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DOE OSTI2023

Complete Performance Comparison Between the Optimized Image Construction Algorithm (U-MBIR) and the Existing Reconstruction Algorithm for Detecting Defects and Damage in Concrete

Reinforced concrete (RC) is a composite material subjected to mechanical, thermal, and chemical loads throughout its service life. Because of these external stressors and the susceptibility of RC structural members to shrinkage and microcracking, the material degrades throughout its life cycle. This deterioration can lead to a decrease in member capacity and, ultimately, poses a threat to the structural integrity. Thus, it is crucial that the damage caused by aging and degradation be monitored and assessed at regular intervals throughout the material’s service life. Since coring of the material is typically not feasible for in-service structural systems, non-destructive evaluation (NDE) methodologies are used to assess remaining structural capacity. NDE methods enable surface and subsurface examination without damaging or degrading the medium. Moreover, RC is a critical component of nuclear power plants; thus, its safety and reliability must be thoroughly examined throughout the life cycle of the structural system. Ultrasonic measurements have been an industry standard for both surface and subsurface inspections. To this end, Oak Ridge National Laboratory (ORNL) has researched and developed advanced image reconstruction algorithms to capture internal damage. The results and discussion presented herein summarize the current state of the ultrasonic model–based iterative reconstruction (U-MBIR) algorithm developed at ORNL. More specifically, this report presents a comparison between reconstruction images produced via a widely employed ultrasonic NDE technique—the synthetic aperture focusing technique (SAFT)—and the ORNL-developed U-MBIR algorithm. These NDE methodologies are demonstrated using ultrasonic data collected from four concrete specimens. Overall, the U-MBIR algorithm eliminates artifacts and noise that are typically present within the SAFT reconstructions, and it shows defects and anomalies more clearly than the SAFT images. In conclusion, this algorithm is suitable for identifying concrete defects, although more improvements and optimization could be done to better define internal defects.

36 MATERIALS SCIENCE↗

DOE OSTI2021

In situ TEM observation of calcium silicate hydrate nanostructure at high temperatures

Fire poses a substantial threat to concrete structures because calcium silicate hydrate (C-S-H) is not thermally stable at high temperatures. Herein, in situ TEM at temperatures from 20 to 800 °C was conducted to study the thermal-induced structural changes in C-S-H. We captured C-S-H shrinks at an average rate of 0.02 μm{sup 2}/{sup o}C upon heating with three stages observed, including induction, constant, and rapid periods. Our observation revealed that an 800 nm pore could be healed during pore structure evolution owing to the reconstruction and deformation among C-S-H blocks. The Ca/Si ratio was dropped at higher temperatures because of the leakage of calcium ions from the C-S-H structure to form CaO precipitates. The temperature-driven phase transformation and degradation were also detected using electron diffraction that C-S-H was transformed into metastable calcium silicate minerals above 800 °C. This work provides insights into the nanoscale transformation of C-S-H at high temperatures.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Modeling Interfaces to Support Low-Level Waste Disposal System Performance Assessments - 20366

In low-level waste (LLW) disposal sites, interfaces between cementitious materials, used as waste forms and/or engineered barriers, and the surrounding soil or backfill material are often encountered. Reactions across these interfaces may lead to chemical and structural alteration of the cementitious and backfill materials that may ultimately affect long-term performance. For example, the ingress of carbon dioxide from soil gas into a waste tank concrete shell at the Hanford Site or Savannah River Site (SRS) or a low-level waste disposal vault at SRS leads to carbonation of cement hydration products (e.g. portlandite, calcium-silicate hydrate or C-S-H, and ettringite). The result of carbonation is a decrease in pH in the cement paste portion of the tank wall or vault concrete structure that could ultimately lead to de-passivation of the embedded structural steel and, eventually, to cracking. Cracking can lead to increased ingress of water into the structure, corresponding increased release of constituents of concern, and to subsequent increased transport of these constituents to the surrounding environment. The mobility of trace constituents (e.g., radionuclides of concern) in waste forms may increase in response to changes in pH, pore structure, and mineralogical gradients. Depletion of portlandite results in subsequent decalcification of C-S-H that leads to changes in the cement strength and may lead to structural failure. At the boundary between waste forms and concrete vaults, the migration of sulfate ions from the salt waste form (e.g., the saltstone waste form used for LLW disposal at the Savannah River Site in Saltstone Disposal Units, or SDUs) into the barrier or vault concrete has been predicted to result in expansive mineral phase formation (or 'sulfate attack'). Expansive mineral formation could result in cracking and potential loss of structural integrity in the vault concrete. In the performance assessment (PA) for the SRS SDUs, prediction of both the carbonation and sulfate attack fronts are critical to assessing the long-term performance of these LLW disposal vaults. A general purpose geochemical reactive transport model has been developed in LeachXS/ORCHESTRA to evaluate typical interfaces for LLW disposal environments, including SDUs at SRS. The model may be used to predict the long-term performance of interfaces between cementitious materials and saltstone or backfill materials with respect to primary phases by considering mass transport, cement chemistry, geochemical speciation, and multi-ionic diffusion over interfaces between different materials. The present study compares the performance and interactions between vault concrete and saltstone waste form for a representative SDU scenario. The materials were characterized using pH-dependence leaching test (EPA 1313) and semi-dynamic transport tests (EPA 1315) with test results used to parameterize the interface model. The development of a credible inter model provides an important basis for supporting future PAs where carbonation and sulfate attack are important aging and degradation mechanisms. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Electronically induced defect creation at semiconductor/oxide interface revealed by time-dependent density functional theory

Carrier induced defect creation at the semiconductor-oxide interface has been known as the origin of electronic device degradation for a long time, but how exactly the interface lattice can be damaged by carriers (especially low-energy ones) remains unclear. Here we carry out real-time time-dependent density functional theory simulations on concrete Si/SiO 2 interfaces to study the interaction between excited electrons and interface bonds. We show that the normal interface Si-H bonds are generally resistant to electrons due to the delocalized nature and high energy level of the Si-H antibonding states, and due to the high-energy barrier to break the Si-H bond. However, if an additional hydrogen atom exists by attaching to a nearby oxygen atom (forming a “Si-H···H-O” complex), the Si-H bond will be greatly weakened, including the reduction of energy barrier for bond breaking, and the lowering of the antibonding state energy level which favors electron injection. Together with the multiple vibrational excitation process, the corresponding Si-H bond can be broken much more easily. Thus we propose that the Si-H···H-O complex will be the center for defect creation and device degradation. Finally, we also explain why such a center might be relatively easy to form during the hydrogen annealing process.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2022

MELCOR Accident Progression and Source Term Demonstration Calculations for a FHR

MELCOR is an integrated thermal hydraulics, accident progression, and source term code for reactor safety analysis that has been developed at Sandia National Laboratories for the United States Nuclear Regulatory Commission (NRC) since the early 1980s. Though MELCOR originated as a light water reactor (LWR) code, development and modernization efforts have expanded its application scope to include non-LWR reactor concepts. Current MELCOR development efforts include providing the NRC with the analytical capabilities to support regulatory readiness for licensing non-LWR technologies under Strategy 2 of the NRC's near- term Implementation Action Plans. Beginning with the Next Generation Nuclear Project (NGNP), MELCOR has undergone a range of enhancements to provide analytical capabilities for modeling the spectrum of advanced non-LWR concepts. This report describes the generic plant model developed to demonstrate MELCOR capabilities to perform fluoride-salt-cooled high-temperature reactor (FHR) safety evaluations. The generic plant model is based on publicly-available FHR design information. For plant aspects (e.g., reactor building leak rate and details of the cover-gas system) that are not described in the FHR references, the analysts made assumptions needed to construct a MELCOR full-plant model. The FHR model uses a TRi-structural ISOtropic (TRISO)-particle fuel pebble-bed reactor with a primary system rejecting heat to two coiled tube air heat ex changers. Three passive direct reactor auxiliary cooling systems provide heat removal to supplement or replace the emergency secondary system heat removal during accident conditions. Surrounding the reactor vessel is a low volume reactor cavity that insulates the reactor with fire bricks and thick concrete walls. A refractory reactor liner system provides water cooling to reduce the concrete wall temperatures. Example calculations are performed to show the plant response and MELCOR capabilities to characterize a range of accident conditions. The accidents selected for evaluation consider a range of degraded and failed modes of operation for key safety functions providing reactivity control, the primary system decay heat removal and also a piping leak of the line to the coolant drain tank.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Quantum optimal control of superconducting qubits based on machine-learning characterization

Implementing fast and high-fidelity quantum operations using open-loop quantum optimal control relies on having an accurate model of the quantum dynamics. Any deviations between this model and the complete dynamics of the device, such as the presence of spurious modes or pulse distortions, can degrade the performance of optimal controls in practice. Here, we propose an experimentally simple approach to realize optimal quantum controls tailored to the device parameters and environment while specifically characterizing this quantum system. Concretely, we use physics-inspired machine learning to infer an accurate model of the dynamics from experimentally available data and then optimize our experimental controls on this trained model. We show the power and feasibility of this approach by optimizing arbitrary single-qubit operations in detailed numerical simulations of a superconducting transmon qubit. Furthermore, we demonstrate that this framework produces an accurate description of the device dynamics under arbitrary controls, together with the precise pulses achieving arbitrary single-qubit gates with a high fidelity of ∼99.99%.

Artificial neural networks↗

NASA NTRSDate not supplied

Geopolymer Lunar Concrete under Reduced-Pressure Curing and Vacuum Exposure

Geopolymer concrete has emerged as a promising material for constructing habitats, roads, landing pads, and other necessary infrastructure on the lunar surface. Proper in-situ resource utilization (ISRU) is an important aspect of any candidate lunar construction material. Geopolymer concrete is effective at ISRU, as the lunar regolith can be used in the “as found” condition and mixed with an alkaline solution, such as sodium silicate and sodium hydroxide, which could also be mostly produced in-situ. It is envisioned that any concrete material on the lunar surface would be deposited with additive manufacturing techniques and would require a temporary environmental enclosure to mitigate the harsh lunar environment during construction and curing, such as pressure and temperature. Data are needed to set a benchmark for the conditions to be achieved within the deployable environmental enclosure during printing, as well as the resultant material properties after the enclosure is removed. This novel study investigated the influence of reduced atmospheric pressures during the initial curing period on the compressive strength of geopolymer lunar concrete created with two types of lunar regolith simulant. The simulants used in the study represent the regolith for the lunar mare (OPRL2N) and lunar highlands (OPRH2N) regions. Results show a noticeable decline in strength with a reduction in pressure and visible degradation. After the deployable environmental enclosure is removed from the printed segment of infrastructure on the lunar surface, it would then be exposed to the harsh vacuum environment. To understand the importance of curing time required within the deployable enclosure before it is removed, samples were cast and cured for different lengths of time and then exposed to a vacuum chamber capable of achieving 100 mTorr. Control samples were also cast at the same time but left in a standard laboratory environment after curing. The weight change and compressive strength as a function of time in the vacuum chamber were monitored for up to 56 days. The results showed that exposure to the vacuum limits further reaction and strength gain in comparison to control samples left in the laboratory environment. This research is in strong alignment with the 2020 NASA Technology Taxonomy materials area (12.1) and the in-situ resource utilization area (7.1), as it provides necessary data on the geopolymer lunar concrete material to support safe and sustainable options for maintaining a human presence on the Moon.

Moon to Mars Planetary Autonomous Construction Tec↗

DOE OSTI2019

Innovative Approach to SCC Inspection and Evaluation of Canister in Dry Storage

This final report is summarized four years of work of awarded project number DE-NE0008442\IRP-15-9318, title as: “Innovative Approach to SCC Inspection and Evaluation of Canister in Dry” supported by Nuclear Energy University Program\Integrated Research Program, NEUP\IRP, where last year is an extension of no cost project. This research work is focused on the possible degradation of dry storage canister holding used nuclear fuel (UNF), due to exposure of marine atmospheric environment. The canister usually designed from a welded austenitic stainless-steel of types 304 or 316, encased in a concrete shell known as an overpack. This canister designed and licensed for 20 years with an extension period need to be determined, in addition 20 years or more. These casks are located throughout the country in Independent Spent Fuel Storage Installations (IFSFIs) often on the site close to nuclear power plants, that are often located in coastal or lake-side regions where a marine atmospheric salt air environment exist. As a result of several recent decisions, such as not pursuing Yucca Mountain as a long-term geologic repository storage facility for UNF and the cancellation of the construction of a reprocessing facility in the 1980s due to proliferation risks, interim dry storage facilities for long-term storage are vital option for the immediate future of nuclear energy technology. One of the primary concerns with respect to the long-term performance of the storage casks is the potential for corrosion initiation due to deliquescence of salts deposited on the canister surface as aerosols; in regions of high residual weld stresses, this may lead to localized stress corrosion cracking (SCC). Dust and aerosols in the air being drawn though ventilation openings in the overpacks of passively ventilated dry canister storage systems may be deposited on the stainless-steel canister outer surfaces. Under these conditions, localized corrosion attack can occur. Chlorides ions deliquescence on the canister surface, a process by which the salts essentially pull moisture out of the air, creating a concentrated salt brine on the canister surface. Chloride-induced stress corrosion cracking (CISCC) of welded zones is of special concern, as it a well-documented mode of attack for austenitic stainless steels marine environments, and many independent spent fuel storage installations (ISFSIs) are located in coastal areas. Recent canister inspections have shown that chloride salts are present on the surface of in-service canisters at near-marine settings. However, canister surface inspections of satisfactory resolution to detect SCC have never been carried out in details, because access to the canister surfaces through vents in the overpacks is extremely limited, and high radiation fields make removal of the canisters from the overpacks undesirable. Although no significant corrosion damage has not been detected up to date, there is concern that this may lead to pitting corrosion, which could in turn develop into stress corrosion cracking, causing canister structural failure. The initiation and propagation of pitting corrosion has been widely studied in full immersion conditions, but limited studies have been done on marine atmospheric conditions, which is a focus of this work, these includes laboratory and field experimentally and propagation rate modeling of pitting and transition to crack, and crack propagation rate. It is known, however, that the rate of atmospheric corrosion can be affected by variations in relative humidity (RH), salt density, salt type, droplet area, temperature and stainless-steel microstructure and residual stress particularly in welded zones of canisters. The integrated research program has identified potential deterioration mechanisms for the stainless-steel canisters containing the UNF in dry storage that require detailed research and investigation and will have an impact on the performance of long-term interim storage under the normal and extreme environmental conditions experienced during the duration of this storage. The prediction and monitoring of canister corrosion processes while in storage can provide important data for the assessment of interim storage performance and the safety to the public. This report comprises of mainly the published results in different journals and covering a wide range spectrum from susceptible stainless-steel types 304\316 microstructural analysis through the implant test at atmospheric environment, residual stress analysis of mockup canister and simulation, inspection methodology includes nonlinear ultrasound and vibro-thermography and stochastic modeling and simulation of pits propagation rates and transition to cracks. Description of field specimens installed at ISFSI of Main Yankee is given in the following chapters and will be analyzed in the coming future and serves as an important data for a stochastic analysis as it was drawn from the publications and reports described in this report. The purpose of this report is to bring together the project efforts that includes improving in our understanding of structural aging performance at current dry storage casks fleet at ISFSIs in the context of remedial maintenance and inspection applications technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

The Hanford Lead Canister Collaboration

The Hanford Site, Management of the Cesium and Strontium Capsules (MCSC) Project is preparing to place cesium and strontium material into dry storage using stainless steel canisters that are housed inside vertical concrete casks. Part of the site’s aging management program is to use a spare canister system as a leading indicator canister to provide advance warning for signs of chloride induced stress corrosion cracking (CISCC) or any other kind of canister degradation that might take place over the system’s 300-year design life. The Hanford Lead Canister (HLC) will use electric heaters to simulate the decay heat of cesium and strontium to ensure the HLC simulates the environment and conditions the capsule-loaded systems will experience. In addition to being part of the site’s aging management program, the HLC is being made available for research and development that advances the industry regarding canister mitigation and repair (M&R). The HLC system closely resembles current spent nuclear fuel (SNF) dry storage systems, making the HLC useful for demonstrating technologies or performing research that is relevant to SNF canisters . Several organizations are collaborating to prepare the HLC, along with technical strategy, to perform the needed research and development activities, to include the Pacific Northwest National Laboratory (PNNL), Central Plateau Cleanup Company (CPCCo), and the Electric Power Research Institute (EPRI). In one example of this collaborative research, EPRI is planning to use the HLC to demonstrate in-situ M&R technologies under controlled conditions before the HLC is deployed at Hanford. There is a period of several years when the HLC will be available for canister M&R activity before it is deployed. This paper describes the development of the HLC, its role in the Hanford site’s MCSC Project aging management program, and some of the research activities that are planned for the HLC.

Klymyshyn, Nicholas A.↗

DOE OSTI2020

Modeling the corrosion of steel casing and the damage of well cement in a borehole system

This work presents the effect of chloride-induced corrosion in an underground borehole system, specifically considering the steel casing and further damage to the surrounding cement. In the past four decades, oil well cement has been successfully used for underground construction. However, high concentrations of chloride ions gradually degrade the functionality and durability of well cement. Gasses and liquids can leak from deterioration caused by chloride ions, and thus motivate this study. The chemo-mechanical coupling diffusion model and a classical fracture mechanics model are applied. The model is validated with OPC concrete corrosion data from literature, and then applied to well cement using experimentally obtained material property data. The time when crack initiation, peak pressure, and complete fracture is reached in the model of the cement sheath provides a quantification of the expected service life of the borehole system.

36 MATERIALS SCIENCE↗

DOE OSTI2021

Development of a Reconstruction Methodology Based on X-Ray Computed Tomography to Generate Realistic 3D Concrete Microstructures in MOSAIC

Despite being passive components, concrete structures represent a major capital investment in nuclear power plants. However, some concrete components are critical for the safety and long-term operation of the reactors: the concrete containment building protects the reactors from external aggression. The concrete biological shield (CBS) contains the radiation exiting the reactor to protect equipment and personnel. Depending on the design, the CBS also has a function for supporting the reactor’s systems. This function needs to be ensured in-service and during accident scenarios. After several decades of use, the CBS is exposed to high-energy irradiation over an extended operation time, exceeding the fluence level currently considered detrimental for the concrete’s structural properties. Such conditions occur at different times depending on the reactor’s operation and design. Nevertheless, from the perspective of subsequent license application, it is mandatory to assess the potential effects of prolonged irradiation on concrete. Unlike steel components, concrete structures are built with local constituents (cement, sand, and coarse aggregates) to avoid prohibitive transportation costs. Thus, concrete performance is determined by the tolerance or the susceptibility of its constituents against any aging mechanism. The tolerance against irradiation of the CBSs may vary greatly among the 56 nuclear power plants and 93 nuclear power reactors that form the current US light-water reactor fleet. Irradiation-induced degradation is mainly a result of amorphization-induced expansions of the silicated minerals present in the aggregates. The neutron irradiation–induced expansion varies from one rock-forming mineral to another: well framework–structured silicates (e.g., quartz and feldspars) exhibit higher swelling than chained (e.g., pyroxene) or isolated silicates (e.g., garnet). Silicates make up nearly 90% of the Earth’s crust. Other common minerals present in construction aggregates are carbonates found, for example, in limestone. Irradiated carbonates exhibit only very minor expansions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1990

Experimental validation study of an analytical model of discrete frequency sound propagation in closed-test-section wind tunnels

The principal objective is to assess the adequacy of linear acoustic theory with an impedence wall boundary condition to model the detailed sound field of an acoustic source in a duct. Measurements and calculations are compared of a simple acoustic source in a rectangular concrete duct lined with foam on the walls and anechoic end terminations. Measurement of acoustic pressure for twelve wave numbers provides variation in frequency and absorption characteristics of the duct walls. Close to the source, where the interference of wall reflections is minimal, correlation is very good. Away from the source, correlation degrades, especially for the lower frequencies. Sensitivity studies show little effect on the predicted results for changes in impedance boundary condition values, source location, measurement location, temperature, and source model for variations spanning the expected measurement error.

Mosher, Marianne↗

NASA NTRS1979

Friction and wear characteristics of wire-brush skids

The testing technique consisted of towing the skids with a ground test vehicle over asphalt and concrete surfaces at ground speeds up to 80 km/hr (50 mph) and bearing pressures up to 689 kPa (100 psi) over sliding distances up to 1585 m (5200 ft). Results indicate that the friction coefficient developed by wire brush skids is essentially independent of ground speed, is slightly increased with increasing bearing pressure, is noticeably affected by surface texture, and is not degraded by surface wetness. Skid wear is shown to increase with increasing bearing pressure and with increasing ground speed and is dependent on the nature of the surface. Runway surface damage caused by the skids was in the form of an abrasive scrubbing action rather than physical damage.

Dreher, R. C.↗

DOE OSTI2021

Annual Status Report (FY 2020): Performance Assessment for the Disposal of Low Level Waste in the 200 East Area Burial Grounds

This annual review provides the projected dose estimates of radionuclide inventories disposed in the 200 East Area Low-Level Waste Burial Grounds (LLBGs) since September 26, 1988. These estimates are calculated using the original dose methodology developed in the performance assessment (PA) analysis (WHC-SD-WM-TI-730). The estimates are compared with requirements of DOE O 435.1 Chg 1 and performance objectives defined in companion documents DOE M 435.1-1 Chg 1and DOE-STD-5002-2017). All performance objectives are currently satisfied, and operational waste acceptance criteria (HNF-EP-0063) and waste acceptance practices continue to be sufficient to maintain compliance with performance objectives. Inventory estimates and associated dose estimates from future waste disposal actions are unchanged from previous years’ evaluations that indicate potential impacts well below performance objectives; therefore, future compliance with DOE O 435.1 Chg 1 is expected. A new PA study was initiated in fiscal year (FY) 2019 for evaluation of active disposal sites within the 200 East and 200 West Areas (Trench 94 in 200 East; Trenches 31 and 34 in 200 West) due to extended time elapsing between the current annual status report and the original PA for the active disposal sites. The new PA for the active disposal sites is expected to be completed in FY 2021. Within the active burial grounds in the 200 East Area, low-level waste and mixed low-level waste will continue to be disposed of in the dedicated U.S. Navy reactor compartment trench at the 218-E-12B Burial Ground (Trench 94). During this reporting period (FY 2020, from October 1, 2019, through September 30, 2020), two reactor compartments were disposed in Trench 94. Results from sorption experiments are summarized for this reporting period to quantify the efficacy of concrete waste forms in retaining key radionuclides (e.g., technetium-99 and iodine-129). The test durations ranged from 1 to 3 months. Continued groundwater monitoring of the 200 East Area LLBGs indicates no groundwater contamination due to LLBG waste. Current assumptions about future land use at the Hanford Site are consistent with PA analysis1 assumptions of a post-closure facility that will not be degraded by human activity. The LLBGs are located in an area identified for waste management and containment of residual contamination. This area will remain after final environmental remediation and the proposed shrinkage of Hanford Site boundaries to small sections within the 200 East and 200 West Areas in the Central Plateau (DOE/EIS-0391). The current closure plan for the LLBGs (DOE/RL-2000-707) estimates that the 200 East LLBGs will be closed in the 2050 timeframe. The Disposal Authorization Statement, other technical basis documents, and the radioactive waste management basis are of continued adequacy to meet the performance objectives of DOE O 435.1 Chg 1. Overall, there are no substantive changes to primary PA assumptions nor the PA analysis conclusion; therefore, compliance with DOE O 435.1 Chg 1 and the Disposal Authorization Statement is maintained.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Evaluation of H-Canyon Ventilation Exhaust Tunnel Inspection at the Savannah River Site - 20052

The Department of Energy (DOE) H-Canyon Facility at the Savannah River Site (SRS) is operated by Savannah River Nuclear Solutions (SRNS). The Facility Structural Integrity Program [1] provides a process to inspect, evaluate, and document the conditions of passive safety related structural systems and components, their degradation mechanisms and their impact on the safety envelope. The Structural Integrity Program performs periodic visual inspections to confirm the systems and components can perform their safety function and recommend corrective actions before the function is compromised. The H-Canyon Exhaust Tunnel is part of the H-Canyon Ventilation System and is periodically inspected under the Structural Integrity Program. The Tunnel performs a Safety Class passive design feature function that is available 100% of the time. The underground reinforced concrete Tunnel directs Canyon process air from the Canyon to the Sand Filter System. The radiological and chemical airborne activity coupled with the physical configuration of the tunnel precludes a manned entry into the tunnel to perform the inspection. Unlike active systems that can be subjected to physical testing to confirm capability, qualification calculations, including modeling, is performed to assess passive design feature capabilities relative to operational, accident and natural phenomena hazards events. Input for these calculations come from knowledge of field conditions, design drawings and conservative assumptions. The knowledge of field conditions comes from concrete and soil testing activities as well as inspection data. The use of cameras on crawlers to obtain inspection data and how that data is used in qualification calculations is the focus of this discussion. Remote Tunnel inspections have been performed using cameras on a stick, referred to as pole cameras, prior to Calendar Year 2003 (CY2003). Since CY2003, six camera equipped crawlers have been used. Crawlers have evolved and improved by incorporating lessons learned from the previous inspections. The latest 2019 crawler successfully traveled the entire length of the tunnel between the Canyon and Sand Filter System and captured images that were previously unseen, thus providing key visual data for the Structural Integrity Program report [2]. The inspection data serves as 1) input to qualification calculations, 2) validates current conditions remain bounded by qualification calculations, and 3) documents a record of change over time. This data is vital for the effective evaluation of the tunnel structure and aids with predicting the life cycle service life safety envelope to protect the public, environment, and facility worker. The goal of the 2019 deployment was to use a more robust vehicle to ensure travel over obstacles, perform the inspection with higher resolution cameras, and have the ability to elevate the cameras to view surfaces previously obscured by abandon-in-place ducts in the tunnel. The inspection information will serve to strengthen the input basis for qualification calculations, versus using excessive conservative calculation assumptions. This 2019 inspection was compared with previous crawler visual evidence to identify change over time to project remaining service life and confirm field conditions remain bounded by qualification calculations. The results of the 2019 Exhaust Tunnel inspection continue to improve and provide a firm basis to make sound engineering judgements regarding tunnel capability and approaches in projecting remaining service life with an aging system and component. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Challenges to detection: Humidity as a spur to chemical agent change

Detecting chemical agents in outdoor environments such as a battlefield is made challenging by not only the spurious signatures from background chemicals and surfaces (e.g. asphalt, dirt, concrete), but also by the chemical transformation of the actual agents. The change of CW agents to other species can be catalyzed by other chemicals present in the scene, by different substrates, as well as by local weather conditions. Some of the final environmental transformation products are known (e.g. for the G agents methylphosphonic acid), but many of the intermediate chemical states are not, nor are the rates of transformation to the other intermediates or the end products. In this study we have made preliminary optical investigations into the degradation products of a G-agent intermediate, namely methylphosphonic anhydride and its rate of conversion to the more stable methylphosphonic acid. Using infrared and Raman spectroscopies, we have found that the relative humidity (RH) greatly affects the rate of change and we report first results from these studies.

: chemical warfare agent detection, Methylphosphon↗

DOE OSTI2022

Critical problems of energy frontier Muon Colliders: optics, magnets and radiation

This White Paper brings together our previous studies on a Muon Collider (MC) and presents a design concept of the 6 TeV MC optics, the superconducting (SC) magnets, and a preliminary analysis of the protection system to reduce magnet radiation loads as well as particle backgrounds in the detector. The SC magnets and detector protection considerations impose strict limitations on the lattice choice, hence the design of the collider optics, magnets and Machine Detector Interface (MDI) are closely intertwined. As a first approximation we use the Interaction Region (IR) design with beta-star=3 mm, whereas for the arcs we re-scale the arc cell design of the 3 TeV MC. Traditional cos-theta coil geometry and Nb3Sn superconductor were used to provide field maps for the analysis and optimization of the arc lattice and IR design, as well as for studies of beam dynamics and magnet protection against radiation. The stress management in the coil will be needed to avoid large degradation or even damage of the brittle SC coils. In the assumed IR designs, the dipoles close to the Interaction Point (IP) and tungsten masks in each IR (to protect magnets) help reducing background particle fluxes in the detector by a substantial factor. The tungsten nozzles in the 6 to 600 cm region from the IP, assisted by the detector solenoid field, trap most of the decay electrons created close to the IP as well as most of the incoherent electron-positron pairs generated in the IP. With sophisticated tungsten, iron, concrete and borated polyethylene shielding in the MDI region, the total reduction of background loads by more than three orders of magnitude can be achieved.

43 PARTICLE ACCELERATORS↗

NASA NTRS1990

Evaluation of discrete frequency sound in closed-test-section wind tunnels

The principal objective of this study is to assess the adequacy of linear acoustic theory with an impedance wall boundary condition for modeling the detailed sound field of an acoustic source in a duct. This study compares measurements and calculations of a simple acoustic source in a rectangular concrete duct lined with foam on the walls and anechoic end terminations. Measuring acoustic pressure for 12 wave numbers provides variation in frequency and absorption characteristics of the duct walls. The cases in this study contain low frequencies and low wall absorptions corresponding to measurements of low-frequency helicopter noise in a lined wind tunnel. This regime is particularly difficult to measure in wind tunnels due to high levels of the reverberant field relatively close to the source. Close to the source, where the interference of wall reflections is minimal, correlation is very good. Away from the source, correlation degrades, especially for the lower frequencies. Sensitivity studies show little effect on the predicted results for changes in impedance boundary condition values, source location, measurement location, temperature, and source model for variations spanning the expected measurement error.

Mosher, Marianne↗