capsule Irradiation Experiment

Idaho National Laboratory (INL) is developing a Molten salt Research Temperature-controlled Irradiation (MRTI) experimental capsule for the Neutron Radiography (NRAD) reactor. The experiment is envisaged to be versatile and able to host a variety of salt-wall material combinations. The first test is designed to contain actinide-bearing chloride salt, which previously has not been subjected to a neutron field. The main objectives of the experiment are to produce irradiated salt samples for Post Irradiation Examination (PIE), to evaluate salt behavior under irradiation, and to test in-situ sensors/instrumentation. The final design for the experiment completed in 2022, and fabrication of the experiment is underway. The salt is double encapsulated providing insulation and redundancy in the case of leakage. A gas-gap between the inner capsule and the outer container enables tuning of thermal conduction to the NRAD coolant by altering its thickness and its gas-mixture. An inert gas mixture ratio in the gas-gap ensures the configurations reached a targeted average salt temperature of 600oC, while avoiding salt freezing or exceeding material limits. A resistive heater placed within a thermowell and controls the temperature of the experiment (both when the reactor is on or off). The selected configurations can contain over 10 cm3 of fuel-bearing chloride salt, which is sufficient for the purposes of the post-irradiation examination (PIE) and provides sufficient volume for two submerged thermocouple probes to monitor salt temperature. The capsule wall material is Inconel 625, and weld qualifications for the assembly have been developed. A glove-box laser weld technique was refined as part of this process to ensure a leak-tight assembly for the salt-bearing assembly and reduce the risks of impurities permeating into the salt. This work was supported through the INL Laboratory Directed Research & Development (LDRD) Program under DOE Idaho Operations Office Contract DE-AC07-05ID14517.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

ENYGF 2023 - Design of Temperature-controlled Fuel-salt Irradiation Experiment

Idaho National Laboratory (INL) is developing a Molten salt Research Temperature-controlled Irradiation (MRTI) experimental capsule for the Neutron Radiography (NRAD) reactor. The experiment is envisaged to be versatile and able to host a variety of salt-wall material combinations. The first test is designed to contain actinide-bearing chloride salt, which previously has not been subjected to a neutron field. The main objectives of the experiment are to produce irradiated salt samples for Post Irradiation Examination (PIE), to evaluate salt behavior under irradiation, and to test in-situ sensors/instrumentation. The final design for the experiment completed in 2022, and fabrication of the experiment is underway. The salt is double encapsulated providing insulation and redundancy in the case of leakage. A gas-gap between the inner capsule and the outer container enables tuning of thermal conduction to the NRAD coolant by altering its thickness and its gas-mixture. An inert gas mixture ratio in the gas-gap ensures the configurations reached a targeted average salt temperature of 600oC, while avoiding salt freezing or exceeding material limits. A resistive heater placed within a thermowell and controls the temperature of the experiment (both when the reactor is on or off). The selected configurations can contain over 10 cm3 of fuel-bearing chloride salt, which is sufficient for the purposes of the post-irradiation examination (PIE) and provides sufficient volume for two submerged thermocouple probes to monitor salt temperature. The capsule wall material is Inconel 625, and weld qualifications for the assembly have been developed. A glove-box laser weld technique was refined as part of this process to ensure a leak-tight assembly for the salt-bearing assembly and reduce the risks of impurities permeating into the salt. This work was supported through the INL Laboratory Directed Research & Development (LDRD) Program under DOE Idaho Operations Office Contract DE-AC07-05ID14517.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

EVALS: Enhanced VALidation of Advanced Battery Supply Chains

This is an overview of EVALS (Enhanced VALidation of advanced battery Supply chains), funded by the U.S. Department of Energy Vehicle Technologies Office (VTO) and some initial data from a transformational laboratory directed research and development (LDRD) program project funded at NREL. EVALS aims to accelerate the process to bring domestic and allied primary sources of battery materials online, from production to deployment.

ADVANCED PROPULSION SYSTEMS↗

Working with Trusted Local Organizations to Address Impediments to Cost-Effective and Holistic Building Performance Improvements

Commercial buildings are at the heart of any community. Investments towards the health and performance of community buildings can support a wide range of critical community and individual benefits; however, many buildings have been left behind in these investments due to a spectrum of barriers, especially those located in underserved communities. This portfolio of work stands with the intent that if we can create solutions to support the hardest-to-reach commercial buildings in gaining equitable access to energy efficiency, zero energy, and zero carbon solutions, we can support all commercial buildings with these goals. To understand success and barriers, and identify new, innovative, and impactful solutions, NREL is taking a four-part approach to: 1. Develop a methodology for defining affordability in commercial buildings and understand its impact on underserved communities, 2. Work with trusted, local organizations to address barriers that impede access to cost-effective, holistic, building performance improvements, 3. Support minority-focused developer incubator programs for, and within, underserved communities, and 4. Identify overlapping value streams that support underserved community and utility needs.

barriers↗

A Methodology for Defining Affordability in Commercial Buildings and its Impact on Underserved Communities

Commercial buildings are at the heart of any community. Investments towards the health and performance of community buildings can support a wide range of critical community and individual benefits; however, many buildings have been left behind in these investments due to a spectrum of barriers, especially those located in underserved communities. This portfolio of work stands with the intent that if we can create solutions to support the hardest-to-reach commercial buildings in gaining equitable access to energy efficiency, zero energy, and zero carbon solutions, we can support all commercial buildings with these goals. To understand success and barriers, and identify new, innovative, and impactful solutions, NREL is taking a four-part approach to: 1. Develop a methodology for defining affordability in commercial buildings and understand its impact on underserved communities, 2. Work with trusted, local organizations to address barriers that impede access to cost-effective, holistic, building performance improvements, 3. Support minority-focused developer incubator programs for, and within, underserved communities, and 4. Identify overlapping value streams that support underserved community and utility needs.

building electrification↗

Initial Split Disk Mechanical Testing Studies with a Digital Image Correlator

Nuclear graphite plays a pivotal role in the advanced reactor program. Consequently, developing a viable method to measure and evaluate small samples is essential. The Digital Image Correlation (DIC) technique allows for the examination of much smaller components in real-time applications, particularly concerning reactor capsule design and placement within the reactor. These samples become crucial for understanding the behavior of nuclear graphite before and after radiation exposure. They serve as a means to analyze changes and effectively measure the mechanical strength properties, which may or may not alter following irradiation.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗