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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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DOE OSTI2017

M4SF-17LL010301071: Thermodynamic Database Development

This progress report (Level 4 Milestone Number M4SF-17LL010301071) summarizes research conducted at Lawrence Livermore National Laboratory (LLNL) within the Argillite Disposal R&D Work Package Number M4SF-17LL01030107. The DR Argillite Disposal R&D control account is focused on the evaluation of important processes in the analysis of disposal design concepts and related materials for nuclear fuel disposal in clay-bearing repository media. The objectives of this work package are to develop model tools for evaluating impacts of THMC process on long-term disposal of spent fuel in argillite rocks, and to establish the scientific basis for high thermal limits. This work is contributing to the GDSA model activities to identify gaps, develop process models, provide parameter feeds and support requirements providing the capability for a robust repository performance assessment model by 2020.

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DOE OSTI2018

US Sections Prepared for Future NEA Crystalline Club (CRC) Report on Status of R&D in CRC Countries Investigating Deep Geologic Disposal in Crystalline Rock

U.S. knowledge in deep geologic disposal in crystalline rock is advanced and growing. U.S. status and recent advances related to crystalline rock are discussed throughout this report. Brief discussions of the history of U.S. disposal R&D and the accumulating U.S. waste inventory are presented in Sections 3.x.2 and 3.x.3. The U.S. repository concept for crystalline rock is presented in Section 3.x.4. In Chapters 4 and 5, relevant U.S. research related to site characterization and repository safety functions are discussed. U.S. capabilities for modelling fractured crystalline rock and performing probabilistic total system performance assessments are presented in Chapter 6.

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DOE OSTI2018

GeoVision Analysis: Exploration Task Force Report

The GeoVision study effort included ground-breaking, detailed research on current and future market conditions and geothermal technologies in order to forecast and quantify the electric and non-electric deployment potentials under a range of scenarios, in addition to their impacts on the Nation’s jobs, economy and environment. Coordinated by the U.S. Department of Energy’s (DOE’s) Geothermal Technologies Office (GTO), the GeoVision study development relied on the collection, modeling, and analysis of robust datasets through seven national laboratory partners, which were organized into eight technical Task Force groups. The purpose of this report is to provide a central repository for the research conducted by the Exploration Task Force. The Exploration Task Force consists of four individuals representing three national laboratories: Patrick Dobson (task lead) and Christine Doughty of Lawrence Berkeley National Laboratory, Anna Wall of National Renewable Energy Laboratory, Travis McLing of Idaho National Laboratory, and Chester Weiss of Sandia National Laboratories. As part of the GeoVision analysis, our team conducted extensive scientific and financial analyses on a number of topics related to current and future geothermal exploration methods. The GeoVision Exploration Task Force complements the drilling and resource technology investigations conducted as part of the Reservoir Maintenance and Development Task Force. The Exploration Task Force however has focused primarily on early stage R&D technologies in exploration and confirmation drilling, along with an evaluation of geothermal financing challenges and assumptions, and innovative “blue-sky” technologies. This research was used to develop geothermal resource supply curves (through the use of GETEM) for use in the ReEDS capacity expansion modeling that determines geothermal technology deployment potential. It also catalogues and explores the large array of early-stage R&D technologies with the potential to dramatically reduce exploration and geothermal development costs, forming the basis of the GeoVision Technology Improvement (TI) scenario. These modeling topics are covered in detail in Potential to Penetration task force report. Most of the research contained herein has been published in peer-reviewed papers or conference proceedings and are cited and referenced accordingly. The sections that follow provide a central repository for all of the research findings of the Exploration and Confirmation Task Force. In summary, it provides a comprehensive discussion of Engineered Geothermal Systems (EGS) and associated technology challenges, the risks and costs of conducting geothermal exploration, a review of existing government efforts to date in advancing early-stage R&D in both exploration and EGS technologies, as well as a discussion of promising and innovative technologies and implementation of blue-sky concepts that could significantly reduce costs, lower risks, and shorten the time needed to explore and develop geothermal resources of all types.

15 GEOTHERMAL ENERGY↗
DOE OSTI2018

Technical and Regulatory Considerations for Use of Fillers in DPCs

There are currently 2,462 dual-purpose canisters (DPCs) containing spent nuclear fuel (SNF) across the United States. Repackaging DPCs into specialized disposal canisters could be financially and operationally costly with additional radiological, operational safety, and management risks. There are several approaches to facilitate direct disposal of DPCs and demonstrate acceptable repository performance. A promising approach is to fill the void space within the DPCs with a material that would significantly limit the potential for criticality through limiting moderation and/or the addition of neutron absorbers in the interstitial spaces within the fuel assemblies and baskets. An acceptable filler would demonstrably show that the probability of criticality in DPCs during the disposal period of interest to be below the probability threshold for inclusion in repository performance assessment. Based on previous work conducted by domestic and international organizations, two approaches were identified as potentially viable for introduction of fillers into DPCs as liquids that would eventually solidify: (1) molten metal fillers introduced at higher temperatures, and (2) resins or cement slurries that solidify at lower temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Developing Design Criteria for Safeguards Seals for Spent Fuel Transportation Casks

Current designs for spent fuel transportation casks cannot ensure a cask's integrity during shipment, nor is there any verifiable means of maintaining continuity of knowledge (CoK) on a cask's contents. Spent fuel destined for encapsulation plants or geological repositories requires additional containment and surveillance (C/S) measures during shipment. Following final safeguards accountancy measurements on spent fuel assemblies, the shipment of verified assemblies will require unprecedented reliance on maintaining CoK on the fuel inside transport casks. Such increased reliance is due to the lack of reverification of spent fuel following encapsulation into disposal canisters and by meeting the requirement of dual C/S measures during such fuel shipments according to recommendations made by the Application of Safeguards to Geological Repositories (ASTOR) International Atomic Energy Agency (IAEA) expert group. By designing spent fuel transportation casks with effective seals integrated into their design, CoK can be more effectively maintained than by ad hoc C/S measures because seal integration ensures that a cask has not been tampered with. Externally applied seals might not be able to provide such assurance for currently designed spent fuel transportation casks, although some combination of seals, detectors, and/or a technology that can verify canister integrity might provide this assurance. This paper examines the design criteria for integrating safeguards seals into transportation casks and provides recommendations for near-term applications.

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DOE OSTI2018

Innovative Plutonium Waste Forms: Acceptability to Shipp to WIPP

The mission of the Waste Isolation Pilot Plant in New Mexico is the safe and permanent disposition of transuranic radioactive waste from atomic energy defense activities of the United States. The plant consists of an underground repository in a salt formation that is free of flowing water, easily mined, impermeable, and geologically stable. The salt rock naturally seals fractures and closes openings over time. The repository is now anticipated to receive a greater amount of plutonium materials than originally envisioned. An engineered waste form for safe and efficient transport and long-term disposal of plutonium materials is an urgent need. An optimal waste form could also reduce the disposition life cycle and its associated costs to the taxpayer.

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DOE OSTI2018

Evaluation of Engineered Barrier Systems in the Disposition of Spent Nuclear Fuel

This document is a summary of the R&D activities associated with the Engineered Barrier Systems Work Package. Multiple facets of Engineered Barrier Systems (EBS) research were examined in the course of FY18 activities. This report is focused on delivering an update on the status and progress of modelling tools and experimental methods, both of which are essential to understanding and predicting long-term repository performance as part of the safety case. Specifically, the work described herein aims to improve understanding of EBS component evolution and interactions. Utlimately, the EBS Work Package is working towards producing process models for distinct processes that can either be incorporated into performance assessment (PA), or provide critical information for implementing better constraints on barrier performance The main objective of this work is that the models being developed and refined will either be implemented directly into the Generic Disposal System Analysis platform (GDSA), or can otherwise be indirectly linked to the performance assessment by providing improved bounding conditions. In either the case, the expectation is that validated modelling tools will be developed that provide critical input to the safety case. This report covers a range of topics — modelling topics include: thermal-hydrologic-mechanical-chemical coupling (THMC) in buffer materials, comparisons of modelling approaches to optimize computational efficiency, thermal analysis for EBS/repository design, benchmarking of thermal analysis tools, and a preliminary study of buffer re-saturation processes. Experimental work reported, includes: chemical evolution and sorption behavior of clay-based buffer materials and high-pressure, high temperature studies of EBS material interactions. The work leverages international collaborations to ensure that the DOE program is active and abreast of the latest advances in nuclear waste disposal. This includes participation in the HotBENT Field Test, aimed at understanding near-field effects on EBS materials at temperatures above 100 °C, and the analysis of data and characterization of samples from the FEBEX Field Test. Both the FEBEX and HotBENT Field Tests utilize/utilized the Grimsel Test Site in Switzerland, which is situated in a granite host rock. These tests offer the opportunity to understand near field evolution of bentonite buffer at in situ conditions for either a relatively long timescale (18 years for FEBEX) or temperature above 100 °C (HotBENT). Overall, this report provides in depth descriptions of tools and capabilities to investigate nearfield performance of EBS materials (esp. bentonite buffer), as well as tools for drift-scale thermal and thermal-hydrologic analysis critical to EBS and repository design. For a more detailed description of work contained herein, please see Section 10 ("Conclusions") of this document.

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DOE OSTI2019

Phase Transformation Kinetics and Alloy Microsegregation in High Pressure Die Cast Magnesium Alloys (Final Report)

The primary objective of this project is to conduct a systematic, quantitative study of phase transformation kinetics and microsegregation in HPDC Mg alloys in both as-cast and heat treated conditions. These results will be used to develop and validate physics-based phase transformation kinetics micro-models that quantitatively capture microstructural evolution and microsegregation during HPDC and heat treatment of Mg alloys. Finally, this knowledge will be transferred to industry and the wider research community through these micro-models and the experimental data that will be stored in the National Institute of Standards and Technology (NIST)-DOE Energy Efficiency and Renewable Energy Advanced Automotive Cast Magnesium Alloys Repository, as well as a new repository, the University of Michigan Department of Energy Predictive Integrated Structural Materials Science (UM DOE PRISMS) Materials Commons. To accomplish these goals, a systematic and comprehensive experimental study is being combined with computational modeling and simulation.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Fracture Toughness Evaluation for Spent Nuclear Fuel Clad Systems Using Spiral Notch Torsion Fracture Toughness Test

Many radioactive materials within the nuclear fuel cycle present a significant hazard. Such materials include spent nuclear fuel, high-level waste, legacy waste and other nuclear materials. These materials are often held in long-term storage as an interim stage within their lifecycle, including reuse or disposal. Safety and security of spent nuclear fuel (SNF) interim storage installations and follow-on SNF transportations to its final repository sites are very important, due to a great concentration of fission products, actinides and activation products. Fracture mechanics approach in applying to SNF system reliability investigation, especially for the high burn up (HBU) SNF, during SNF long term dry storage or SNF transportation is necessary due to the inherited flaws and inhomogeneity structures existed in a SNF system from the reactor operation. For instance, such as hydride and oxide formation, surface flaw induced by SNF assembly contact interactions, or internal flaws induced by pellet-clad mechanical interaction (PCMI). However, none of the existing fracture toughness data deal with fuel cladding specific geometry or SNF material conditions, such as cladding structure with the segment fuel pellets configuration and the PCMI mechanism. Consequently, the application of existing fracture toughness data to SNF system reliability investigation can be error prone; thus, the development of an industry-wide consensus fracture mechanics approach is needed. The objective of this research is to develop an in-situ fracture mechanics testing protocol, including the associated analytical procedure, that is suitable for evaluating the SNF fracture toughness. Fracture toughness data was obtained under quasi-static fracture loading using Oak Ridge National Laboratory (ORNL) developed Spiral Notch Torsion Test (SNTT) technology carried out on a biaxial tension/torsion tester. These data will be used to support SNF reliability investigation during long-tern SNF dry storage stage or in the follow-on SNF system transport to the final repository sites. SNTT has been a recent breakthrough in measuring fracture toughness for different materials, including metals, ceramics, concrete and polymers composites. Due to its high geometry constraint and unique loading condition, SNTT can be used to measure the fracture toughness with smaller specimens without concern of size effects. The application of SNTT to brittle materials has been proved to be successful. The micro-cracks induced by original notches in brittle materials could ensure crack growth in SNTT samples. Therefore, no fatigue pre cracks are needed for brittle materials specimens. The application of SNTT to the ductile material to generate valid toughness data will require a test sample with sufficient pre-crack length to increase the sample localized constraint at crack front. Fatigue pre-crack growth techniques with the associated compliance function estimated was developed for estimating the crack penetration depth to monitor the fatigue crack growth evolution. In order to extend SNTT approach to a thin shell cladding structure material, a new testing protocol and the associated analytical procedure for estimating the SNF fracture toughness was developed. The detailed SNTT approach and its estimated fracture toughness for the Zr-4 cladding structure with segment alumina inserts are presented in this report. The SNTT test results indicate that SNTT method is a reliable test approach with good repeatability in applying to Zr-4 cladding material. For the medium and the short crack length, the estimate J Q upon fracture for the baseline Zr-4 cladding is at 285.7 lb/in with 2-sigma uncertainty of 18.59 lb/in., and the associated K Q is at 61.4 Ksi√in. For a long crack length, the crack initialization orientation is apparently deviated from the principle tensile stress contour – 45° spiral crack front (i.e., Mode I + Mode III failure mechanism); and the estimated J Q is at 108 lb/in., the associated K Q is at 37.7 Ksi√in. The estimate Mode I J Q (along the 45° pitch principle tensile stress contour) is at 200 lb/in.; which indicate a significant reduction in the fracture toughness in a tubing structure under mixed mode loading condition, Mode I + Mode III, compared to that of Mode I (tensile stress dominated) alone.

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DOE OSTI2017

Effect of Iodide on Radiolytic Hydrogen Peroxide Generation

This report fulfills the milestone (M4SF-17PN010501041 Report on Radiolysis Modeling for the Defense Repository) to discuss continued integration of the PNNL Radiolysis Model and the ANL Mixed Potential Model. This work concerns the development of an integrated Radiolysis Model (RM) for evaluating defense waste materials (oxide and metal) degradation and radionuclide mobilization. Within an anoxic repository environment, primary oxidizing species (e.g., hydrogen peroxide (H 2 O 2 ), OH• radicals, as well as chlorate and other oxidizing species depending on the disposal environment) will be generated at the surface of the nuclear waste forms as a function of their specific activity. RM development has included expansion of chemical environments considered to encompass species for various disposal environments. PNNL has been coordinating this effort with ANL on the integration of the radiolysis work with the fuel degradation model. In this study, we demonstrate and approximate possible effects of iodide on H 2 O 2 generation. As has been shown these are conditions for which H 2 O 2 generation is reduced. We find that the presence of the iodide ion reduces the steady-state H 2 O 2 concentration, but not to the same degree as bromide at micro-molar concentrations.

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DOE OSTI2019

Neptunium(IV) Diffusion through Bentonite Clay

DOE Fellow, Silvina A. Di Pietro, completed a 10-week internship at Lawrence Livermore National Laboratory (LLNL) in Livermore, California. The objective of the study was to determine the diffusion profile of neptunium (Np) in bentonite clay (potential geo-engineered barrier for nuclear waste repositories). Under the mentorship of Drs. Claudia Joseph (Karlsruhe Institute of Technology, Germany) and Mavrik Zavarin, she prepared a series of core abrasion peeling samples. This procedure was a continuation of a Np diffusion experiment conducted by Dr. Joseph during her post-doc residency at LLNL in the year 2016. The second part of the experiment (from now on will be referred as abrasive peeling) focuses on understanding (1) the Np movement through the clay core and (2) the influence of temperature (25°C, 65°C, and 85°C) for the Np transport across the clay core under anaerobic conditions. End of 2016, the diffusion core samples were embedded in epoxy resin to stabilize them for long-term. The abraded core powder samples vary in material (epoxy vs. clay), Np activity and concentration, temperature exposed during the diffusion experiment, and mass, or amount of clay peeled off. Throughout the course of the 10-week internship program at LLNL, the experiments implemented and discussed in this report aimed to answer the questions: (1) How is the Np(IV) diffusion profile through bentonite influenced by the temperature? and (2) Will Np(IV) migrate through the geoengineered barrier of a nuclear waste repository? In addition, Ph.D. candidate Di Pietro received invaluable guidance in experimental design and development that will benefit her ongoing research endeavors.

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DOE OSTI2020

2020 LANL Contributions to the BATS Test in WIPP

The DOE Office of Nuclear Energy (DOE-NE) repository research and development program seeks to provide a sound technical basis for multiple viable disposal options, increase confidence in the robustness of generic disposal concepts, and develop science and engineering tools needed to support disposal concept implementation. Sandia, Los Alamos, and Lawrence Berkeley National labs are conducting research into salt, which includes the Brine Availability Test in Salt (BATS) field test at the Waste Isolation Pilot Plant (WIPP), a DOE Office of Environmental Management facility. BATS is leveraging the existing infrastructure associated with the WIPP to advance science informing generic disposal concepts. BATS is an ongoing heated borehole experiment being conducted in the WIPP underground. The goal of this experiment is to reduce the uncertainty associated with spent nuclear fuel disposition in geologic salt formations. The BATS experiments is designed to accomplish this by increasing our understanding of brine migration in salt, assessing damaged zones from mining and drilling, simulating a post-closure environment, confirming salt properties, and providing data for model validation. Phased testing for generic salt repository R&D was proposed in 2015. Subsequently, borehole thermal testing was developed as a first step in this process. This document summarizes the work conducted on the BATS experiment during fiscal year 2020 with an emphasis on the contributions by Los Alamos National Lab.

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DOE OSTI2020

Advances in Uncertainty and Sensitivity Analysis Methods and Applications in GDSA Framework.

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (FCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling. These priorities are directly addressed in the SFWST ''Geologic Disposal Safety Assessment'' (GDSA) control account, which is charged with developing a geologic repository system modeling and analysis capability, and the associated software, ''GDSA Framework'', for evaluating disposal system performance for nuclear waste in geologic media. ''GDSA Framework'' is supported by SFWST Campaign and its predecessor the Used Fuel Disposition (UFD) campaign. This report fulfills the GDSA Uncertainty and Sensitivity Analysis Methods work package (SF-20SN01030403) level 3 milestone — ''Advances in Uncertainty and Sensitivity Analysis Methods and Applications in GDSA Framework'' (M3SF-20SN010304032). It presents high level objectives and strategy for development of uncertainty and sensitivity analysis tools, demonstrates uncertainty quantification (UQ) and sensitivity analysis (SA) tools in GDSA Framework in FY20, and describes additional UQ/SA tools whose future implementation would enhance the UQ/SA capability of ''GDSA Framework''. This work was closely coordinated with the other Sandia National Laboratory GDSA work packages: the GDSA Framework Development work package (SF- 2051\101030404), the GDSA Repository Systems Analysis work package (SF-2051\101030405), and the GDSA PFLOTRAN Development work package (SF-20SN01030406). This report builds on developments reported in previous ''GDSA Framework'' milestones, particularly M2SF- 19SNO1030403.

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DOE OSTI2017

International Collaboration Activities on Engineered Barrier Systems

The Spent Fuel Waste Science and Technology (SFWST) campaign from the DOE Fuel Cycle and Technology (FCT) program has been engaging in international collaborations between repository R&D programs for nuclear waste disposal to leverage on the extensive research investigations and laboratory/field data of engineered barrier system (EBS) components (e.g., near-field) and characterization of transport phenomena in the host rock (e.g., far-field) processes from state-of-the-art underground research laboratories (URL) experiments. Thermal heating from radionuclide decay in the waste canisters will generate increases in temperature that will drive chemical and transport processes in the near- and far-field domains of the repository. URL sites provide the ideal setting to conduct heater test experiments to simulate the thermal effects of heat-generating nuclear waste in disposal galleries and surrounding host rock.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Decovalex-2019 (Task A Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. This document is the final report of Task A which was proposed by the DECOVALEX Project and coordinated by the British Geological Survey (BGS), presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task A under DECOVALEX-2019 is to better understand the processes governing the advective movement of gas in both low-permeability argillaceous repository host rocks and clay-based engineered barriers. Special attention is given to the mechanisms controlling gas entry, flow and pathway sealing and their impact on the performance of the engineered clay barrier. Previous work suggests gas flow may be accompanied by the creation of dilatant pathways whose properties change temporally and spatially within the medium. Thus, new numerical representations for the quantitative prediction of gas migration fluxes through argillaceous rock formations have been developed. These provide an invaluable tool with which to assess the impact of gas flow on repository layout and therefore design of any future facility. In addition, experience gained through this task is of direct relevance to other clay-based engineering issues where immiscible gas flow is a consideration including shale gas, hydrocarbon migration, carbon capture and storage and landfill design.

58 GEOSCIENCES↗
DOE OSTI2020

DECOVALEX-2019 (Task G Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. This document is the final report of Task G which was proposed by the Swedish Radiation Safety Authority (SSM) and coordinated by geomecon GmbH, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. Task G in DECOVALEX-2019 is primarily concerned with the evolution of transmissivity throughout the lifetime of a repository for radioactive waste, and in particular of spent nuclear fuel, in sparsely fractured and competent crystalline rock masses, and the representation of these processes via numerical simulation. The numerical simulations were to be evaluated for their suitability, and to be validated. Over the course of the work presented here, not only were transmissivity changes and the best approaches to simulate those change examined, but also strategies were to be developed on how to monitor transmissivity change. Guidelines on what a control program and a monitoring system for a repository should include and be able to measure, and how this can be done in practice, was a key objective of the research project.

58 GEOSCIENCES↗
DOE OSTI2018

Reporting the Radionuclide Inventory of the DWPF Product

In the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), the radionuclides contained in the liquid high-level waste currently stored at SRS are being converted into a durable borosilicate glass. This glass in its sealed stainless-steel canister, the DWPF product, eventually must be sent to a federal repository for final disposal. In order to ensure that the DWPF product is suitable for acceptance by the U. S. Department of Energy’s (DOE) Civilian Radioactive Waste Management System (CRWMS)*, the DOE Office of Environmental Management (DOE-EM) established the Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS). DOE-EM is tentatively performing surrogate CRWMS waste acceptance functions, as the office was eliminated in 2009, for review and concurrence of technical baseline documents and waste form non conformances. The specifications are divided into five sections, dealing with the waste form (borosilicate glass) itself, the canister, the canistered waste form, quality assurance, and documentation and other administrative requirements. The DWPF is required to document its compliance with the specifications in the Waste Form Compliance Plan (WCP), the Waste Form Qualification Report (WQR), the Production Records, and the Storage and Shipping Records. The WCP provides general information about the DWPF process and product, and a general description of the methods and programs by which the DWPF will demonstrate compliance with each specification in the WAPS. The WCP was vital to DWPF startup with respect to operations planning and also assurance to DOE that the canistered waste forms would be acceptable for disposal at any future High-Level Waste (HLW) geologic repository. The WQR is a compilation of the results of the testing and analysis programs identified in the WCP. The WQR details the method of compliance with each specification, and the testing performed to demonstrate the effectiveness of each method. The WQR is published in separate volumes which each correspond to one or more of the specifications. The WQR also contains a description of the DWPF’s Glass Product Control Program. Each volume contains Part and Item designators which indicate the corresponding item(s) in the WCP. The Production Records, and the Storage and Shipping Records document that each DWPF product has been produced and stored in compliance with the WAPS. This document, Volume 4 of the WQR, responds to WAPS Specifications 1.2.2, 1.6, and 3.14 which require that the DWPF report the radionuclide inventory of each canistered waste form, and of each waste type. This volume describes the systems the DWPF uses to determine and report the radionuclide inventory of the glass product, and provides evidence that these systems allow the DWPF to satisfy the WAPS specification. This volume of the WQR has been prepared to establish that methods are available to determine and report the radionuclide inventory, and that the DWPF is ready to collect the facility-specific data necessary to provide definitive estimates of the errors associated with the use of these methods.

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DOE OSTI2021

Cloud-Feedback Model Intercomparison Project: Tier 2 Simulations (Final Report)

The University of Miami (Subcontractor)’s Research Scientist James Benedict (with oversight by PI Amy Clement) was tasked with producing global climate model simulations as part of the Cloud-Feedback Model Intercomparison Project (CFMIP) Tier 2 protocol, managing the output of these simulations, providing assistance (as requested) to staff at Lawrence Livermore National Lab (LLNL) regarding model output and setup, and meeting with LLNL scientists and team members to coordinate work on the project. The simulations represent an important contribution to the CFMIP data repository and will advance understanding of a wide range of critical cloud, circulation, and precipitation responses to climate change. The Subcontractor completed all proposed simulations, formatted the model output to be compliant with CFMIP protocols, and published the model data to the CFMIP repository. When requested, the Subcontractor provided assistance to LLNL for model output and configuration queries. Benedict and/or Clement also met in person or virtually 1-2 times per year with project scientists to coordinate work and review results.

54 ENVIRONMENTAL SCIENCES↗