Results for “Unit Testing” — Page 72

DOE OSTI2016

SU-F-J-37: Commissioning a Six Degree of Freedom Couch

Purpose: The proper operation of a six degree of freedom couch (6DOFC) to position a patient for treatment requires that a number systems on the treatment unit work together correctly. Commissioning a 6DOFC requires not only the commissioning of each separate system but also commissioning the interactions between the systems. Methods: To commission a 6DOFC, we tested the resolvers that readout the linear and angular positions; the orthogonality of the axes of the laser positioning system, the imaging system, the gantry and the couch as well as their relationship to each other. An automated Winston-Lutz test was used verify the coincidence of the origins of the laser system, imaging system and treatment unit. The IsoCal phantom was used to verify the consistency of axes orientations between systems. The operation of the couch and imaging system was verified under simulated load conditions and an end to end test was performed to verify correct operation of the system. Results: The tests on the couch resolvers, orthogonality and coincidence of the axes of each system were successfully completed and showed excellent agreement. The automated Winston-Lutz test showed that the maximum displacement of the center of the radiation field with respect to the isocenter was <0.6 mm (radius) at isocenter. The analysis of the IsoCal phantom showed that the maximum variation of the radiation field with respect to known locations in the phantom was ∼0.9 mm for all the BBs in the phantom. Load tests showed maximum errors of 1 mm. Conclusion: We tested all system associated with the 6DOFC and found the uncertainty due to imaging was <.9mm and the uncertainty due to loading ∼1 mm, resulting in a total error of ∼1.4 mm.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2018

Crossflow Filtration of Hanford Tank AP-105 Supernatant

A small-scale test platform to demonstrate the solids filtration, Cs removal, and low-activity waste (LAW) vitrification was constructed and installed at the Pacific Northwest National Laboratory (PNNL) 325 Building. Bench scale testing was conducted to demonstrate cross flow filtration using the cells unit filter (CUF) on approximately 14 liters of 241-AP-105 feed to support the confirmation of the Low-Activity Waste Pretreatment System (LAWPS) equipment performance and design basis for entrained solids removal process. Despite the low solids content (~16 ppm) of the AP-105 feed, significant fouling of the filter was observed that required backpulsing of the filter to maintain the permeate flux within the LAWPS baseline design rate. The improvement from backpulsing was significant, but transitory, and lasted a few hours. Increasing the transmembrane pressure was shown to have limited impact on the permeate flux in terms of magnitude and duration, especially compared with backpulsing. Because LAWPS must operate three to eight days to throughput requirements, testing demonstrated that backpulsing or chemical cleaning may need to be done every 7 to 14 hours when processing feed from AP-105. Because of this, backpulsing capability is strongly recommended for the baseline design. The LAWPS prototypic filter cleaning was demonstrated at the end of testing and was shown to be very effective as it largely restored the filter permeability to near pre-test conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2014

AGR-2 Irradiation Test Final As-Run Report, Rev 2

This document presents the as-run analysis of the AGR-2 irradiation experiment. AGR-2 is the second of the planned irradiations for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. Funding for this program is provided by the U.S. Department of Energy as part of the Very High Temperature Reactor (VHTR) Technical Development Office (TDO) program. The objectives of the AGR-2 experiment are to: (a) Irradiate UCO (uranium oxycarbide) and UO 2 (uranium dioxide) fuel produced in a large coater. Fuel attributes are based on results obtained from the AGR-1 test and other project activities. (b) Provide irradiated fuel samples for post-irradiation experiment (PIE) and safety testing. (c) Support the development of an understanding of the relationship between fuel fabrication processes, fuel product properties, and irradiation performance. The primary objective of the test was to irradiate both UCO and UO 2 TRISO (tri-structural isotropic) fuel produced from prototypic scale equipment to obtain normal operation and accident condition fuel performance data. The UCO compacts were subjected to a range of burnups and temperatures typical of anticipated prismatic reactor service conditions in three capsules. The test train also includes compacts containing UO 2 particles produced independently by the United States, South Africa, and France in three separate capsules. The range of burnups and temperatures in these capsules were typical of anticipated pebble bed reactor service conditions. The results discussed in this report pertain only to U.S. produced fuel. In order to achieve the test objectives, the AGR-2 experiment was irradiated in the B-12 position of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for a total irradiation duration of 559.2 effective full power days (EFPD). Irradiation began on June 22, 2010, and ended on October 16, 2013, spanning 12 ATR power cycles and approximately three and a half calendar years. The test contained six independently controlled and monitored capsules. Each U.S. capsule contained 12 compacts of either UCO or UO2 AGR coated fuel. No fuel particles failed during the AGR-2 irradiation. Final burnup values on a per compact basis ranged from 7.26 to 13.15% FIMA (fissions per initial heavy-metal atom) for UCO fuel, and 9.01 to 10.69% FIMA for UO 2 fuel, while fast fluence values ranged from 1.94 to 3.47 x 10 25 n/m 2 (E >0.18 MeV) for UCO fuel, and from 3.05 to 3.53 x 10 25 n/m 2 (E >0.18 MeV) for UO 2 fuel. Time-average volume-average (TAVA) temperatures on a capsule basis at the end of irradiation ranged from 987°C in Capsule 6 to 1296°C in Capsule 2 for UCO, and from 996 to 1062°C in UO 2 -fueled Capsule 3. By the end of the irradiation, all of the installed thermocouples (TCs) had failed. Fission product release-to-birth (R/B) ratios were quite low. In the UCO capsules, R/B values during the first three cycles were below 10 -6 with the exception of the hotter Capsule 2, in which the R/Bs reached 2 x 10 -6 . In the UO 2 capsule (Capsule 3), the R/B values during the first three cycles were below 10 -7 . R/B values for all following cycles are not reliable due to gas flow and cross talk issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Evaluation of a Potential Groundwater Tracer Test in the Ringold Upper Mud Aquifer at the 100-H Area of the Hanford Site

This document presents an evaluation of a potential groundwater tracer test in the uppermost RUM (Ringold Formation member of Wooded Island – upper mud unit) aquifer in the 100-H Area of the Hanford Site. The results from field electrical resistivity tomography (ERT) investigations, numerical flow and transport (F&T) modeling, coupled ERT-F&T model simulations, and laboratory resin experiments provide the technical basis for evaluating tracer test scenarios that minimize impacts to the 100-HX pump and treat (HX P&T) operations, and maximize hydrologic data acquisitions including the use of advanced surface geophysical techniques for monitoring tracer transport.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Irreversible Wash Aid Additive for Cesium Mitigation. Small-Scale Demonstration and Lessons Learned

The Irreversible Wash Aid Additive process has been under development by the U.S. Environmental Protection Agency (EPA) and Argonne National Laboratory (Argonne). This process for radioactive cesium mitigation consists of a solution to wash down contaminated structures, roadways, and vehicles and a sequestering agent to bind the radionuclides from the wash water and render them environmentally immobile. The purpose of this process is to restore functionality to basic services and immediately reduce the consequences of a radiologically-contaminated urban environment. Research and development have resulted in a down-selection of technologies for integration and demonstration at the pilot-scale level as part of the Wide Area Recovery and Resiliency Program (WARRP) under the Department of Homeland Security and the Denver Urban Area Security Initiative. As part of developing the methods for performing a pilot-scale demonstration at the WARRP conference in Denver in 2012, Argonne conducted small-scale field experiments at Separmatic Systems. The main purpose of these experiments was to refine the wash water collection and separations systems and demonstrate key unit operations to help in planning for the large scale demonstration in Denver. Since the purpose of these tests was to demonstrate the operations of the system, we used no radioactive materials. After a brief set of experiments with the LAKOS unit to familiarize ourselves with its operation, two experiments were completed on two separate dates with the Separmatic systems.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2019

Terry Turbopump Expanded Operating Band Modeling and Full-Scale Test Development Efforts in FY19 (Progress Report)

The Terry Turbine Expanded Operating Band Project is currently conducting testing at Texas A&M University, and the resulting data has been incorporated into MELCOR models of the Terry turbines used in nuclear power plants. These improved models have produced improvements in the Fukushima Daiichi Unit 2 simulations while providing new insights into the behavior of the plant. The development of future experimental test efforts is ongoing. Development of and refinements to the plans for full-scale steam and steam-water turbine ingestion testing has been performed. These full-scale steam-based tests will complement the testing occurring at Texas A&M University, and will resolve the remaining questions regarding scale or working fluid. Planning work has also begun for future testing intended to explore the uncontrolled RCIC self-regulation theorized to have occurred in Fukushima Daiichi Unit 2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2013

Clopidogrel Responsiveness in Patients Undergoing Peripheral Angioplasty

Purpose: To investigate the incidence and clinical significance of platelet responsiveness in patients receiving clopidogrel after peripheral angioplasty procedures. Materials and Methods: This prospective study included patients receiving antiplatelet therapy with clopidogrel 75 mg after infrainguinal angioplasty or stenting and who presented to our department during routine follow-up. Clopidogrel responsiveness was tested using the VerifyNow P2Y12 Assay. Patients with residual platelet reactivity units (PRU) {>=} 235 were considered as nonresponders (NR group NR), whereas patients with PRU < 235 were considered as normal (responders [group R]). Primary end points were incidence of resistance to clopidogrel and target limb reintervention (TLR)-free survival, whereas secondary end points included limb salvage rates and the identification of any independent predictors influencing clinical outcomes. Results: In total, 113 consecutive patients (mean age 69 {+-} 8 years) with 139 limbs were enrolled. After clopidogrel responsiveness analysis, 61 patients (53.9 %) with 73 limbs (52.5 %) were assigned to group R and 52 patients (46.1 %) with 66 limbs (47.5 %) to group NR. Mean follow-up interval was 27.7 {+-} 22.9 months (range 3-95). Diabetes mellitus, critical limb ischemia, and renal disease were associated with clopidogrel resistance (Fisher's exact test; p < 0.05). According to Kaplan-Meier analysis, TLR-free survival was significantly superior in group R compared with group NR (20.7 vs. 1.9 %, respectively, at 7-year follow-up; p = 0.001), whereas resistance to clopidogrel was identified as the only independent predictor of decreased TLR-free survival (hazard rate 0.536, 95 % confidence interval 0.31-0.90; p = 0.01). Cumulative TLR rate was significantly increased in group NR compared with group R (71.2 % [52 of 73] vs. 31.8 % [21 of 66], respectively; p < 0.001). Limb salvage was similar in both groups. Conclusion: Clopidogrel resistance was related with significantly more repeat interventions after peripheral angioplasty procedures.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

DOE OSTI2019

Quantifying Climate and Catchment Control on Hydrological Drought in the Continental United States

The evolution of hydrological drought events is a result of complex (nonlinear) interactions between climate and catchment processes. To investigate such nonlinear relationship, we integrated a machine learning modeling framework based on the random forest (RF) algorithms with an interpretation framework to quantify the role of climate and catchment controls on hydrological drought. More particularly, our framework interprets a built RF machine-learning model to identify dominant variables and visualize their functional dependence and interaction effects on hydrological drought characteristics utilizing concepts of minimal depth, interactive depth, and partial dependence. We test our proposed modeling framework based on a set of 652 continental United States catchments with minimal human interference for a period of 1979–2010. Application of this framework indicated presence of three distinct drought regimes, which includes, Regime 1: droughts with longer duration, less frequent and lesser intensity; Regime 2: droughts with moderate duration, moderate frequency, and moderate intensity; and Regime 3: droughts with shorter duration, more frequent, and more intense. RF algorithm was able to accurately model the drought characteristics (intensity, duration, and number of events) for all the three drought regimes as a function of selected variables. It was observed that the type of dominant variables as well as their nonlinear functional relationship with hydrological droughts characteristics can vary between three selected regimes. Finally, our interpretation framework indicated that catchment characteristics have a significant role in controlling the hydrologic drought for catchments (regime 1), whereas both climate and catchment characteristics control hydrological drought in regimes 2 and 3.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2022

Development of an MC&A toolbox for liquid-fueled molten salt reactors with online reprocessing (Final Report)

A critical barrier to the deployment of MSRs is the absence of a well-defined nuclear material control and accounting (MC&A) approach, a vital prerequisite to meet NRC licensing requirements as well as facilitating future international exports. Liquid-fueled MSR variants (especially those incorporating online refueling or reprocessing) present a unique set of challenges to traditional nuclear material control and accountancy. Unlike solid-fueled light-water reactors, traditional item-counting methods cannot be applied as an accountancy strategy. Rather, MC&A approaches to MSR variants (including both uranium and thorium fueled designs) are more analogous to bulk material handling facilities (e.g., enrichment and reprocessing); yet further complicating matters is the fact that fuel medium is also highly radioactive. Meanwhile, the space of MSRs covers a broad range of design parameters, including thermal and fast spectra designs; operation in actinide breeder or burner modes, choice of actinide fuel used (e.g., 235 U, 232 Th / 233 U, denatured 233 U, etc.), pool or loop-type configuration, and even different salt chemistry. Each of these design choices introduce significant challenges to MC&A approaches within MSR facilities. We propose to bridge this gap for liquid-fueled MSRs by developing a modular, component-based test framework for evaluating viable process monitoring and MC&A techniques specifically suited to liquid-fueled MSR system variants employing online refueling or reprocessing. This test platform will consist of a toolbox of independent process modules representing discrete physical units (such as the reactor core, off-gas processing, decay tanks, and actinide separation units), each with its own self-contained physics responsive to the input mass flow, along with appropriate measurement models that can be coupled to key flow points. These dynamic physical signatures thus afford the ability to test the viability and efficacy of potential accountancy techniques under the full range of reactor operating conditions. As process modules are connected via mass flows, the result is a reconfigurable, generic MSR mass flow model capable of serving as an MC&A test platform for a broad spectrum of possible MSR configurations. The resulting MSR MC&A toolbox will enable robust assessment of accountancy strategies for this unique facility type, including analysis of physical feedbacks arising both from depletion of the fuel over time as well as from potential off-normal events, including those introduced by equipment failures (e.g., a pump failure) as well as by malicious action (i.e., attempts to divert material). The proposed toolbox both addresses a critical needs area for the MPACT analysis toolkit while leveraging existing MPACT-sponsored tools, especially with respect to simulation of measurement and accountancy techniques for advanced fuel cycle facilities. Beyond enabling new analysis capabilities for MSR systems, the design of this toolbox will be to such to maximize compatibility with existing MPACT tools, such to enhance existing facility MC&A analysis capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Distributed Automatic Generation Control Considering DPV Using T&D Dynamic Co-Simulation

The increasing adoption of distributed energy resources (DERs) over the last decade warrants a reconsideration of control of generation resources. This paper proposes a distributed Automatic Generation Control (AGC) using transmission-and-distribution (T&D) dynamic co-simulation framework for the efficient DPV frequency regulation services. The co-simulation framework allows AGC units to exchange the information for distributed AGC, based on their adopted communication network topology. As a result, a cost-effective automatic generation control is achieved with DPV and conventional generators. The proposed distributed AGC is based on the gossip algorithm in which the neighboring AGC units share the relevant local information with each other and updates their share of AGC regulation signal. Distributed photovoltaics (DPV) unit contribute to AGC response based on their headroom capacity via DER aggregators. The algorithm is tested on IEEE-14 bus transmission system under conditions of generation failure and random load variation to observe effective frequency regulations service offered by DPVs and other AGC units. The study shows that DPV can effectively participate in AGC with the proposed distributed control framework.

automatic generation control↗

DOE OSTI2024

MARVEL Reactor Fuel Performance Report (Rev.2)

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Multiscale Mechanical Characterization of Mineral-Reinforced Wood Cell Walls

Studying the multiscale mechanics of bio-based composites offers unique perspectives on underlying structure–property relations. Cellular materials, such as wood, are highly organized, hierarchical assemblies of load-bearing structural elements that respond to mechanical stimuli at the microscopic, mesoscopic and macroscopic scale. In this study, we modified oak wood with nanocrystalline ferrihydrite, a widespread ferric oxyhydroxide mineral, and characterized the resulting mechanical properties of the composite at various levels of organization. Ferrihydrite nanoparticles were deposited inside the wood cell wall by an in situ chemical reaction, resulting in increased stiffness and hardness of the functionalized secondary cell wall, as evidenced by region-specific nanoindentation tests under an electron microscope. Chemically modified and pristine wood samples were characterized by using atomic force microscopy in the bimodal frequency modulation mode, which produced topographical images from the cellular ultrastructure with high lateral resolution and localized nanomechanical information across distinct cell wall layers. In conclusion, despite mineral reinforcement at the cell wall level, the macroscopic fracture behavior examined through three-point flexural testing remained unchanged upon modification, as cell–cell adhesion could be impaired by harsh chemical conditions.

Cells↗

DOE OSTI2026

Multiple Hail Impact Testing

For resilient energy delivery PV modules and systems must withstand extreme weather events such as hailstorms, which are the leading cause of PV insurance claims. Currently, the industry testing is limited to one strike at a time, but in real hailstorms multiple hail strikes can occur within a fraction of a second. In this project we test if multiple simultaneous or near-simultaneous hail strikes cause more damage than with single hail strike testing.

14 SOLAR ENERGY↗

DOE OSTI2017

Microstructure of In-Core Molten Salt Corrosion Hastelloy N{sup R} and 316 Stainless Steel

Since 2011, many materials have been discussed for potential application in Fluoride salt-cooled High-temperature nuclear Reactors (FHRs), which are emerging as leading reactor concept for the next generation of nuclear reactors thanks to the benefits of a high degree of passive safety, high thermal efficiency, and low spent fuel volume. Among these discussed materials, Hastelloy N{sup R} and 316L stainless steel have been selected as the most likely candidate alloys for FHRs. To provide direct experience in integrating a real system using molten FLiBe salt for reactor operation, these two candidate alloys were experimentally tested in Li-7 enriched Li{sub 2}BeF{sub 4} (FLiBe) molten salt in the core of MIT Research Reactor (MITR) at 700 deg. C for 1000 hours. After in-core corrosion testing, some basic post-irradiation examinations (PIE) including weight change per unit area and surface profile as well as surface morphology were conducted in the MIT Nuclear Reactor Laboratory (NRL) and compared with out-of-reactor corrosion samples. The deep understanding of these alloys in high-temperature molten salt with strong neutron irradiation requires further microstructural analysis such as electron microscopy and energy dispersive spectroscopy on the surface and cross-section of the in-core corrosion tested samples. The challenge in the microstructural analysis is due to the high radioactivity from the gamma emitting isotopes formed in alloys during neutron irradiation. Additionally, the facilities that are authorized for radioactive materials are very limited on campus. In this study, the microstructure of the in-core corrosion tested Hastelloy N{sup R} and 316L stainless steel specimen was analyzed by using the facilities at the Center for Advanced Energy Studies (CAES) at Idaho National Laboratory (INL) through the Nuclear Science User Facilities (NSUF) Rapid Turnaround Experiment (RTE) program.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Security-Constrained Unit Commitment with Linearized AC Optimal Power Flow.

We propose a mathematical programming-based approach to optimize the security-constrained unit commitment problem with a full AC transmission network representation. Our approach is based on our previously introduced successive linear programming (SLP) approach to solving the non-linear, nonconvex AC optimal power flow (ACOPF) problem. By linearizing the ACOPF, we are able to leverage powerful commercial mixed-integer solvers to iteratively optimize the combined unit commitment plus ACOPF model. We demonstrate our approach on six-bus, IEEE RTS-96, and IEEE 118-bus test systems. We perform a comparative analysis of the relative impacts of singlebus, DC, and AC transmission network models on the unit commitment and dispatch solutions and their associated costs.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2022

Clean Condensing Gas Furnace

Natural gas furnaces are the most common space heating equipment in the U.S. residential and commercial building markets. However, current residential natural gas condensing furnaces generate substantial acidic condensate as well as significant emissions of sulfur oxides (SOx), nitrogen oxides (NOx), carbon monoxide (CO), hydrocarbons (HC), and methane (CH 4 ) contributing to environmental degradation of air, water, and soil. This report describes a novel solution to reduce the environmental impact of natural gas condensing furnaces based on the technology of a monolithic acidic gas reduction (AGR) catalyst for SOx trapping, NOx redox to nitrogen, and oxidation of formic acid, CO, HC, and CH 4 . The AGR technology offers the following benefits: (1) a neutral furnace condensate with a pH of ~7, allowing its safe release into the sewer system thus eliminating a second drainage system; (2) trapping and removing nearly all SOx emissions; (3) NOx emissions nearly at nearly 1-2 ng/J, more than 95% lower than new emissions standards in California; (4) the use of a low-cost heat exchanger as a condensing heat exchanger (HX) since the condensate is not acidic, avoiding the need for expensive stainless steel alloys; and (5) unburnt fuel energy recovery to boost efficiency.The AGR component and AGR-enabled furnace performance were broadly tested to determine their effects on long-term reliability and durability, as well as SOx storage and regeneration activity. The AGR regeneration does not impair the performance in achieving neutral condensate and ultra-low NOx emissions, and the AGR catalyst subjected to regeneration activities continued to function well and achieved slightly better annual fuel utilization efficiency (AFUE). The 400-hour reliability and durability test of the retrofitted condensing furnace with the AGR component shows that the furnace unit still achieves a neutral furnace condensate with a pH of ~7 and enables 0~3 ng/J of NOx emissions. However, the 400-hour operation slightly degraded the AFUE because of soot particle accumulation caused by frequent incomplete combustion owing to inappropriate condensate drainage during testing. Thus, proper condensate drainage is critical for AGR-enabled furnaces. Furthermore, neutron computed tomography was employed to survey the aged AGR component and demonstrate high-resolution 2D and 3D representations for the nondestructive diagnosis of the AGR component. The tomography showed that the AGR component did not deform or suffer broken AGR channels. A new AGR catalyst with low precious metal loading was preliminarily explored to identify a pathway of optimizing AGR material loading and maximizing acidic gas reduction at low cost. The new AGR component can reduce precious metal loading by 38% and still achieve neutral condensate and ultralow NOx emissions. The furnace with the AGR component of low Pt/Rh loading enables a maximum AFUE of 97%, which is meaningfully higher than the original furnace. Long-duration testing for the furnace enabled with the low precious metal loading AGR component will be vital in future research. Although the current work demonstrates a proof of concept for the AGR-enabled furnace, the AGR assembly needs to be optimized and integrated into the design of new OEM furnace products. Furthermore, the AGR technology can be applied not only for residential gas furnaces, but also for commercial rooftop units, gas heat pumps, gas-fired water heaters, combustion boilers, and other systems.

03 NATURAL GAS↗

DOE OSTI2016

Public Housing: A Tailored Approach to Energy Retrofits - Raleigh

More than 1 million U.S. Department of Housing and Urban Development-supported public housing units provide rental housing for eligible low-income families across the country. These units range from scattered single-family houses to high-rise apartments. In this project, the Advanced Residential Integrated Energy Solutions Collaborative (ARIES) worked with two public housing authorities (PHAs) to develop packages of energy efficiency retrofit measures the PHAs can cost-effectively implement with their own staffs in the normal course of housing operations at the time when units are refurbished between occupancies. ARIES conducted a survey of PHAs to assess their receptiveness to this concept and the applicability of the concept to PHA units. The results of the survey, to which more than 100 PHAs responded, support the proposed approach. The project consisted of a field evaluation in which energy audits were performed on a sample of PHA units at two housing authorities. Energy efficiency turnover protocols were developed for typical units, the protocol was implemented by PHA staff, and the effectiveness of the protocol was quantified through field testing and modeling. The energy efficiency turnover protocols emphasized air infiltration reduction, duct sealing, and measures that improve equipment efficiency. In the 10 housing units in which ARIES documented implementation, reductions in average air leakage of 16%-20% and duct leakage of 38% were obtained. Total source energy consumption savings was estimated at 6%-10% based on Building Energy Optimization modeling with a simple payback of 1.7-2.2 years. Implementation challenges were encountered, mainly related to required operational changes and budgetary constraints. Lack of complete training and inadequate quality control can prevent PHAs from effectively retrofitting units to their full potential. Nevertheless, despite these hurdles, simple improvements, such as caulking and sealing penetrations, windows, and doors; sealing duct boots; and adding pipe insulation into a standardized turnover protocol can feasibly be accomplished by PHA staff at low or no cost. At typical housing unit turnover rates, these measures could impact hundreds of thousands of units per year nationally. Islip Housing Authority - Single Story Typical home - Raleigh, NC Raleigh Housing Authority - 2 story typical home in Raleigh housing authority - Terrace Park

air filtration↗