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At least 199 records · Page 11

Radiation Shielding Analysis of Niowave’s Uranium Target Assembly 2 (UTA-2) Facility for Molybdenum-99 Production

Niowave is seeking to develop a commercial medical isotope production process using a low-enriched uranium accelerator-driven subcritical reactor (ADSR) system to produce molybdenum-99. As part of the US Department of Energy national laboratory support to develop low enriched uranium molybdenum-99 production technologies, Oak Ridge National Laboratory is providing research and development resources to assist Niowave in the design and licensing process of its technology. The goals of this work under the radiation shielding task are to assist Niowave in developing a methodology and framework for efficient radiation shielding simulations of their Uranium Target Assembly 2 facility and to provide preliminary radiation shielding calculations and design recommendations for the facility.

07 ISOTOPE AND RADIATION SOURCES↗

Mean Time to Failure Analysis for the Los Alamos National Laboratory Weapons Engineering Tritium Facility Tritium Containment Vessels

The Weapons Engineering Tritium Facility (WETF) is a Hazard Category 2 nonreactor nuclear facility located in Technical Area 16 (TA-16) of Los Alamos National Laboratory. Tritium related research, development, and processing in support of stockpile stewardship program activities is conducted at WETF. Within WETF, various devices can be used to contain tritium and certain devices are designated as Tritium Containment Vessels (TCVs). Several different device types are identified as TCVs within WETF such as LP-50s and AL-M1s. Historically, a threshold tritium mass value has been used as the mechanism for designating a device as a TCV. When the tritium content of the device was below the tritium mass threshold it was not considered to be a TCV in the formal context. When the tritium content was above the administrative threshold, the device was considered to be a TCV. An upper tritium mass limit was also established for TCVs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Assessment and Improvement Proposals for High-Pressure Rinsing Processes during Cleanroom Preparation of Superconducting Radio-Frequency Cavities for Particle Accelerators

The goal of this internship is to develop a fluid dynamics characterization of the high-pressure flow used for cleaning the internal surface of the SRF cavities. Initially, the High Pressure Rinsing (HPR) process is introduced, focusing on the problems of the SRF cavities and how this method can be used to increase their acceleration efficiency and performance. The characterization of the problem is then presented, including an introduction to the multiphase flow modeling and the analysis of the Fermilab facility for the HPR. To accomplish the goal, a preliminary estimation of the variables was performed using MATLAB. Then, multiple 2D simulations were developed with Ansys Fluent, and the influence of the main parameter was investigated. The simulation results are then presented and analysed, with a focus on the possible future developments of the HPR process fluid dynamic characterization.

43 PARTICLE ACCELERATORS↗

Federal Water Management Planning Manual

FEMP developed this Federal Water Management Planning Manual to provide a resource for designing a comprehensive water management program. The manual describes strategies to optimize water use at federal facilities, use water balance analysis methods, identify design elements and procurement best practices, and expand use of alternative water.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

The Effects of Beta Irradiation on the Performance of ATR Fiber Probes

The use of real-time inline monitoring to track the chemical composition of slurries is recognized as a promising method to reduce the processing time of Hanford tank waste. The conversion of waste into glass suitable for long-term storage requires adherence to waste acceptance criteria and process control limits that respectively ensure safe operation of the vitrification facility and suitable glass quality. Real-time measurements of blended waste materials can ensure that the various criteria are met or can be addressed in a timely manner.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Updated Estimate of Tritium Permeation from TPBAR Disposal Containers in ILV (U)

A tritium source term analysis was performed for TPBAR disposal in the E-Area Intermediate Level Vaults for the E-Area Low-Level Waste Facility Performance Assessment (PA). This analysis is based on an earlier source term analysis which treated the bulk oft he tritium residual as tightly bound by the TPBAR getter material, with only a small fraction existing as tritiated moisture in the lithium aluminate ceramic pellets. Together with atmospheric moisture trapped in the free volume, the tritiated water vapor is assumed to corrode steel surfaces inside the disposal container, covering them with a magnetite film while generating hydrogen. The carbon steel walls of the disposal container are permeable to hydrogen, providing a path way for tritium to escape containment. The rate of hydrogen generation is assumed to be limited by the rate of corrosion, which is assumed to be governed by parabolic reaction kinetics obtained from the literature. This relies on the further assumption that the water vapor consumed by the corrosion reaction is continually replaced by moisture from the lithium aluminate pellets until all of that moisture is gone. In addition to tritium permeation, the analysis also includes a slow leak through the disposal container walls at the maximum allowable leak rate, 1 x 1CH standard cm3/s. Results were obtained for four different combinations of internal and container wall temperatures, established in an earlier thermal analysis that considered two different vault loadings and two different TPBAR activity levels. Instantaneous release rates for these four cases are plotted in Figure ES-1 below

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Compounding Risks from Natural Phenomena Hazards at U.S. Department of Energy Facilities

U.S. Department of Energy (DOE) facilities are exposed to adverse effects of natural phenomena hazards (NPHs). While the DOE Standard (STD) 1020-2016 provides criteria and guidance for assessing these effects, certain gaps exist, particularly related to combinations of NPHs that can compound the risk to DOE facility assets. This report describes a methodology to systematically consider compounding NPHs, identified gaps in DOE-STD-1020-2016, and recommends ways to update the standard. At any given DOE facility, NPH effects depend on the site-specific hydrometeorologic and geoseismic conditions. Therefore, compounding NPHs should be determined using site-specific assessments. DOE facility assets generally are assessed using site-specific NPHs with the assessment usually limited to the DOE site extents. However, an inventory analysis of a selected DOE facility also revealed that some assets may have off-site dependencies. These dependencies may be part of threat pathways that can lead to on-site asset failures because of off-site effects of the same NPHs that also affect the DOE site. These threat pathways usually are not accounted for in traditional NPH analyses. The report also describes the effects of duration of NPH effects. Duration of disruption, particularly for off-site dependencies can lead to on-site asset failures, disruption of evacuation roadways, and consequent human health and safety effects. Finally, the report recommends some considerations for updating DOE-STD-1020-2016 to include explicit consideration of compounding NPHs.

42 ENGINEERING↗

Defensive Cybersecurity Architecture Design Using Force-on-Force Cyber-Physical Modeling

Currently, nuclear power plant physical security systems are highly dependent on air-gaps as a protective measure against cyber-threats. Cyber-physical threats become more likely as advanced cyber-threat capabilities to jump air-gaps transition into common use. Defending against the emerging threat of cyber-enabled physical intrusions is poorly understood. The consequence of these cyber-physical attacks has no quantitative analysis method to inform risk-informed, performance-based cybersecurity approaches. By modifying the physical security simulation tool Dante, cyber-physical threat consequence was able to be analyzed on a notional facility. The results of this analysis are used to design a Defensive Cybersecurity Architecture (DCSA) for the physical security system to produce example resilience measures for this notional facility. A DCSA defines security levels to provide a graded approach for defending plant functions, and security zones for trusted communication between systems. This approach can be applied to real world systems to produce physical protection systems and response measures that are resilient to cyber-physical threats.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

Anchoring Strategies for Marine Renewable Energy at the Sandia Water Impact Facility

This report documents the formulation and preliminary analysis of potential anchoring strategies for use in future WEC testing at the Sandia Water Impact Facility given the existing 5-ton and 10-ton concrete blocks at the facility. Three anchoring strategies are created that are applicable to different mooring configurations for existing WECs of interest for testing. The three anchoring strategies have a maximum allowable load of between 24.57 kN and 281.17 kN, depending on whether the 5-ton or 10-ton blocks are used, with Strategy 3 with 10-ton blocks having the highest allowable load. The AquaHarmonics and CalWave xWave WECs can be tested up to a 1:1.5 and 1:1.15 scale, respectively, while maintaining a mooring configuration very close to the one used in ocean deployment. The requirements for the mooring lines, shackles, and triplates needed to support the three strategies are also calculated, with commercially available products specified. Finally, the padeyes connected to the existing concrete blocks are assessed and determined to be satisfactory for the loading expected with these anchoring strategies.

16 TIDAL AND WAVE POWER↗

Analysis of Solids Obtained from 201O, 201P, 202A, and 202B Contactors in the Salt Waste Processing Facility

In December 2022, Savannah River Mission Completion personnel recovered solids samples from four contractors (EXT-001, EXT-007, EXT-031, and EXT-040) employed at the Salt Waste Processing Facility (SWPF). These solids were submitted to Savannah River National Laboratory for analysis and characterization. This Technical Report outlines the analytical findings and observations associated with these samples and explores the potential source of solid generation within the SWPF. The following conclusions are offered as a result of this work: the light color solids recovered from EXT-007 are consistent with crystallized, soluble salt components (such as sodium nitrate, sodium hydroxide, and sodium carbonate), all of which can be easily re-dissolved in process water and do not represent a threat to SWPF CSSX processing. While the light solids from the top of EXT-040 were not analyzed, it is believed that they share similar characteristics to those recovered from EXT-007; The dark color solids recovered from EXT-001, EXT-031, and the inside of EXT-040 appear to be complex, consisting of several different types of solids: High concentrations of mercury (Hg, 40-75% by mass) partially present as elemental Hg are consistent with the precipitation of Hg 0 /Hg 2+ from disproportionation of Hg + species in pH swing conditions; Moderate concentrations of titanium (Ti) (2-7%, by mass) are also present, often co-located with Hg. This is consistent with co-precipitation, entrainment, and/or amalgamation of Ti and Hg; Moderate concentrations of iron (Fe) (1-3%, by mass) are observed in SWPF solids and are likely attributable to erosion of stainless-steel components within the CSSX process; Small concentrations of tungsten (W) and cobalt (Co) (<1%, by mass) are observed in SWPF solids. The presence of these solids is consistent with the erosion of components coated with W and Co (e.g., Stellite); Small concentrations of aluminum (Al) (<1%, by mass) are observed in the form of gibbsite and potentially sodium aluminosilicate. The presence of these compounds is likely attributable to the pH swing observed between the scrub and extraction cycles within the CSSX process. The following recommendations are made as a result of this work: Testing should be performed to determine the chemical drivers, process, and mechanism of Hg precipitation in CSSX processing. Special care should be taken to include titanium in tests to evaluate the potential for co-precipitation, entrainment, and amalgamation; Hg-mitigation options that would minimize or eliminate risks of Hg precipitation and solids accumulation (e.g., Hg-absorption techniques, ion exchange/adsorbents, flowsheet changes, etc.) should be assessed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization of Core from Sanford Underground Research Facility

Computed tomography and special core analysis data associated with the technical report series document Computed Tomography Scanning and Geophysical Measurements of the Enhanced Geothermal Systems (EGS) Collab SURF Core. Paronish, T.; Mackey, P.; Schmitt, R.; Crandall, D.; Moore, J.; Brown, S.; Roggenthen, W.; Schwering, P. C.; Dobson, P. F.; Kneafsey, T. Computed Tomography Scanning and Geophysical Measurements of the Enhanced Geothermal Systems (EGS) Collab SURF Core; DOE.NETL-2021.2866; NETL Technical Report Series; U.S. Department of Energy, National Energy Technology Laboratory: Morgantown, WV, 2022; p 76.

computed tomography↗

1016516285-AA - Handling Frame Assembly Table Structural and Seismic Analysis

Within the NIF (National Ignition Facility) sustainment project, LLNL is replacing all 1,728 blast shields with new ones. Before the new blast shield glass can be installed on NIF, it must pass through the AMOL (Amplifier Main Optics Loop). The AMOL process involves cleaning and coating the glass, assembling it into a Line Replaceable Unit (LRU), and performing testing and inspection. This work will be conducted in the OPFX cleanroom facility (Building 391, Room 1250) and requires several pieces of custom equipment. One key piece of equipment within the loop is the Handling Frame Assembly Table (see Figure 1.1). Its purpose is to assist operators in installing the blast shield glass into handling frames (1014442709) (see Figure 1.2). The table is also used to store handling frame assembly components. The primary components stored are the L-frames (1015208271), which are kept in a drawer underneath the table (see Figure 1.3). The purpose of this safety note is to ensure that the system does not pose a risk of injury to personnel during a seismic event. The system owner is OMST Engineering & Maintenance.

42 ENGINEERING↗

Accelerating Nuclear Facility Mission Completion Through an Improved Chemical Hazard Control Strategy - 20460

Code of Federal Regulation (CFR) 10 CFR 830, Nuclear Safety Management [1], explicitly identifies hazardous materials in addition to radioactive materials when establishing the scope of materials that must be evaluated in Hazard Category (HC) 1, 2, and 3 nuclear facilities. This inclusion has led to a near decade-long progression of more and more rigorous application of chemical hazard controls in these facilities. The progressive addition of more rigorous chemical hazard controls has occurred across the Department of Energy (DOE) Complex, having been driven by contractors and regulators alike. However, chemical hazards in other DOE facilities have been, and continue to be, controlled using traditional industrial safety centered strategies. Thus, identical chemical hazards are controlled using dramatically different approaches based solely on the facility type. Not only are the controls different, the analysis and derivation rigor and requirements are also significantly different. In particular, chemical control derivation and application in non-nuclear facilities are much less onerous than the comprehensive hazard analysis process used in nuclear facilities, even in circumstances in which chemicals are not initiators to, or do not exacerbate a nuclear event. While the derivation and control approaches used in these two types of facilities are disparate, the safety results are similarly effective based on a review of reported chemical events from 2014 through 2018. While both approaches achieve comparable safety outcomes, the approach used in nuclear facilities results in significantly greater preparation costs and implementation costs. Furthermore, the problem with the differing approach is not relegated to the control preparation and implementation burden. The operational impact of potentially unnecessary controls slows mission completion without a corresponding increase in safety, not to mention the dilution of the importance of Technical Safety Requirements. The proposed solution must first include a critical and thorough examination of DOE Complex and top-tier chemical industry chemical hazard control practices, coupled with a similar examination of the driving facility regulatory requirements. Results from the examination can then be used to fuel a bold departure from current chemical control practices in Hazard Category 1, 2, and 3 nuclear facilities. Development of an improved and consistent chemical hazard control process would follow, one that continues to ensure safety while accelerating mission completion and significantly reducing the burden associated with chemical hazard control development and implementation. The improved process must establish a hierarchy where radiological hazards are addressed using safety class and safety significant controls and chemical hazards are addressed using overarching safety management programs (with limited unique exceptions). The improved Hazard Category 1,2, and 3 nuclear facility chemical hazard control strategy, while compliant with 10 CFR 830 [1], would be founded on the requirements in 10 CFR 851, Worker Safety and Health Program [2], and the complementary DOE O 151.1D, Comprehensive Emergency Management System [3]. In this way, the Hazard Category 1, 2, and 3 nuclear facility chemical hazard control strategy would, wherever appropriate, be the same process used at other DOE Complex facilities. Ultimately, the improved process will maintain safety and significantly reduce costs and life cycle risk through faster mission completion. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A Monte-Carlo Method for Evaluating the Economic Performance of Plastics Recycling Systems Using Historical Pricing

The management of plastic waste is a considerable emerging global concern. Conventional plastics recycling rates range from 5–10% in the US, which results in considerable material being sent to landfill. New sorting technologies using artificial intelligence have the potential to dramatically increase the ability of waste management companies to sort out valuable fractions of plastic waste and create high-purity streams for secondary markets. However, several fractions of the plastics recycling stream, i.e., low-density polyethylene, polypropylene, and polystyrene, do not have well-developed markets. These materials may be sent to pyrolysis facilities for thermal recycling, but the economics of the pyrolysis industry are currently uncertain. This work aims to identify the breakeven price of the plastic fractions that would be sent to pyrolysis in order to determine the economic viability of the sorting facility. The work will use Monte Carlo analysis to evaluate several scenarios and understand how compositional variation, price variability, and facility attributes affect the breakeven price. Here, the presorting conditions in Materials Recovery Facilities strongly affects the profitability of the sorting facility. High residual high density polyethylene (HDPE) and polyethylene terephthalate (PET) content in presorted bales are important revenue drivers for the facility, even in presorted 3-7 bales. Key variables that lead to profitability include a high sale price of polyethylene terephthalate, a high sale price of high-density polyethylene, and if polypropylene can be sold to recycling markets (as opposed to pyrolysis facilities).

Plastics↗

Thermal loading analysis of the ring injection dump for the Spallation Neutron Source facility

The ring injection dump (RID) is the largest beam dump in the Spallation Neutron Source facility at Oak Ridge National Laboratory and accepts a fraction of the beam from the ring that is not captured in the ring during injection. Thermo-fluid modeling of the RID components was performed during the initial design by using the software ANSYS CFX to characterize the power rating for the RID. From the calculations, the power rating was lowered from 200 kW to 150 kW, at 1.3 GeV beam energy due to concerns regarding the heating of the concrete structure. The Proton Power Upgrade project scope included reevaluating the power limit to the RID in anticipation that increasing it would provide more operational flexibility. The focus was on the shielding because it was the limiting factor, and data collected from thermocouples installed on the structure were used to benchmark the analysis. A three-stage process was adopted in the latest study. First, the already existing steady-state calculations were validated and extended with the latest Monte Carlo N-Particle source term calculations. Second, transient calculations were conducted to capture the dynamic state of the system in response to the energy dumped to the RID during the following operational period of the beam from 2005–2019 (14 years), and sensitivity analyses of crucial parameters were performed to benchmark the model with reasonable accuracy. Finally, an idealized transient cycle analysis was conducted with realistic duty factors to predict the temperature distribution at higher beam powers. Overall, the results confirmed the accuracy of the original steady state model but allowed the development of a new validated transient model with higher accuracy for future analyses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Alpha-heating analysis of burning plasma and ignition experiments on the National Ignition Facility

A recent experiment conducted on the National Ignition Facility (NIF) described in the study by Abu-Shawareb et al. achieved a fusion yield output of 1.3 MJ from ~ 220 kJ of x-ray energy absorbed by the capsule, demonstrating remarkable progress in the field of laser driven inertial confinement fusion. In the study by A. R. Christopherson [“Effects of charged particle heating on the hydrodynamics of inertially confined plasmas,” Ph.D. thesis (2020)], the plasma conditions needed to claim the onset of ignition and burn propagation were outlined and multiple criterion were provided to assess progress in inertial fusion experiments. In this work, we modify the metrics from A. R. Christopherson to accurately calculate performance metrics for indirect-drive experiments on the NIF. We also show that performance metric trends observed in NIF data are consistent with theory and simulations. This analysis indicates that all the identified criterion for ignition and burn propagation have been exceeded by experiment 210 808.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Oak Ridge Reservation Meteorological Overview Report, Oak Ridge, TN

This report is purposed to provide a summary of the local climate and data from meteorological (Met) towers and wind profilers at the East Tennessee Technology Park (ETTP), the Y-12 National Security Complex Site (Y-12), the Oak Ridge National Laboratory (ORNL), as well as other federal, state, or local agencies as necessary to support the natural phenomena hazards (NPH) analysis 10-year update required by Department of Energy (DOE)- STD-1020-2016, Natural Phenomena Hazards Analysis and Design Criteria for DOE Facilities. The data from this report will be used to support high wind analyses, precipitation analysis, flood analysis, lightning analysis, and other severe weather phenomena for the ETTP, the ORNL, and Y-12.

54 ENVIRONMENTAL SCIENCES↗

Thermohydraulic Design Analysis of the Target Assembly in the Material Plasma Exposure Experiment Facility

The Material Plasma Exposure eXperiment (MPEX) project seeks to design a steady-state linear plasma facility at Oak Ridge National Laboratory that will be used to study plasma-material interactions (PMIs) at fusion prototypic levels, supporting the evaluation and development of materials for the next generation of fusion devices. This study is focused on PMI exposure of small-size neutron-irradiated specimens, which are clamped onto an actively cooled component. A thermohydraulic evaluation of a new MPEX target assembly design to assess the appropriate operation during MPEX operation is presented. To further guide the design and assess the structural integrity of the components under expected loads, preliminary thermomechanical stress analyses were also conducted. To ensure good thermal contact between the components, thermal interface materials, such as silver flexible graphite, were used in the assembly. It was found that the maximum target temperatures of 1572, 1463, and 1315 K were obtained for Grafoil thicknesses of 0.61, 0.38, and 0.25 mm, respectively. The distribution of the axial deformation at high heat fluxes showed that there are no gaps between components, indicating good contact at material interfaces. Moreover, the contact pressure between the target and other components indicated that very good contact was established at these interfaces. Furthermore, the stress-strain conditions for the target will be further used to assess the appropriate operation during MPEX experiments and gain insight into materials science phenomena during PMI experiments.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗