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At least 91 records · Page 5

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DOE OSTI2021

Nimble Feedthrough Qualification - 125% High Explosive Overpressure Test Plan RevA

The primary purpose of this High Explosive (HE) Over-Pressure Test (OPT) is to qualify top cover diagnostic feedthroughs that will be used on LLNL Nimble Subcritical Experiment (SCE) Series designs per experimental design verification requirements specified in ASME Boiler and Pressure Vessel Code Case 2564, Impulsively-Loaded Pressure Vessels, Section VIII, Division 3; and to satisfy the over-test requirement of DOE-STD-1212. The diagnostic feedthroughs are part of the Vessel Confinement System (VCS), which is credited as a Safety Significant Design Feature per the U1a Facility Documented Safety Analysis (DSA). The OPT will be conducted at the LANL Area 1, R306 Firing Site (TA-15-R306) in a 3-foot diameter VCS depicted in Figure 1.

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers” for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field’s Metal eutectic alloy with a melting point of 62 °C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Temperature Effect of Cesium Exchange onto Crystalline Silicotitanate in AP-107 and AP-105 Hanford Tank Wastes and Two Simulants

Washington River Protection Solutions, LLC (WRPS) is charged with the development of the Tank Side Cesium Removal (TSCR) system to process Hanford tank waste supernates in preparation for vitrification. In addition to a filtration step, TSCR will remove cesium (Cs) using ion exchange columns filled with crystalline silicotitanate (CST) ion exchange media. CST is produced by Honeywell UOP, LLC. The documented safety analysis (DSA) developed for the TSCR system limits a single column loading to 141,600 Ci 137 Cs. Given a 137 Cs isotopic mass fraction of 20% and the planned CST bed size of 596 L (157.5 gal) in a TSCR column, this equates to 0.10 mmole Cs per g CST (Cs distribution coefficient, K d , 1400 mL/g). Factors that influence Cs uptake by CST include (but are not limited to) (1) CST production (lot-to-lot variations), (2) contact temperature, (3) contact duration, (4) competitors in the tank waste feed, (5) anionic composition of the tank waste feed, and (6) the 137 Cs isotopic mass fraction (differs slightly among tank wastes and decreases with time).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Environmental Protection Agency Drinking Water Protective Action Guides Implementation Recommendations

This report documents the analysis conducted by Sandia National Laboratories on the effect of various alterations to published methodologies to calculate Derived Response Levels for Environmental Protection Agency Drinking Water Protective Action Guides. Specifically, this study sought to assess and provide recommendations on calculation of the Derived Response Level accounting for decay during the consumption period, assess the impact of decay on laboratory Minimal Detectable Concentration in water samples as compared to the Derived Response Level, make a recommendation on the calculation of Derived Response Level consistent with existing Public Protection Methods, and make a recommendation on the use of six age groups versus eight age groups based on available dose coefficients for calculation of the Derived Response Level. The authors analyzed these various factors using nominal radionuclide mixes from four scenarios and compared calculation of the Derived Response Level accounting for decay and no decay and then compared those results to the laboratory Minimal Detectable Concentrations. The authors concluded that decay should be included in the calculation of the Derived Response Level, existing Public Protection Methods should be employed to calculate the Derived Response Level, six age groups should be used versus eight, and the use of both decay and Public Protection Methods result in little to no concern for water samples meeting the Minimum Detectable Concentration requirements. The results of this study may be used in further developing and implementing a method for the Environmental Protection Agency Water Derived Response Level calculation in the Federal Radiological Monitoring and Assessment Center Assessment Manual.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2021

As-Run Physics Analysis for the EPRI Zirconium Growth C Capsule

This engineering calculation and analysis report documents the as-run nuclear-heating term, source-term, and radiation damage (measured in displacements per atom [dpa]) accumulation for the Electric Power Research Institute zirconium growth experiment's C capsule (EPRI-ZG-C). These data have been calculated using MCNP 6.1 and SCALE 6.2.3. These data can be used for: the shipping of the EPRI-ZG-C capsule, the acceptance of the capsule at the Materials and Fuels Complex, and programmatic scientific needs.

36 MATERIALS SCIENCE↗

DOE OSTI2022

3013 Surveillance Program Interim Summary Report (FY16-FY21)

The K-Area Documented Safety Analysis (DSA) requires the facility participate in the SRS Site Surveillance Program as a part of Shipping Package Qualification and Storage Surveillance Program. The Site Surveillance Program outlines activities for field surveillance and laboratory tests that demonstrate the 3013 containers meet the functional performance requirements described in the DSA. The SRS Surveillance Program also supports the complex-wide Integrated Surveillance Program (ISP) for 3013 containers in accordance with DOESTD-3013. The purpose of this report is to provide a summary of the SRS portion of the surveillance program activities from FY16 through FY21 and formally communicate the interpretation of these results by the Surveillance Program Authority (SPA).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

H-Canyon Jumper Gasket Sealing Evaluation

H-Canyon Documented Safety Analysis approved December 2019 credits the leak integrity of HCanyon (HCA) process tanks, piping, and jumpers in preventing the release of radioactive material in a post seismic environment. Demonstration of leak tightness capability for this equipment was achieved through substantial analytical effort. However, one potential weak point that affects leak integrity exists at the gasketed joint between Hanford Connectors (HC) and nozzles. Review of the HCA Crane Log over a three-year period indicates the HCs periodically leak. Savannah River National Laboratory (SRNL) was requested to perform gasket compression tests that evaluate gasket relaxation characteristics, i.e., creep, and provide improvements to the HC installation process that can reduce or eliminate the frequency of leaks at HCs. A HC torque value of 312 ft-lbf is the target value used because it is the minimum torque value, with 95% confidence, when an unlubricated HC is impacted for standard sealing period of 9 seconds.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

Packaging Capacity Calculation: Pu Oxide Packaging Options with the 2 Quart SAVY Container

This calculation supports programmatic efforts to Dilute and Dispose Pu. Metal is to be oxidized at LANL, then transported to SRS for dilution at SRS. Ultimately, the diluted oxide will be disposed of at the Waste Isolation Pilot Plant (WIPP). Currently 3013 containers are used to package the oxide, which are placed in to 9975 Type B shipping containers. The Oxide Packaging Technology Maturation Plan determined that a transition to the used of 2-qt SAVY containers placed in 9977 would increase efficiency in shipping and handling and increase programmatic flexibility. The 2-quart SAVY container meets DOE M 441.1 and TA-55 Documented Safety Analysis requirements for handling and storage of Pu oxide at LANL.

36 MATERIALS SCIENCE↗

DOE OSTI2023

Analytics Supporting Stockpile and Enterprise Planning [Slides]

A robust and growing modeling effort supports Los Alamos weapons leadership as well as NNSA (NA18, NA12, NA19) and DoD (DASD-NM, USSTRATCOM). Current work are valuable to several key NNSA working groups supporting strategic planning: Stockpile Stewardship Management Plan (SSMP); Requirements & Planning Document (RPD) analysis; Requirements and Capacity Working Group (RCWG); Production Integration Collaboration Working Group (PICWG); Enterprise Modeling and Analysis Consortium (EMAC); Cost Estimating Analysis Group (CEAG).

45 MILITARY TECHNOLOGY, WEAPONRY, AND NATIONAL DEF↗

DOE OSTI2020

Annual Transuranic Waste Inventory Report – 2020 (Data Cutoff Date 12/31/2019)

The purpose of this Annual Transuranic Waste Inventory Report (ATWIR) – 2020 is to document the inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2019. This report also notes major changes to the inventory used for the ATWIR-2019 that had a data cutoff date of December 31, 2018. This updated inventory information is available to the U.S. Department of Energy (DOE) TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports the DOE Carlsbad Field Office (CBFO) input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act analyses), planned changes, and other design changes as needed for the WIPP project.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2021

Annual Transuranic Waste Inventory Report – 2021 (Data Cutoff Date 12/31/2020)

The purpose of this Annual Transuranic Waste Inventory Report (ATWIR) – 2021 is to document the inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2020. This report also notes major changes to the inventory used for the ATWIR-2020 that had a data cutoff date of December 31, 2019. This updated inventory information is available to the U.S. Department of Energy (DOE) TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports the DOE Carlsbad Field Office (CBFO) input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act analyses), planned changes, and other design changes as needed for the WIPP project.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

Annual Transuranic Waste Inventory Report – 2022 (Data Cutoff Date 12/31/2021)

The purpose of this Annual Transuranic Waste Inventory Report (ATWIR) – 2022 is to document the inventory estimate of transuranic (TRU) waste reported by the TRU waste generator sites as of December 31, 2021. This report also notes major changes to the inventory since the ATWIR-2021, which had a data cutoff date of December 31, 2020. This updated inventory information is available to the U.S. Department of Energy (DOE) TRU waste complex, Waste Isolation Pilot Plant (WIPP) stakeholders, and regulators. The TRU waste inventory information is used for strategic planning, and supports the DOE Carlsbad Field Office (CBFO) input into documents (e.g., WIPP Documented Safety Analysis and National Environmental Policy Act evaluations), performance assessments, planned changes, and other design changes as needed for the WIPP project.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Exploring Grid-Interactive Efficient Building Strategies for Laboratories Through Energy Modeling

Laboratories are often overlooked in demand flexibility research due to constraints on their operations as mission critical facilities, despite the major role they play in an organization's emissions. Laboratories consume 3-4 times more energy than a typical office building and are commonly the largest energy users on any campus. Consequently, most laboratories in the United States are significant contributors to their organization's carbon footprint if their energy needs are met through the combustion of fossil fuels. As part of the initiative to decarbonize laboratories, this report documents an analysis on specifically grid-interactive efficient building (GEB) opportunities for reducing energy costs and emissions associated with laboratory operations. The goal of this initiative was to provide a case study and guidance on how to use OpenStudio and REopt as modeling tools for GEB technologies and strategies in laboratory environments across different climate zones in the United States. The analysis found that efficiency-based GEB strategies had the most significant impact on laboratory operations, while load-shedding and load-shifting GEB strategies produced smaller results. The culmination of these approaches applied across all five climate zones generated on average: 1) 28% energy cost savings and 30% greenhouse gas (GHG) emissions reductions, and 2) 4% enhanced energy cost savings under a time-of-use (TOU) pricing schedule compared to traditional pricing schemes. Grid-interactive efficiency building measures were found to produce the greatest energy savings in both electricity and natural gas, particularly in regions with high electrical loads, such as warm climates for cooling. Laboratories that had high levels of natural gas consumption, meanwhile, experienced the greatest emission reductions. The report concludes with an analysis on the opportunities for flexible loads in lab spaces and how small-scale measures in addition to opaque pricing structures for peak demand could become barriers to demand flexibility planning. The report also explores how electrifying laboratory buildings with heat pumps could reduce energy costs and GHG emissions.

24 POWER TRANSMISSION AND DISTRIBUTION↗

DOE OSTI2024

Evaluation of Second Target Station Target Streaming for the Preliminary Design of the Target Assembly

This report documents the analysis of dose rate fields from a combination of two sources stemming from the target system of the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source (SNS) Second Target Station (STS). This analysis supports the target drive components’ lifetime estimates and conventional facilities’ preliminary design. The two sources detailed in this report are the radiation generated by the proton beam incident on the target and the activated cooling water in the pipes at the top of the target assembly. The analysis considers the STS target system in operational configuration, with all shielding in place. The dose rate fields are calculated in the target drive room and through the target drive room ceiling. This report calculates the thickness necessary for the target drive room’s high-density concrete ceiling to meet the 0.25 mrem/hr dose limit requirement in the high bay directly above the target drive room. In addition to sizing the target drive room ceiling, this report includes details of the dose field surrounding the target drive assembly to estimate the deposited dose in the target drive assembly components resulting from both streaming and decay of activated water. Prompt neutrons and gammas generated by the proton beam incident on the target primarily stream along the gaps between the target assembly and core vessel shielding. Because the models and analysis detailed in this report are based on the STS preliminary design, future analysis is needed to support the STS final design.

43 PARTICLE ACCELERATORS↗

DOE OSTI2025

HFIR LEU High Density Silicide Dispersion Optimized Design Steady-State Heat Transfer Analyses

Steady-state heat transfer simulations of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) with the low-enriched uranium (LEU) high-density silicide dispersion Optimized fuel design were performed to support comprehensive performance and safety metric studies concerning this design. The LEU Optimized design operates at 95 MW to maintain HFIR’s current highly enriched uranium (HEU) core performance level at 85 MW. Full cycle Mode 1 full flow Case 1 (inlet temperature), Case 2 (flux-to-flow), and Case 3 (inlet pressure) safety limit analyses were performed to assess the margins to critical heat flux. Under the prescribed conditions, this LEU design meets the safety limit and limiting control setting requirements outlined in HFIR’s documented safety analysis; however, the safety margins are less than those for the 85 MW HEU core, and several assumptions were made where fuel fabrication and qualification data are currently lacking for the silicide fuel design. Effects of changes to pertinent fuel fabrication assumptions and uncertainty factors on thermal safety margins were also evaluated, showing that the margins are sensitive to many of these parameters. Power and pressure perturbations were also performed, indicating that significant steady-state thermal margins could be gained by increasing the coolant inlet pressure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Sierra/SD - Users Manual - 5.28

Sierra/SD provides a massively parallel implementation of structural dynamics finite element analysis, required for high-fidelity, validated models used in modal, vibration, static and shock analysis. This document provides a user's guide to the input for Sierra/SD. Details of input specifications for the different solution types, output options, element types and parameters are included.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2026

Sierra/SD - Users Manual - 5.30

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2026

Sierra/SD - Users Manual - 5.32

Beale, Dagny Marie [Sandia National Laboratories (↗