Search NASA⌕ Search

SEARCH · Search NASA

Results for “REPOSITORY”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Performance Confirmation Strategies for the Waste Isolation Pilot Plant - A Historical Perspective from an Operating Disposal Facility - 12248

Performance confirmation is an important element of the Waste Isolation Pilot Plant (WIPP) program. Performance confirmation was first used during the early WIPP site characterization phase to focus experimental activities that address the development of probabilistic repository performance models and to address stakeholder assurance needs. The program is currently used to analyze the conditions of the repository and its surroundings to ensure that the basis for the repository's long-term radioactive waste containment predictions is valid. This basis is related to the parameters, assumptions, conceptual and numerical models that are used to predict or validate the potential radioactive waste containment performance of the system. The concept of performance confirmation for the WIPP is one that has evolved since the first repository work was initiated decades ago and plays an important role in assuring adequate repository performance both now and in the long-term. The WIPP mission has progressed from a pilot project to an operational disposal facility and will progress to eventual site closure when disposal operations are completed. Performance confirmation is an important part of each of these progressions. The concept of disposing radioactive waste in a geologic repository today involves a complete understanding of many technical, political, regulatory, societal and economic elements. Many of these elements overlap and solving all relevant issues necessary to site, operate and decommission a disposal facility should be done with knowledge of each element's requirements and impacts. Performance confirmation is one tool that can help to coordinate many of these elements into a program that actively investigates what is thought to be adequately understood about the system and what information is lacking. A performance confirmation program is used to determine ways to challenge and verify those areas that are thought to be understood and to find ways to understand those areas that are not well understood. Performance confirmation programs have been used twice at WIPP, first during site characterization and PA development and later in a Compliance Monitoring program. At first, only certain technical aspects of the system were deemed important because it was a scientifically-based, government project. Early site characterization work was design to gather information about the geology and hydrology of the area and the mechanical properties of the natural barrier. The information would be used in a PA to determine the long-term containment performance of the disposal system. A performance confirmation element was used to identify the sensitive elements of the system that were certain, well understood or justified and those that were not. It identified experimental and analytical programs that could be used to reduce uncertainty, confirm sensitive assumptions and provide useful data. This performance confirmation program provided data to justify the adequacy of the information used in PA to demonstrate compliance with EPA's containment requirements. Performance confirmation will continue to be used in the post-closure period for at least 30 years and likely up to the end of the 100-year institutional controls period. As the technical basis for the repository matures throughout the operational period, the currently planned post-closure monitoring program will need to be reassessed prior to implementation. However, the intent of the program will be the same as it was for the previous programs, to ensure the ultimate goal of the repository. This goal is to safely isolate waste from the accessible environment and ensure public and environmental safety. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Investigations of dual-purpose canister direct disposal feasibility - 15106

Direct disposal of commercial spent nuclear fuel (SNF) in dual-purpose canisters (DPCs) could avoid the cost and complexity of re-packaging spent fuel into smaller purpose-built containers, avoid disposing of the DPC hulls as low-level waste, and potentially decrease worker dose. Technical objectives for disposal are: 1) safety of workers and the public, 2) engineering feasibility; 3) thermal management; and 4) criticality control after permanent closure of a repository. The postclosure safety case for DPC direct disposal would resemble that for any repository-waste isolation would be enhanced by choosing host geology in which radionuclide transport is diffusion dominated, and for which transport properties and pathways are insensitive to the expected temperature history. Containment functions would be assigned to the disposal overpack, and published data show that suitable materials exist for most possible disposal environments. Earlier studies showed that DPC-based waste packages would be only slightly larger and heavier than those proposed for a repository in volcanic tuff. Engineered solutions are available for transporting and emplacing them underground, although some could be the largest of their kind. The cost of DPC direct disposal in various geologic settings has been compared to options involving a change to loading standardized multi-purpose canisters, or continuing to load DPCs and re-packaging all of them for disposal. In general DPC direct disposal would cost tens of billions of dollars less, in packaging and disposal costs. Calculations for emplacement in hard rock and sedimentary rock show that host rock peak temperature targets of 100 to 200 deg.C could be met by managing fuel age and burnup, with repository closure at up to 150 years from reactor discharge. For disposal in salt, DPC-based waste packages could be emplaced at only 50 to 100 years from discharge, depending on burnup. Sedimentary media such as shales pose special challenges especially if they have low thermal conductivity. If backfill is used around DPC-based packages its peak temperature could be 150 to 200 deg.C. Without flooding of waste packages with ground water, criticality can never occur. However, some small number of disposal overpacks could fail within the postclosure performance period due to corrosion or disruptive events. Once flooded, the aluminum-based neutron absorber materials used in most DPCs would readily degrade. Analysis has shown that even without neutron absorbers, virtually all DPCs could be subcritical if flooded with chloride brines such as would occur in a salt repository. The licensing basis for existing DPCs involves criticality analysis that may be conservative compared to fuel that is actually loaded, and the difference provides uncredited reactivity margin. Uncredited margin can offset increases in reactivity associated with flooding and basket degradation. For flooding with fresh water, with loss of neutron absorbers, many DPCs have been shown to be subcritical using uncredited margin. Significantly fewer are subcritical if the fuel basket also degrades. Thus, the materials of DPC basket construction (e.g., aluminum or carbon steel) could be deleterious if they fully corrode, along with the neutron absorbers, during the performance period. Logistical simulations are used to better understand the relationship between the needed DPC decay storage time for disposal, and the timing of future events such as the repository opening date, or a transition to loading multi-purpose (disposable) canisters at nuclear power plants. Results show that the maximum benefit from implementing a new canister design, with respect to shortening the cooling time for all SNF including that in DPCs, is associated with the earliest repository start date. As time passes more of the total SNF inventory will be in DPCs such that the value of a transition to multi-purpose canisters will decline, and the remaining options are DPC direct disposal and re-packaging in purpose-designed disposal canisters. Fuel age at emplacement is of interest to evaluate potential impacts from future changes in the fuel management system that could place fuel age limits on dry storage or transportation. The minimum fuel age at emplacement (best-case) is obtained by re-packaging all DPCs into smaller canisters for disposal, thus drastically decreasing the needed decay storage time for any geologic disposal setting. If the industry transitions to smaller, multi-purpose canisters, and without re-packaging, the fuel age at emplacement would be comparable to the best case if: 1) the emplacement power limit is high enough, or 2) both the transition and the repository start date occur early (e.g., 2036 was analyzed). Technical feasibility evaluation continues to show that direct disposal could be safe and cost effective, for some DPCs and in some geologic settings. Direct disposal of a substantial fraction of existing DPCs, limited mainly by thermal management, uncredited criticality margin, materials of DPC basket construction, and the salinity of ground water in the disposal environment, could be part of the overall waste management system. (authors)

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

Deep Isolation-Development of the Safety Case for Disposal of Radioactive Wastes in Horizontal Boreholes - 20028

Deep Isolation has developed a safe, secure, and permanent geological disposal method for high-level waste, including spent nuclear fuel as well as sealed sources and other highly radioactive materials. The method leverages well established directional drilling technology to create horizontal repositories deep underground. The combination of great depth (1-3 km) and the ability to precisely position repositories in a horizontal orientation provides access to geologic strata that are inaccessible for typical mined and deep vertical borehole repositories. A number of potential direct and indirect safety benefits accrue including:1) greater depth below regional freshwater aquifers; 2) increased flexibility and more geologic options for siting; 3) access to deep formations with persistent and sustained reducing conditions. In addition there are a number of safety elements related specifically to the horizontal repository geometry including: 1) passive direction of thermally driven fluid and radionuclide movement away from the vertical access hole and toward the 'dead end' portion of the repository; 2) mitigation of seismic hazards by orienting repositories parallel to local and regional fault structures. In this paper we explore and discuss some of the key technologic, geologic and hydrologic elements that support the deep horizontal borehole safety case. The stalemate seen across the globe on the disposal of spent nuclear fuel and high-level waste can be broken. Deep Isolation offers a novel option for safe, secure, and permanent deep geological disposal of nuclear waste that can be developed as centralized repositories or adapted to smaller regional or site-specific repositories located near waste sources. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Models, data, and scripts associated with “Prediction of Distributed River Sediment Respiration Rates using Community-Generated Data and Machine Learning”

This data package is associated with the publication “Prediction of Distributed River Sediment Respiration Rates using Community-Generated Data and Machine Learning’’ submitted to the Journal of Geophysical Research: Machine Learning and Computation (Scheibe et al. 2024). River sediment respiration observations are expensive and labor intensive to obtain and there is no physical model for predicting this quantity. The Worldwide Hydrobiogeochemisty Observation Network for Dynamic River Systems (WHONDRS) observational data set (Goldman et al.; 2020) is used to train machine learning (ML) models to predict respiration rates at unsampled sites. This repository archives training data, ML models, predictions, and model evaluation results for the purposes of reproducibility of the results in the associated manuscript and community reuse of the ML models trained in this project. One of the key challenges in this work was to find an optimum configuration for machine learning models to work with this feature-rich (i.e. 100+ possible input variables) data set. Here, we used a two-tiered approach to managing the analysis of this complex data set: 1) a stacked ensemble of ML models that can automatically optimize hyperparameters to accelerate the process of model selection and tuning and 2) feature permutation importance to iteratively select the most important features (i.e. inputs) to the ML models. The major elements of this ML workflow are modular, portable, open, and cloud-based, thus making this implementation a potential template for other applications. This data package is associated with the GitHub repository found at Please see the file level metadata (flmd; “sl-archive-whondrs_flmd.csv”) for a list of all files contained in this data package and descriptions for each. Please see the data dictionary (dd; “sl-archive-whondrs_dd.csv”) for a list of all column headers contained within comma separated value (csv) files in this data package and descriptions for each. The GitHub repository is organized into five top-level directories: (1) “input_data” holds the training data for the ML models; (2) “ml_models” holds machine learning models trained on the data in “input_data”; (3) “scripts” contains data preprocessing and postprocessing scripts and intermediate results specific to this data set that bookend the ML workflow; (4) “examples” contains the visualization of the results in this repository including plotting scripts for the manuscript (e.g., model evaluation, FPI results) and scripts for running predictions with the ML models (i.e., reusing the trained ML models); (5) “output_data” holds the overall results of the ML model on that branch. Each trained ML model resides on its own branch in the repository; this means that inputs and outputs can be different branch-to-branch. Furthermore, depending on the number of features used to train the ML models, the preprocessing and postprocessing scripts, and their intermediate results, can also be different branch-to-branch. The “main-*” branches are meant to be starting points (i.e. trunks) for each model branch (i.e. sprouts). Please see the Branch Navigation section in the top-level README.md in the GitHub repository for more details. There is also one hidden directory “.github/workflows”. This hidden directory contains information for how to run the ML workflow as an end-to-end automated GitHub Action but it is not needed for reusing the ML models archived here. Please the top-level README.md in the GitHub repository for more details on the automation.

13C↗
DOE OSTI2013

Temperature-package power correlations for open-mode geologic disposal concepts.

Logistical simulation of spent nuclear fuel (SNF) management in the U.S. combines storage, transportation and disposal elements to evaluate schedule, cost and other resources needed for all major operations leading to final geologic disposal. Geologic repository reference options are associated with limits on waste package thermal power output at emplacement, in order to meet limits on peak temperature for certain key engineered and natural barriers. These package power limits are used in logistical simulation software such as CALVIN, as threshold requirements that must be met by means of decay storage or SNF blending in waste packages, before emplacement in a repository. Geologic repository reference options include enclosed modes developed for crystalline rock, clay or shale, and salt. In addition, a further need has been addressed for open modes in which SNF can be emplaced in a repository, then ventilated for decades or longer to remove heat, prior to permanent repository closure. For each open mode disposal concept there are specified durations for surface decay storage (prior to emplacement), repository ventilation, and repository closure operations. This study simulates those steps for several timing cases, and for SNF with three fuel-burnup characteristics, to develop package power limits at which waste packages can be emplaced without exceeding specified temperature limits many years later after permanent closure. The results are presented in the form of correlations that span a range of package power and peak postclosure temperature, for each open-mode disposal concept, and for each timing case. Given a particular temperature limit value, the corresponding package power limit for each case can be selected for use in CALVIN and similar tools.

Hardin, Ernest.↗
DOE OSTI2020

Engineered Barrier System R&D and International Collaborations – LANL (FY20): Spent Fuel and Waste Disposition

The United States Department of Energy’s Spent Fuel and Waste Disposition (SFWD) program is investigating the design and safety function of generic nuclear geologic repositories in a variety of geologic settings (salt, argillite, and crystalline rock). Different configurations and loadings of spent nuclear fuel and waste within disposal canisters are also being investigated, some of which have the potential to generate repository temperatures higher than previously considered (i.e., temperatures >100ºC) by foreign and domestic concepts. This report expands on engineered barrier material stability in a high temperature crystalline rock repository through high temperature hydrothermal experiments. Experiments were designed to develop engineered barrier system (EBS) concepts in a hightemperature crystalline environment in 1) bentonite-Grimsel Granodiorite interactions, 2) bentonite-cement reactions, and 3) interaction between waste canister materials and bentonite. Experiment results are applied to understanding long-term repository performance in terms of radionuclide isolation. One hydrothermal experiment was completed in the rocking autoclaves at LANL in FY-20: IEBS-6 (Grimsel Granodiorite + Wyoming bentonite + cured ordinary Portland cement + Grimsel Granodiorite synthetic groundwater, 250ºC/150 bar, 8 weeks). Several other experiments were planned but were delayed due to the COVID-19 pause in laboratory work. In addition, some characterization of the reaction products of experiments conducted in FY-20 was prevented by COVID-19. Mineral phase chemistry was not measured via electron microprobe analyses for IEBS-6. Quantitative X-ray diffraction results from IEBS-6 and HBT-1 were not completed. The missing analyses and discussion of the results will be included in next year’s report. New characterization that was completed includes scanning electron microscopy of reaction products from IEBS-6, quantitative X-ray diffraction results are presented from IEBS-1 through IBES-5, preliminary scanning electron microscope images and chemical analyses for IEBS-6, X-ray diffraction of the clay fraction from all experiments, measurement/imaging of mineral growth on the surface of steel coupons. Major observations pertaining to bentonite stability in a Grimsel Granodiorite environment include the stability of Na-montmorillonite at 250°C, the formation of trace CSH phases, and the formation of bentonite colloids on experiment cooling. The addition of a cured chip of Portland cement to the bentonite-Grimsel system results in slightly higher pH values and the formation of diverse secondary mineral phases that were not observed in the previous experiments (e.g., analcime, garronite, CSH phases). The new characterization efforts related to the interaction of stainless-steel coupons and bentonite clay focused on thickness and mineralogy of phases that formed at the steel surface. In Wyoming bentonite + Grimsel Granodiorite systems, newly formed minerals at the bentonitesteel coupon interface included alteration of the outermost steel edge to Fe,Cr-oxide phases, followed by Fe-rich phyllosilicates (Fe-saponite, chlorite) and interbedded Fe,Cr,Ni-sulfide phases (pentlandite). Hydrothermal experiments were completed to assess uranium-carbonate complexation at conditions relevant to high-temperature disposal. Autoclave solubility experiments were conducted at 150 to 250ºC with a range of carbonate and uranium concentrations. The experiment results were characterized via situ UV-Visible spectroscopy and synchrotron-based in situ XAS techniques. Results show a significant decrease in the stability of uranyl-carbonate complexes at temperatures above 100ºC along with the precipitation of uranium oxides. Further, at T > 200ºC, results show that uranyl-hydroxyl complexes control solubility of uranium instead of the previously predicted uranyl-carbonate species. These results are significant for understanding the mobility of uranium in the EBS, which will likely contain carbonate-rich fluids. International research efforts focused on three main areas: 1) participation in international conferences, 2) building collaborations with foreign repository programs, and 3) the initiation of an experimental program to complement the full-scale HotBENT test at the Grimsel test site. This experiment included Wyoming bentonite + low carbon steel + Grimsel Granodiorite synthetic groundwater and was run at the planned maximum temperature of the HotBENT test (200°C). Complete characterization of reaction products was hindered by disruptions to laboratory work but will be reported in the next FY. The experimental results obtained in FY-20 continue to document the wide-ranging effects of bulk composition and pressure-temperature conditions in the mineralogical and geochemical evolution of a high-temperature repository environment. Concepts developed will be used to inform models of long-term material stability in a generic crystalline rock-hosted repository.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

An Updated Analysis of Clad Degradation

This report provides an analysis of the clad barrier function associated with the direct disposal of dual purpose canisters (DPCs) under hypothetical conditions in a shale repository and in an alluvial repository, including the effect of a postulated criticality event inside a disposed DPC. Should a postulated criticality event occur in a hypothetical shale repository, cladding will primarily degrade by general corrosion. Stress corrosion cracking, hydride cracking, creep failure, pitting and crevice corrosion, rod pressurization, and clad unzipping are calculated to have little impact on cladding persistence. At the higher temperature expected during a postulated criticality event in a saturated shale repository, general corrosion of cladding would be rapid - on the order of 0.034 microns/yr. A few hundred years after onset of a postulated criticality event in a shale repository complete general corrosion of fuel assembly grid spacer walls and guide tubes will likely result in settling of fuel rods upon each other. This rod consolidation should displace the water moderator and possibly terminate a postulated criticality. The primary potential degradation pathway for cladding in a hypothetical alluvial repository is localized corrosion by fluoride, which cannot occur in a shale repository. Fluoride-enhanced corrosion of cladding would be accelerated under the slightly higher (< 100°C) temperatures associated with a postulated criticality event. The impact of criticality in both cases (shale and alluvial) would be to increase the amount of failed cladding. But it would require very specialized transport pathways.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

DPC Disposal Concepts of Operations. Final Report

The U.S. Department of Energy supports an R&D program for evaluating approaches to direct disposal of commercial spent fuel in dual-purpose canisters (DPCs). The major thrusts include alternative measures for treating the possibility of internal criticality events in DPC-based waste packages after thousands of years in a repository. These measures include: 1) injectable fillers, 2) analysis of the consequences of criticality events in a repository should they occur, and 3) options for modifying fuel assemblies or baskets in DPCs at the time they are loaded. This report presents a snapshot of progress in each of these areas drawing on deliverable reports generated during FY18 through FY20. Another aspect of the R&D program is to develop concepts of operations for repositories that would permanently dispose of DPC-based waste packages, considering different generic host media (not site-specific). The idea is to examine whether the disposal of large, heavy, heat-generating waste packages is technically feasible, and to identify the engineering challenges that would arise during implementation of the different disposal concepts. Descriptions of repository features are presented for repositories in salt media, argillite (clay/shale) media, crystalline (e.g., granitic) media, and unsaturated media (considering either alluvium or hard rock). Thermal management criteria for each concept are presented in terms of the maximum waste package thermal power at emplacement, when the repository could be opened, and the duration of repository emplacement operations. The overall message of this report is that direct disposal of commercial spent fuel is technically feasible in different types of geologic host media, but that thermal management and postclosure criticality impose different constraints on each concept. Engineering challenges are recognized and discussed. Treatment of postclosure criticality is identified as an important technical question that receives the majority of attention in the R&D program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

DECOVALEX-2023: Task F2 Salt Final Report

The subject of Task F of DECOVALEX-2023 concerns performance assessment modelling of radioactive waste disposal in deep mined repositories. The primary objectives of Task F are to build confidence in the models, methods, and software used for performance assessment (PA) of deep geologic nuclear waste repositories, and/or to bring to the fore additional research and development needed to improve PA methodologies. In Task F2- (salt), these objectives have been accomplished through staged development and comparison of the models and methods used by participating teams in their PA frameworks. Coupled-process submodels and deterministic simulations of the entire PA model for a reference scenario for waste disposal in domal salt have been conducted. The task specification has been updated continuously since the initiation of the project to reflect the staged development of the conceptual repository model and performance metrics. Thermal, hydrological, mechanical, and chemical properties of individual components of the engineered and natural system were chosen for relevance by participating teams. The salt reference case system was characterized using data and measurements collected at relevant underground research laboratories (URLs), field sites, and simulation results from teams with specialized modelling capability. Participating teams made a wide range of model assumptions from compartmentalized networks to full 3D models of the salt formation. No single contributed model includes full-fidelity representation of all the features, events, and processes (FEPs) detailed in the task specification, but almost all features and processes are represented in at least one model. Despite differences in the modelling strategies developed by participating teams, all models indicate that salt compaction and radionuclide diffusion are key processes in the repository, and for the FEPs and model scenario considered, little of the disposed radionuclides will migrate beyond the repository seal over the 100,000 year simulations. In general, the model output quantities have the largest differences over the short term and near the waste. The models tend to be more similar further from waste and at later time. Disparities between the models are believed to be due to differing simplifications from the task specification, some of which are chosen simplifications to reduce complexity, and some are restrictions imposed by the modelling tools. A second round of this task has been accepted for DECOVALEX-2027 in conjunction with Task F1 on crystalline PA modelling. The future round includes waste package heating, improved modelling of salt creep closure, additional comparisons of coupled-process sub-models, and the impact of repository engineering design on radionuclide migration in the repository. Participants will also propose and finalize a set of uncertain inputs for the reference case simulations, propagate these uncertainties in a set of realizations, and conduct sensitivity analyses on the simulation results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Effect of Anoxic Iron Corrosion on WIPP Brine Geochemistry FY23 Final Report (U)

A 280-day study was completed to evaluate the effect of zero-valent iron (Fe 0 ) on the Waste Isolation Pilot Plant (WIPP) brine geochemistry under anticipated reducing conditions. Hydrogen (H 2 ) gas is expected to be present in the repository after closure due to the anoxic corrosion of a vast quantity of iron contained in the waste forms disposed at WIPP; therefore, a background argon atmosphere containing H 2 was chosen for this study. WIPP groundwater brine pH and E h will impact the mobility and fate of plutonium within the repository. Modeling and laboratory results for Castile WIPP brine indicate that equilibrium fa values relative to the standard hydrogen electrode (SHE) are 40 mV more reducing (i.e., more negative) than those for Salado WIPP brine (-480 mV vs. -440 mV, respectively) because of the higher pH of the Castile brine (pH 9 .3 for Castile vs. pH 8.8 for Salado). The E h and pH data were corrected for the effects of high ionic strength. The experimental results for both brines are consistent with thermodynamic predictions using OLI Systems' Mixed Solvent Electrolyte chemical equilibrium model. The measured and corrected pH and E h data from this study are provided in Table ES-I and Table ES-2, respectively. The experimental study, with four test conditions in triplicate, was performed in a dual glovebox with a nominally 3 vol.% H 2 in argon atmosphere (target H 2 range: 3 ± I vol.%). Simulants containing MgO only ( experimental control) and MgO+Fe 0 (WIPP base case) were prepared for both the Salado and Castile brines. MgO was included in all simulants to account for the use of bulk magnesium oxide in the WIPP repository. Fe 0 was included in some simulants to incorporate the effects of the anoxic corrosion of iron and in-situ hydrogen generation in the study. The brine compositions were developed by Sandia National Laboratory (SNL; Xiong, 2008) and have been used in previous WIPP evaluations. The test method (agitation, etc.) is partially based on ASTM D3987-12. Twelve rounds of periodic measurements of pH and E h were performed over the course of the study. Chemical analysis results for liquids and solids (ICP-MS, ICP-ES, IC Anion, TIC, SEM-EDX) are consistent with the pH, E h , and thermodynamic modeling results. This study included the following conditions that deviate from anticipated post-closure conditions following brine intrusion, but were selected to facilitate bench-scale testing to validate modeling of pH and E h for the post-closure WIP P repository: an anoxic glove box atmosphere containing ≤ 4 vol. % H 2 vs. substantially higher H 2 gas concentrations assumed in the WIPP Performance Assessment (PA); a significantly higher liquid-to-solid test ratio compared to the much lower phase ratio anticipated in the WIP P repository; agitation of the simulant bottles to maximize mass transfer; and finally the use of Fe 0 reagents having a much greater surface area than expected in the WIP P repository. Non-representative conditions were chosen for various reasons such as: to provide bounding conservative results, to provide a margin of safety for testing, or to facilitate simulant sub-sampling and analysis. In a parallel effort, aqueous electrolyte thermodynamic models were developed for the synthetic Salado and Castile brines to inform the experimental design, facilitate laboratory data interpretation, and allow extension of evaluations beyond the parameters tested. Thermodynamic modeling simulations including the MgO and Fe 0 additives that are directly relevant to the experimental measurements (e.g., pH calibration curve, ORP corrections) are included in this report. The measured fa of the simulants was close to the OLI model predictions for both brines and was largely controlled by the background H 2 partial pressure in the vapor phase as well as H 2 generated in situ in the aqueous phase by the Fe 0 corrosion. The H 2 gas-phase concentration tested and thermodynamically evaluated was much lower than is assumed in the WIPP PA; however, H 2 (g) concentrations significantly below this level are still predicted to result in very reducing conditions. In conclusion: • The experimental results are consistent with thermodynamic model predictions for fa, pH, and the effects of high ionic strength. • Evidence to date suggests that the H2 concentration in the glovebox atmosphere ultimately determined the final E h values of the simulants and resulted in highly reducing conditions. As a result, little difference was observed between the control simulants containing only MgO and the WIPP base-case simulants that contained MgO and Fe 0 . • This test methodology is recommended for future studies evaluating WIPP repository conditions. The methodology includes: (1) background H 2 in argon with agitation ( or could alternatively include in-situ-generated H 2 in sealed bottles); (2) carefully measured and corrected ORP data ( with much effort focused on allowing the probes to fully stabilize); and (3) ionic-strength-corrected pH data. Other best practices, such as simulant sparging/handling, ORP probe replacement, etc., should also be considered. • The coupling of experimental studies and thermodynamic modeling is also highly recommended because these methods inform and direct one another leading to greater confidence in and understanding of the results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

In-Situ Testing and Performance Assessment of a Redesigned WIPP Panel Closure - 13192

There are two primary regulatory requirements for Panel Closures at the Waste Isolation Pilot Plant (WIPP), the nation's only deep geologic repository for defense related Transuranic (TRU) and Mixed TRU waste. The Federal requirement is through 40 CFR 191 and 194, promulgated by the U.S. Environmental Protection Agency (EPA). The state requirement is regulated through the authority of the Secretary of the New Mexico Environment Department (NMED) under the New Mexico Hazardous Waste Act (HWA), New Mexico Statutes Annotated (NMSA) 1978, chap. 74-4-1 through 74-4-14, in accordance with the New Mexico Hazardous Waste Management Regulations (HWMR), 20.4.1 New Mexico Annotated Code (NMAC). The state regulations are implemented for the operational period of waste emplacement plus 30 years whereas the federal requirements are implemented from the operational period through 10,000 years. The 10,000 year federal requirement is related to the adequate representation of the panel closures in determining long-term performance of the repository. In Condition 1 of the Final Certification Rulemaking for 40 CFR Part 194, the EPA required a specific design for the panel closure system. The U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) has requested, through the Planned Change Request (PCR) process, that the EPA modify Condition 1 via its rulemaking process. The DOE has also requested, through the Permit Modification Request (PMR) process, that the NMED modify the approved panel closure system specified in Permit Attachment G1. The WIPP facility is carved out of a bedded salt formation 655 meters below the surface of southeast New Mexico. Condition 1 of the Final Certification Rulemaking specifies that the waste panels be closed using Option D which is a combination of a Salado mass concrete (SMC) monolith and an isolation/explosion block wall. The Option D design was also accepted as the panel closure of choice by the NMED. After twelve years of waste handling operations and a greater understanding of the waste and the behavior of the underground salt formation, the DOE has established a revised panel closure design. This revised design meets both the short-term NMED Permit requirements for the operational period, and also the Federal requirements for long-term repository performance. This new design is simpler, easier to construct and has less of an adverse impact on waste disposal operations than the originally approved Option D design. The Panel Closure Redesign is based on: (1) the results of in-situ constructability testing performed to determine run-of-mine salt reconsolidation parameters and how the characteristics of the bedded salt formation affect these parameters and, (2) the results of air flow analysis of the new design to determine that the limit for the migration of Volatile Organic Compounds (VOCs) will be met at the compliance point. Waste panel closures comprise a repository feature that has been represented in WIPP performance assessment (PA) since the original Compliance Certification Application of 1996. Panel closures are included in WIPP PA models principally because they are a part of the disposal system, not because they play a substantive role in inhibiting the release of radionuclides to the outside environment. The 1998 rulemaking that certified WIPP to receive transuranic waste placed conditions on the panel closure design to be implemented in the repository. The revised panel closure design, termed the Run-of-Mine (ROM) Panel Closure System (ROMPCS), is comprised of 30.48 meters of ROM salt with barriers at each end. The ROM salt is generated from ongoing mining operations at the WIPP and may be compacted and/or moistened as it is emplaced in a panel entry. The barriers consist of bulkheads, similar to those currently used in the panels as room closures. A WIPP performance assessment has been completed that incorporates the ROMPCS design into the representation of the repository, and compares repository performance to that achieved with the approved Option D design. Several key physical processes and rock mechanics principles are incorporated into the performance assessment. First, creep closure of the salt rock surrounding a panel entry results in consolidation of the ROM salt emplaced in the entry. Eventually, the ROM salt comprising the ROMPCS will approach a condition similar to intact salt. As the ROM salt reaches higher fractional densities during consolidation, back stress will be imposed on the surrounding rock mass leading to eventual healing of the disturbed rock zone above and below the panel closure. Healing of the disturbed rock zone above and below the ROMPCS reduces the porosity and permeability in those areas. Analysis of the new design demonstrates that: (1) the WIPP continues to meet regulatory compliance requirements when the ROMPCS design is implemented instead of Option D, and (2) there is no impact on the short-term effectiveness of the panel closure to limit the concentration of VOCs at the WIPP site boundary to a fraction of the health-based exposure limits (HBLs) during the operational period. (authors)

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

End of the (Fuel) Cycle: What can South Australia learn from the world? - 17047

This project consists of a study of five countries' nuclear waste policy histories, a review of the presence and role of bureaucracy, how these factors have influenced their citizens' attitudes towards nuclear waste and progress made towards the planning, siting and construction of nuclear waste repositories. A survey was sent to a selected stratified sub-group of the community questioning their attitude towards the location of a potential nuclear waste repository in South Australia and what the most influential factor responsible for their opinion was. France is currently in the process of converting its underground nuclear waste research facility in Bure into the country's first high-level nuclear waste repository. Sweden is also constructing an underground high level nuclear waste repository, after decades of planning, community engagement and candidate site selection processes. The Nuclear Energy Agency estimates that Switzerland will construct a low and intermediate level nuclear waste repository by 2035 and a high level nuclear waste repository by 2045. Finland will be the first country in the world to successfully construct and operate a high-level nuclear waste repository and it is expected to be operational in 2023. The project summarises the most influential factors affecting the success of five case-study countries in managing their nuclear and radioactive waste and apply them to an Australian context. It also highlights the outcome of the surveys and connects the result to those previously conducted in the five case-study countries. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

The SD-20K, a large Type B(U)F Package for Waste Containers - 18181

The type B(U)F package SD-20K, currently being developed by DAHER Nuclear Technologies GmbH (DNT), will facilitate the transport of legacy waste from interim storage facilities to the German final repository Konrad. The package SD-20K could also be used in other cases where legacy packages are not fit for a final repository and/or for transport but could be easily conditioned into a suitable package. Since the 1990's German utilities, reactor operators and service providers conditioned waste according to the requirements of the preliminary acceptance criteria for the final repository Konrad. Some of this waste was conditioned in containers which were at that time licensed as type B packages but for which the license expired due to various reasons. Other waste was reclassified so that the used package type does no longer comply with the content. In recent times Konrad containers of type IP-2 are loaded with e. g. core components in compliance with the Konrad acceptance criteria but overloaded with respect to the transport regulations. DAHER Nuclear Technologies GmbH (DNT) in cooperation with Siemens AG investigated possible solutions in a general approach. The result of the evaluation was that the best solution for the transport problem was to develop a versatile type B(U)F package large enough to accommodate all containers and packages accepted for the final repository Konrad. Possible solutions like repacking (high costs and high dose exposure) or requalification (high costs and risk of failure to achieve the goal) were not considered to be generally applicable. The SD-20K packaging is currently designed to accommodate Konrad Containers type III, IV and VI; however, the dimensions of its cavity are sufficient to accommodate any container and package that meets the acceptance criteria for the final repository Konrad. The SD-20K packaging consists of an inner container which constitutes the containment, surrounded by a rather large shock absorbing structure which also contains the thermal protection system. The outer shape of the SD-20K resembles a 20' ISO container. A special feature of the SD-20K is the automatic unloading system to minimize dose exposure of the personnel in the final repository. After arrival in the repository the shock absorbing structure as well as the door of the containment system are opened automatically and the content is moved out of the SD-20K by a remote-controlled carrier system. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

The future low-temperature geochemical data-scape as envisioned by the U.S. geochemical community

Data sharing benefits the researcher, the scientific community, and the public by allowing the impact of data to be generalized beyond one project and by making science more transparent. However, many scientific communities have not developed protocols or standards for publishing, citing, and versioning datasets. One community that lags in data management is that of low-temperature geochemistry (LTG). This paper resulted from an initiative from 2018 through 2020 to convene LTG and data scientists in the U.S. to strategize future management of LTG data. Through webinars, a workshop, a preprint, a townhall, and a community survey, the group of U.S. scientists discussed the landscape of data management for LTG – the data-scape. Currently this data-scape includes a “street bazaar” of data repositories. This was deemed appropriate in the same way that LTG scientists publish articles in many journals. The variety of data repositories and journals reflect that LTG scientists target many different scientific questions, produce data with extremely different structures and volumes, and utilize copious and complex metadata. Nonetheless, the group agreed that publication of LTG science must be accompanied by sharing of data in publicly accessible repositories, and, for sample-based data, registration of samples with globally unique persistent identifiers. LTG scientists should use certified data repositories that are either highly structured databases designed for specialized types of data, or unstructured generalized data systems. Recognizing the need for tools to enable search and cross-referencing across the proliferating data repositories, the group proposed that the overall data informatics paradigm in LTG should shift from “build data repository, data will come” to “publish data online, cybertools will find”. Funding agencies could also provide portals for LTG scientists to register funded projects and datasets, and forge approaches that cross national boundaries. Finally, the needed transformation of the LTG data culture requires emphasis in student education on science and management of data.

58 GEOSCIENCES↗
DOE OSTI2011

Radioactive Waste Partitioning and Transmutation within Advanced Fuel Cycles: Achievements and Challenges

If nuclear power should become a sustainable source of energy, a safe, robust and acceptable solution must be pursued for existing and projected inventories of high-activity, long-lived radioactive waste. Remarkable progress in the last two decades has been made in the field of geological disposal. Some countries have reached important milestones and geological disposal (of spent fuel) is expected to start in 2020 in Finland and in 2022 in Sweden and in fact the licensing of the geological repositories in both countries is now entering into their final phases. In France disposal of Intermediate Level Wastes (ILW) and vitrified High Level Wastes (HLW) is expected to start around 2025, according to the roadmap defined by a Parliament Act in 2006. In this context, transmutation of part of the waste through use of advanced fuel cycles, probably feasible in the coming decades, has the potential of reducing the burden on the geological repository. This article presents the physical principle of transmutation and reviews several strategies of P&T (Partitioning and Transmutation). Many recent studies have demonstrated that the impact of P&T on geological disposal concepts is not overwhelmingly high. However, by reducing waste heat production a more efficient utilization of repository space is likely. Moreover, even if radionuclide release from the waste to the environment and related calculated doses to the population are only partially reduced by P&T, it is important to point out that a clear reduction of the actinide inventory in the High Level Waste definitely reduces risks arising from less probable evolutions of a repository, i.e. increase of actinide mobility in certain geochemical situations and radiological impact by human intrusion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Self-accelerated corrosion of nuclear waste forms at material interfaces

The US plan for high-level nuclear waste includes the immobilization of long-lived radionuclides in glass or ceramic waste forms in stainless-steel canisters for disposal in deep geological repositories. Here we report that, under simulated repository conditions, corrosion could be significantly accelerated at the interfaces of different barrier materials, which has not been considered in the current safety and performance assessment models. Severe localized corrosion was found at the interfaces between stainless steel and a model nuclear waste glass and between stainless steel and a ceramic waste form. The accelerated corrosion can be attributed to changes of solution chemistry and local acidity/alkalinity within a confined space, which significantly alter the corrosion of both the waste-form materials and the metallic canisters. The corrosion that is accelerated by the interface interaction between dissimilar materials could profoundly impact the service life of the nuclear waste packages, which, therefore, should be carefully considered when evaluating the performance of waste forms and their packages. Moreover, compatible barriers should be selected to further optimize the performance of the geological repository system.

Guo, Xiaolei↗
DOE OSTI2018

A large-scale analysis of bioinformatics code on GitHub

In recent years, the explosion of genomic data and bioinformatic tools has been accompanied by a growing conversation around reproducibility of results and usability of software. However, the actual state of the body of bioinformatics software remains largely unknown. Here, the purpose of this paper is to investigate the state of source code in the bioinformatics community, specifically looking at relationships between code properties, development activity, developer communities, and software impact. To investigate these issues, we curated a list of 1,720 bioinformatics repositories on GitHub through their mention in peer-reviewed bioinformatics articles. Additionally, we included 23 high-profile repositories identified by their popularity in an online bioinformatics forum. We analyzed repository metadata, source code, development activity, and team dynamics using data made available publicly through the GitHub API, as well as article metadata. We found key relationships within our dataset, including: certain scientific topics are associated with more active code development and higher community interest in the repository; most of the code in the main dataset is written in dynamically typed languages, while most of the code in the high-profile set is statically typed; developer team size is associated with community engagement and high-profile repositories have larger teams; the proportion of female contributors decreases for high-profile repositories and with seniority level in author lists; and, multiple measures of project impact are associated with the simple variable of whether the code was modified at all after paper publication. In addition to providing the first large-scale analysis of bioinformatics code to our knowledge, our work will enable future analysis through publicly available data, code, and methods.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2011

Salt disposal of heat-generating nuclear waste.

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United States repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, helps define a clear strategy for a heat-generating nuclear waste repository in salt.

02 PETROLEUM↗