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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Low-Level Waste Minimization and Management Planning at the Western Waste Management Facility, Ontario Power Generation - 19235

The purpose of this study was to examine the strategies intended to mitigate the temporary storage restrictions for Low-Level radioactive waste at the Western Waste Management Facility to enable continued accommodation of waste generation rates, and to recover storage capacity from 2018-2023 to optimize long-term sustainability for radioactive waste management. The management options considered were based on existing waste processing technologies and programs available at the Western Waste Management Facility, in combination with third party volume-reduction services. Completed analysis measured the consumption of Low-Level Waste storage space as influenced by the projected volumes from nuclear generating stations, and the various processing scenarios developed; such as reducing the volume of surplus processable waste on-site, and the reduction of specific waste streams previously considered to be non-processable. Results from the quantitative analysis indicated which strategies will be most effective for minimizing waste volumes for long-term storage and for recovering low-level storage space at the Western Waste Management Facility. The strategic path forward is presented in short- to long-term management objectives. Results from this study will contribute to the waste stream planning and management of major reactor refurbishment projects and to the technological advancements and innovation of radioactive waste processing and minimization. With further understanding of the refurbishment-specific waste types and volumes, it is expected that Nuclear Waste Management and the nuclear generating stations will achieve increasingly improved minimization results with the progression of each refurbished reactor. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Developing Waste Characterization Capabilities in Support of the Near Surface Disposal Facility - 19245

Upon approval from the Canadian Nuclear Safety Commission, Canadian Nuclear Laboratories (CNL) Proposed Near Surface Disposal Facility (NSDF) to be constructed at CNL's Chalk River Laboratories will be the first permanent disposal facility in Canada for the disposal of low level radioactive waste. The NSDF is currently in the approval process and as such, a preliminary Waste Acceptance Criteria (WAC) has been implemented for the proposed facility. Waste currently generated at CNL is being characterized to verify that it meets the WAC and will be 'disposal ready' once the NSDF is operational. A robust waste characterization program is required for the following reasons: 1. Improved waste segregation to support waste minimization and recycling to ensure non-radioactive waste is reliably segregated and removed from radioactive waste streams destined for the NSDF. This includes characterization and discrimination of Naturally Occurring Radioactive Material (NORM) and anthropogenic radioactive material. 2. To ensure low level waste (LLW) is properly characterized as per the limits stipulated within the WAC to ensure criteria is met for NSDF packaging requirements 3. To ensure intermediate level waste is reliably segregated from NSDF destined wastes streams The purpose of this paper is to provide a review of the development of a waste characterization program strategy at Chalk River Laboratories and applicability of various waste streams; including consideration of important factors such as measurement methods, techniques, and the management of data. Emphasis is necessarily placed on non-destructive gamma spectrometry systems which are central to CNL's approach. A review of near-term developments is also provided. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Operational Safety Analysis for Disposal Period of the Future Belgian Geological Disposal Facility - 19249

For more than 40 years, ONDRAF/NIRAS (the federal agency responsible for managing radioactive waste and enriched fissile materials in Belgium) has been studying the concept of a geological disposal facility (GDF) in poorly indurated clays as an option for the long-term management of both category B waste (low-level and intermediate-level long-lived waste - LILW-LL) and category C waste (high-level waste - HLW and spent fuel). In the absence of a policy decision, the proposed Belgian reference option for disposal of such waste is a GDF in poorly indurated clays, located at a depth between 200 m and 600 m. ONDRAF/NIRAS has developed a reference disposal concept for research, development and demonstration (RD and D) purposes. At the current stage of this program, the primary objective of ONDRAF/NIRAS is to establish the safety and feasibility of constructing, operating and closing a Geological Disposal Facility (GDF) at the conceptual design level, without any presumption about the exact location of the facility. This will be evaluated in a first version of the safety and feasibility case (SFC-1 V1) for the geological disposal of category B and C radioactive wastes. Analysis of the operational safety is an important part of the SFC-1 V1. ONDRAF/NIRAS, supported by BGE Technology GmbH, has prepared an operational safety analysis of the current GDF reference design and the processes occurring during the operation and closure period. A SWIFT methodology (Structured What If Technique) is used to identify risks and the measures required to mitigate their consequences. In parallel to this SWIFT analysis, specific attention is also given to fire risk, especially in the blind-ending disposal galleries (DG). Based on understanding of fire risk analysis, recommendations suggest using exclusively electrically driven equipment to reduce fire loads and minimize the fire hazard. In addition, strict working procedures could prevent the presence of workers between the DG dead end and equipment presenting significant fire load. The use of mobile refuge chambers should also be part of the safety concept. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Performance Assessment for Closure of the Calcined Solids Storage Facility at the INL Site - 19289

The Calcined Solids Storage Facility is composed of six discrete stainless steel 'bin sets' at the Idaho Nuclear Technology and Engineering Center at the INL Site. Approximately 4,440 m{sup 3} of the dry, granular calcine waste is stored in the facility. Closure of the facility will entail removal of the calcine waste followed by grouting. A 5-cm thick residual amount of calcine was assumed to remain at the bottom of the bins. Important radionuclides in the residual layer for the groundwater pathway were identified as the fission and activation products Tc-99, Se-79, I-129, Cs-135, H-3, and actinides Pu-239, and Pu-240. The stainless-steel bins prevent near-term release of grouted residual calcine waste to the environment. Future releases were conservatively based on the corrosion rate of the weld seams on the stainless-steel bins, which site-specific studies measured mean tunnel corrosion rates of 0.0132 mm/yr with a coefficient of variation of 0.333. Based on tunnel corrosion rates, mean bin hydrologic containment times were estimated to be between 238 years and 1,191 years, depending on the thickness of the bin. Any breach in the weld seam was assumed to result in water flow and radionuclide leaching through the bin. The structural integrity of the bin was assumed to be controlled by the general stainless-steel corrosion rate, which was much slower (3.5 x 10{sup -5} mm/yr), resulting in structural integrity times between 24,000 years to over 100,000 years. A mixing-cell model was used to compute water flow and transport through the hydrologically failed bins and the vadose zone. Credit was taken for the geochemical properties of the grout used to fill the bins at closure, which significantly slows the releases from the bins for key radionuclides like Tc-99. A semi-analytical model for three-dimensional dissolved-phase transport in an aquifer of constant unidirectional flow, infinite length and width, and finite thickness was used to compute concentrations in the aquifer across the model domain for a time varying source. Cumulative impacts from all bin sets were calculated by summation across the bin set sources that were superimposed on a uniform Cartesian grid. Effective doses were calculated for direct groundwater ingestion from wells located 100-m downgradient from each of the bin sets and ingestion of meat, milk, and vegetables irrigated with water derived from the well. Intruder doses were based on a well drilled through the grouted waste and cuttings spread around a home site after a 500-year institutional control period. Peak doses during the 1,000-year compliance period were less than 5% of the 250 μSv/yr (25 mrem/yr) performance objective and intruder doses were generally less 5% of the of the 5000 μSv/yr (500 mrem) and 1000 μSv/yr (100 mrem) performance objective for acute and chronic intruder exposure, respectively. Groundwater concentrations were all less than the maximum contaminant limits. In all cases, doses were highest for the oldest bin set because, compared to the other bin sets, it had the highest radionuclide concentrations in the calcine waste. The performance assessment results indicate that performance objectives can be met leaving considerably more than 5-cm of residual calcine waste in the bin sets. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Advanced Modeling of Contaminant Fate and Transport in Soil and Groundwater from F-Canyon and the Central Laboratory Facility

F-Canyon (221-F) was constructed in the early 1950's and, together with FB Line, began operations in 1954 as the main processing facilities in FArea of SRS. The primary operations conducted in F-Canyon and FB Line included the separation and recovery of Pu-239 and U-238 from irradiated materials and the stabilization of plutonium residues. The F/H Laboratory Complex consists of buildings 772-F, 772-1F, and 772-4F which began operations in 1954, 1987, and 1993, respectively. The primary mission of the F/H Laboratory Complex for the last 45 years has been to support the chemical separations processing activities at Building 221-F and H-Canyon. The objective of this study is to determine the radionuclide and hazardous material holdup in F-Canyon and its support facilities, and to evaluate In-Situ Disposal (ISD) alternatives for deactivation and decommissioning (D and D) of these facilities to ensure groundwater protection and environmental compliance with the South Carolina Department of Health and Environmental Services (SCDHES) and the United States Environmental Protection Agency (USEPA). GoldSim is a dynamic probabilistic simulation software that allows modeling of chemical mass transport through engineered and environmental systems For this study, it will model the potential future release of radiological contamination through the vadose zone, Upper Three Runs (UTR) Aquifer and the Upper Three Runs Creek. Using a GoldSim model developed for building 235-F D and D, contaminant inventories as well as building geometry were updated to create the basis of the new F-Canyon model. The inventories from F-Canyon, FB Line, and the 800 Series Tanks (Tanks 804, 808, and 809) as well as F/H labs (including buildings 772-F, 772-1F, and 772- 4F) were used to make a total of four sources. The concentrations of total uranium and beta-gamma are within compliance for the study, while radium and alpha concentrations exceed their MCL's around 30,000 years and total effective water dose exceeds its MCL around 1,200 years. This suggests that some level of action will be needed to meet SCDHES and USEPA requirements. This evaluation was conducted only for the no-action alternative. Consideration and modeling of other alternatives will be required in the future to determine the best ISD alternative for D and D of F-Canyon and F/H labs with regard to groundwater protection, worker safety, and financial cost.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Panel Session 46: Final Closure of Unlined LLW Disposal Facilities Worldwide (R3.5)

This panel discussed the on-going and emerging issues with the design and construction of final cover system of unlined disposal facilities (e.g., material disposal areas (MDAs), subsurface disposal areas (SDAs), burial grounds, and similar historical waste trenches, pits and vaults). The discussion focused on key technical and regulatory issues as well as lessons learned from years of engineering and construction in US and non-US locations. Panelists with presentations: Closure of Unlined LLW Disposal Sites (Craig Benson); Closure of LLW Facilities (Paul Pointon); Final Closure of Unlined LLW Disposal Facilities - Lessons Learned (Kevin Foye)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Operations and Recent Upgrades to a Tritium Recycling Facility - 20375

Nuclear Sources and Services Inc. (NSSI) is a waste handling treatment company that has been operating since 1971. In 2000, NSSI began building a 1,000 m{sup 2} (∼10,000 sqft) tritium recycling facility on the Houston, Texas premises to address a growing need in the waste disposal market. Over the years, the capability of the facility grew to address many different waste streams with the goal of recycling > 99.9% of all tritium activity back into the fusion community. This is achieved by maintaining an effluent stack release well below the 150 GBq/year (C Ci/year) operating limit and maintaining all liquid and solid waste disposal below the low-level radioactive release limit of 0.C MBq/kg (10 μCi/kg). The unique processing systems in the facility attracts many challenging contracts. Within the past two years, NSSI has engaged with customers to treat; organic mixed wastes that simultaneously contain tritium and carbon-14, safely extract tritium from large quantities of exit sign lights, and remove tritium from uncharacterised legacy depleted uranium beds from a superfund site. This paper will outline the initial testing, the system upgrades to treat these varied streams, and the progress in processing the feedstocks provided by these projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Nuclear Science User Facilities Irradiation Capabilities at Oak Ridge National Laboratory

The Nuclear Science User Facilities (NSUF) within the Department of Energy (DOE) Office of Nuclear Energy offer researchers the opportunity to access world-class nuclear facilities at no cost to the user. Oak Ridge National Laboratory (ORNL) offers users access to the High Flux Isotope Reactor (HFIR) in addition to a wide variety of post-irradiation examination facilities. The ORNL irradiation engineering team has compiled a large number of successful designs for irradiating nuclear fuel and material specimens. NSUF users can leverage these existing designs to significantly reduce the time, cost, and risk of failure when designing new irradiation experiments. This work summarizes some of the irradiation experiment designs and capabilities that have been developed at ORNL in the hope that this information will help improve future NSUF proposals and inspire new ideas for irradiations that can be performed in the HFIR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Counterpart Test on IBLOCA of 13% Cold Leg Break in ATLAS Facility

In the analysis for Loss-of-Coolant Accident (LOCA) as a design-basis accident (DBA) of a pressurized water reactor (PWR), redefinition of the break size based on the risk-informed regulation (RIR) is being extensively investigated due to the potential safety benefits. The pipe integrity and break frequency has been found to depend on the pipe size, so that a complete rupture of a smaller pipe such as a pressurizer surge-line or an emergency core cooling system (ECCS) nozzle is more probable than a double-ended guillotine break (DEGB) of primary coolant circuit piping. The scenario with the break of these pipes can be regarded as intermediate break (IB) LOCA. With an aim of expanding code validation to address safety issues of the IBLOCA and developing an evaluation model (EM) for full spectrum LOCA, the database for the IBLOCA test in an integral effect test (IET) facility is essential. In particular, an experimental data from the counterpart test can contribute to enhance our understanding on the phenomena and resolve the scaling issues in the scaled-down test facilities. The objective of this study is to investigate thermal hydraulic phenomena during the IBLOCA by the counterpart test for the 13% cold leg break in ATLAS facility, which was the A.2 test in the OECDATLAS project. This integral effect test would contribute to evaluate the scaling effect of the ATLAS design and test condition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2010

ATR National Scientific User Facility 2009 Annual Report

This report describes activities of the ATR NSUF from FY-2008 through FY-2009 and includes information on partner facilities, calls for proposals, users week and education programs. The report also contains project information on university research projects that were awarded by ATR NSUF in the fiscal years 2008 & 2009. This research is university-proposed researcher under a user facility agreement. All intellectual property from these experiments belongs to the university per the user agreement.

36 MATERIALS SCIENCE↗
DOE OSTI2011

MANAGING BERYLLIUM IN NUCLEAR FACILITY APPLICATIONS

Beryllium plays important roles in nuclear facilities. Its neutron multiplication capability and low atomic weight make it very useful as a reflector in fission reactors. Its low atomic number and high chemical affinity for oxygen have led to its consideration as a plasma-facing material in fusion reactors. In both applications, the beryllium and the impurities in it become activated by neutrons, transmuting them to radionuclides, some of which are long-lived and difficult to dispose of. Also, gas production, notably helium and tritium, results in swelling, embrittlement, and cracking, which means that the beryllium must be replaced periodically, especially in fission reactors where dimensional tolerances must be maintained. It has long been known that neutron activation of inherent iron and cobalt in the beryllium results in significant {sup 60}Co activity. In 2001, it was discovered that activation of naturally occurring contaminants in the beryllium creates sufficient {sup 14}C and {sup 94}Nb to render the irradiated beryllium 'Greater-Than-Class-C' for disposal in U.S. radioactive waste facilities. It was further found that there was sufficient uranium impurity in beryllium that had been used in fission reactors up to that time that the irradiated beryllium had become transuranic in character, making it even more difficult to dispose of. In this paper we review the extent of the disposal issue, processes that have been investigated or considered for improving the disposability of irradiated beryllium, and approaches for recycling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Entrainment of Juvenile and Adult American Shad at a Pumped Storage Facility

The American Shad Alosa sapidissima has been targeted for restoration to the upper Susquehanna River, and entrainment losses at hydroelectric facilities on the river, including the Muddy Run Pumped Storage Facility, are of concern for the potential growth of the American Shad population. Based on the integration of pumping volume, time of entrainment, and literature–reported diel emigration of juvenile American Shad, the entrainment rate (N = 145; 53 exposed to pumping) was estimated at 3.5%. The entrainment rate for adults (N = 507) was estimated at 0.3–3.9%. Using multistate mark–recapture models, the estimated entrainment probabilities (Ψ) of radio–tagged juveniles were higher during periods of extended pumping (>1 h; Ψ = 0.093; 95% confidence interval [CI] = 0.054–0.156) than at periods of no pumping or short–duration pumping (<1 h; Ψ < 0.0001; 95% CI = 0.0–0.001); the high extended pumping occurred between 2300 and 0600 hours. Entrainment probabilities for adults were low (Ψ < 0.02); the highest probability occurred for fish detected downstream of the intake during the peak portion of the run (Ψ = 0.015; 95% CI = 0.004–0.047). Entrainment probability for most adult fish did not differ from zero. The low values of Ψ for both life stages were attributed to (1) the deep location of the intake (intake ceiling >11.7 m below the water surface), (2) the surface orientation of American Shad (upper 3.1 m), (3) the low overlap between high–volume pumping and peak emigration/migration times, (4) the pumping volume relative to prevailing river flows, and (5) the prolonged, robust swimming speed of American Shad, particularly that of adults (>2.2 m/s), which exceeded the intake velocity (0.2–0.9 m/s). Entrainment of juveniles increased with co–occurrence of low incoming river flows, high pumping volume, and peak emigration times. Furthermore, quantification of migratory species’ entrainment at pumped storage facilities requires integration of diel migration/emigration times with the frequency, timing, and duration of pumping volume.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Acoustic Exposure to Turbine Operation Quantifies Risk to Bats at Commercial Wind Energy Facilities

ABSTRACT Turbine‐related bat mortality at commercial wind energy facilities may threaten populations of migratory tree‐roosting bat species in North America. Industry stakeholders and regulatory agencies alike are investigating strategies to reduce risk of population‐level consequences as the wind energy industry grows. Bats collide with turbines only when turbine rotors are spinning and curtailing turbine operation at low wind speeds can effectively reduce bat fatality rates. Nonetheless, few quantitative data exist to determine appropriate threshold wind speeds below which turbine operations should be curtailed. Carcass monitoring is labor‐intensive and does not provide information on factors linked to bat fatality rates on any scale finer than nightly. We tested whether acoustic bat data recorded at turbine nacelles could provide a more precise and sensitive measure of fatality risk to bats by analyzing acoustics, weather, turbine operation, and carcass data collected at 2 commercial wind energy facilities in West Virginia over 7 years. Each wind facility implemented several distinct curtailment treatments during our study, allowing us to compare fatality rates and acoustic bat activity across multiple operational strategies. We found that bat passes exposed to turbine operation explained close to 80% of the variation in carcass‐based estimates of bat fatality rates and accounted for significant variation in raw carcass counts per turbine and probability of finding bat carcasses during individual turbine searches. Conversely, bat activity occurring when turbines were not operating had little or no relationship to fatality rates. We also found that patterns in bat activity exposure could be predicted accurately among turbines and years. Our results demonstrate that measuring exposure of acoustic bat activity provides a quantitative basis for designing, evaluating, and adaptively managing curtailment strategies. This is an important advance towards using curtailment to reduce bat fatality rates strategically while allowing for increased generation of renewable energy. © 2021 The Authors. Wildlife Society Bulletin published by Wiley Periodicals LLC on behalf of The Wildlife Society.

17 WIND ENERGY↗
DOE OSTI2018

Narrowing transient testing pulse widths to enhance LWR RIA experiment design in the TREAT facility

The Transient Reactor Test (TREAT) Facility provided thousands of transient irradiations and plays a crucial role in nuclear-heated safety research. The facility’s flexible design and multi-mission nature saw historic experiments for numerous reactor fuels and transient types but was never specifically adapted to address very-brief pulse transients akin to postulated Light Water Reactor (LWR) Reactivity Initiated Accidents (RIA). Since the behaviors of fuel under these conditions depends strongly on energy input duration and resulting cladding time/temperature response under pellet cladding mechanical interactions, this three-year project was conceived to investigate new pulse-narrowing capabilities. Kinetic models showed incremental improvements for minor facility enhancements including increased reactivity step insertions (to initiate the power pulse) and slightly-increased transient rod drive speed for pulse termination (“clipping”). Replacing peak fuel assemblies, repositioning non-transient control rods to hold down the “hot side” of the core, and balancing against required excess reactivity needs, the limiting fuel assembly power can be reduced by ~20 %. Furthermore, this is a remarkable discovery of latent capability in TREAT, not only in enabling the subject capabilities for reduced pulse widths, but also significantly “uprating” the core’s transient energy capacity. Incremental improvements can likely enhance TREAT’s capability into BWR-relevant missions; briefer PWR pulse shapes can only be achieved with an advanced clipping system. Helium-3 ( 3 He) was found to offer the greatest benefits for clipping design and overall performance. A unique concept showed great promise for enabling a 3 He-based system with a reasonable cost; this design concept will likely become the focal point of future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Design of an experimental facility with a unit cell test section for studies of the lower plenum in prismatic high temperature gas reactors

A proposed design for a Generation IV reactor is the very-high temperature reactor (VHTR), which employs helium as the primary coolant. Although many advantages exist with this design, a number of challenges remain. One such area of concern is the turbulent mixing of non-isothermal flows in the lower plenum, where an array of coolant jets collectively mix and traverse around a series of structural support posts before exiting the core. Neighboring jets in the lower plenum could experience temperature differences as high as 300–400 K, causing potential hot streaking in the support posts. Additional problems could occur further downstream if the level of non-uniformity in the temperatures is too high (e.g, inducing thermal stresses on gas turbine blades). Therefore, it is important that adequate tools exist to predict the turbulent mixing within the VHTR lower plenum. This task remains difficult to achieve due to the complex geometry and flow physics that dictate the mixing. Various modeling approaches can only be assessed using data acquired from a carefully designed experimental facility. This paper presents details of such a facility, where high-fidelity measurements can be acquired for a scaled portion of the lower plenum defined herein as a ‘‘unit cell”, which consists of a hexagonal array of jets whose flow is directed into a test section with support posts positioned between neighboring jets, and subjected to a crossflow. Through adjusting velocity and temperature conditions at the jet inlets, different locations within the lower plenum can be experimentally simulated. System response quantities (SRQs) include both temperature and velocity measurements, captured via multiple thermocouples embedded in each post, and particle image velocimetry (PIV), respectively. These SRQs provide assessment metrics for both the momentum and energy models present in current computational solvers, and therefore the experimental facility is a valuable source for validation data. Another benefit is a more comprehensive understanding of the thermal-hydraulic loading conditions and flow physics present in the VHTR lower plenum.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Analysis of an optical imaging system prototype for autonomously monitoring zooplankton in an aquaculture facility

Traditional approaches to biomonitoring in aquatic systems, such as sample collection, sorting, and identification, require significant time and effort, thereby limiting the spatiotemporal resolution of sample collection. Additionally, collection and preservation of samples for subsequent taxonomic identification and enumeration leads to mortality of organisms. Recent advances in technologies that utilize optical imaging and machine learning have provided new opportunities to expedite biomonitoring and lead to significant cost savings. These technologies can be advantageous to scientists or managers that conduct routine biomonitoring to inform operations, as in the case of aquaculture facilities. The Small Aquatic Organism optical imaging system (SAO) is a high-throughput optical imaging and classification prototype system that relies on computer vision and machine learning (Support Vector Machines, or SVMs) to autonomously identify and enumerate aquatic organisms. The SAO provides a more sustainable method of collecting large volumes of data and has the benefit of being used in situ. In this study, we tested the performance of the SAO in providing comparable results to manual zooplankton community monitoring in ten ponds at an aquaculture facility. We performed a side-by-side study comparing the sampling methods of plankton tow nets, where major zooplankton taxonomic classes were manually identified and enumerated, to sampling with the SAO. Vouchered samples were used to develop a training library for the SAO, where classes consisted of water boatman and zooplankton groups: cladocerans, copepod adults, copepod nauplii, and rotifers. SAO imagery was manually classified and compared with predicted results for validation. Accuracy for the SVM classifier of the SAO was 37.4 %. Convolutional Neural Networks (CNN) and Random Forest classifiers were also applied to SAO imagery and image features for comparison. The best CNN model and our Random Forest model had accuracies of 80.4 % and 46.6 % respectively. Challenges faced included the small size of copepod nauplii and rotifers and the limited resolution of the imaging camera, although there are tradeoffs between imaging resolution and the sample processing rate. Furthermore, our comparison shows that advancement in both optical imaging and ML are needed in order for the SAO prototype to yield comparable results to manual community monitoring in an aquaculture facility.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2010

Fusion materials irradiations at MaRIE's fission fusion facility

Los Alamos National Laboratory's proposed signature facility, MaRIE, will provide scientists and engineers with new capabilities for modeling, synthesizing, examining, and testing materials of the future that will enhance the USA's energy security and national security. In the area of fusion power, the development of new structural alloys with better tolerance to the harsh radiation environments expected in fusion reactors will lead to improved safety and lower operating costs. The Fission and Fusion Materials Facility (F{sup 3}), one of three pillars of the proposed MaRIE facility, will offer researchers unprecedented access to a neutron radiation environment so that the effects of radiation damage on materials can be measured in-situ, during irradiation. The calculated radiation damage conditions within the F{sup 3} match, in many respects, that of a fusion reactor first wall, making it well suited for testing fusion materials. Here we report in particular on two important characteristics of the radiation environment with relevancy to radiation damage: the primary knock-on atom spectrum and the impact of the pulse structure of the proton beam on temporal characteristics of the atomic displacement rate. With respect to both of these, analyses show the F{sup 3} has conditions that are consistent with those of a steady-state fusion reactor first wall.

43 PARTICLE ACCELERATORS↗
DOE OSTI2023

Development of experimental and computational frameworks to predict subcooled flow boiling in the LANL Isotope Production Facility

Cooling is crucial to maintain the integrity of target systems in isotope production facilities. At Los Alamos National Laboratory (LANL)’s Isotope Production Facility (IPF), multiple encapsulated targets are stacked and irradiated in tandem with a 100 MeV, ~250μA proton beam. To facilitate effective heat removal, these stacked targets are separated and cooled via a series of water channels. At these beam currents, this high-energy proton beam heats the target system, likely initiating subcooled flow boiling in the cooling channels. However, in-beam monitoring of the IPF target system is not possible due to the extreme radiation environment, and the necessarily significant shielding. To better understand high-power target performance, we developed ex-situ experimental and computational frameworks to predict the behavior of subcooled flow boiling at IPF. Subcooled flow boiling experiments on Inconel 625 samples under IPF conditions (2 bar pressure, 10 GPM flow rate (i.e., 2249 kg/m 2 /s), 85 K subcooling) revealed that IPF's average operating power is at the early stage of boiling with a heat transfer coefficient of 48,000 W/m 2 /s. The proposed modeling framework enables us to predict a complete boiling curve, i.e., single-phase heat transfer, onset of nucleate boiling, two-phase heat transfer, and critical heat flux (CHF), with specification of input boiling parameters up to intermediate heat flux levels. The estimated CHF under IPF conditions is 5.2 MW/m 2 . Experimental data under reduced conditions (2 bar pressure, 1.5 GPM flow rate (i.e., 337 kg/m 2 /s), 45 K subcooling) served as validation cases for the computational modeling. This computational model can be further extended to more complicated systems replicating the real IPF configuration, for instance, to study void distribution as a function of the incident proton beam profile and coolant velocity profile of multiple cooling channels. Finally, the proposed experimental and computational frameworks provide a means to better understand cooling systems in the isotope production facilities at different accelerators, where in-beam monitoring of the cooling process is not available.

07 ISOTOPE AND RADIATION SOURCES↗