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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Initial Simulations of Empty Room Collapse and Reconsolidation at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is a geologic repository for defense-related nuclear waste. If left undisturbed, the virtually impermeable rock salt surrounding the repository will isolate the nuclear waste from the biosphere. If humans accidentally intrude into the repository in the future, then the likelihood of a radionuclide release to the biosphere will depend significantly on the porosity and permeability of the repository itself. Room ceilings and walls at the WIPP tend to collapse over time, causing rubble piles to form on floors of empty rooms. The surrounding rock formation will gradually compact these rubble piles until they eventually become solid salt, but the length of time for a rubble pile to reach a certain porosity and permeability is unknown. This report details the first efforts to build models to predict the porosity and permeability evolution of an empty room as it closes. Conventional geomechanical numerical methods would struggle to model empty room collapse and rubble pile consolidation, so three different meshless methods, the Immersed Isogeometric Analysis Meshfree, Reproducing Kernel Particle Method (RKPM), and the Conformal Reproducing Kernel method, were assessed. First, the meshless methods and the finite element method each simulated gradual room closure, without ceiling or wall collapse. All three methods produced equivalent room closure predictions with comparable computational speed. Second, the Immersed Isogeometric Analysis Meshfree method and RKPM simulated two-dimensional empty room collapse and rubble pile consolidation. Both methods successfully simulated large viscoplastic deformations, fracture, and rubble pile rearrangement to produce qualitatively realistic results. In addition to geomechanical simulations, the flow channels in damaged salt and crushed salt were measured using micro-computed tomography, and input into a computational fluid dynamics simulation to predict the salt's permeability. Although room for improvement exists, the current simulation approaches appear promising.

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DOE OSTI2019

Scenario Development: Proposal of a Methodology for Different Host Rocks and Disposal Concepts - 19007

Scenario development is an essential part of the assessment of the long-term safety of a repository for radioactive waste. The new act on site selection in Germany in 2017 requires that salt, clay and crystalline rock are studied as possible sites for repositories. So far, a comprehensive methodology for scenario development has been developed only for a site in salt rock in Germany within the Preliminary Safety Analysis (VSG). The question arose whether this methodology is transferable to other host rocks and disposal concepts and which modifications may be necessary. The transferability was shown in the project ANSICHT for clay rock sites while introducing some modifications of the methodology. This project studied the transferability of crystalline rock sites, however, there is no suitable safety and disposal concept or even a generic site yet. To study the transferability, initial barriers and FEP for different concepts were selected and the first steps of the methodology for crystalline rock were applied and discussed. It was concluded that the definition of barriers as containment providing rock zone (CPRZ) is difficult but important for the application of the methodology. Concepts for crystalline rock can foresee a CPRZ within the host rock or above. If the concept is using engineered barriers only and no CPRZ, the number of initial barriers is low. This will subsequently lead to a more complex scenario development since long FEP chains have to be considered. Based on the discussion so far, the transferability of the methodology seems to be feasible in general. (authors)

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DOE OSTI2017

Influence of Calcium Concentration on Permeability Changes in Flow-paths Altered by Highly Alkaline Groundwater - 17129

Groundwater around the geological repository of HLW is altered to the highly alkaline and Ca-rich condition by the constituents leaching from cementitious materials. Under such condition, the formation of CSH will affect the migration of radionuclides with the change in the hydraulic and chemical properties of the fracture (micro flow-path) in the host rock. In particular, the clogging effect with the deposition of CSH in the flow path the may contribute the retardation of radionuclide migration. In this study, the influence of Ca concentration on such clogging effect in the micro flow-path under the highly alkaline condition was examined by estimating the permeability changes with the deposition of CSH. In the experiment, the micro fractures in host rock were simulated by the micro flow-cell nipping the Teflon sheet (80 μm of thickness) between a granite chip and Teflon basement. The flow experiment was conducted by continuously injecting Ca solution to the micro flow-cell with constant flow rate. The concentrations of Ca ions for the injection solution were set in the range from 0 to 7.5 mM, and the value of pH was 12.5. As a result, under all conditions of Ca concentrations, the permeability in the flow path decreased to the range of 1. 10{sup -10} to 2.0x10{sup -10} m{sup 2} with the deposition of CSH. Besides, such decreased permeability was almost constant with increasing and decreasing periodically. Such increase and decrease in the permeability will be caused by the deposition and outflow of CSH formed in the flow path. In other words, because CSH is stable in pH higher than at least 10, under the condition with no man-made outlet such as real underground around the repository of radioactive waste, most CSH might clog through the flow-paths in the spatial range where pH exceeds at least 10, contributing the retardation effect of the migration of radionuclides released from the repository. (authors)

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DOE OSTI2024

RANGERS: Methodology Report on Design and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF

Salt formations are one of the potential host rocks for the final disposal of high-level radioactive waste (HLW) in deep geological repositories, both in Germany and the United States. The safe isolation of radioactive waste in these repositories relies on a multi-barrier system, combining engineered and natural barriers. The natural barrier is provided by the salt rock itself, known for its self-sealing properties and long-term stability. The engineered barrier, on the other hand, comprises sealing components strategically placed within the repository to enhance its containment capabilities. In both Germany and the United States, long-term safety assessments require demonstrating the integrity of the natural barrier for a period of up to 1 million years. Concurrently, the engineered barrier system (EBS) must maintain its structural and functional integrity until the long-term sealing, such as the granular salt backfill material, has re-consolidated to its final low porosity and permeability. Based on extensive expertise and experience with engineered barriers in salt formations, BGE TECHNOLOGY GmbH and Sandia National Laboratories have partnered to develop a robust methodology for the integrity and performance assessment of EBS in HLW repositories through the RANGERS project. This collaborative effort aims to establish a unified approach to geotechnical engineering, repository design, integrity and performance evaluation of EBS in salt repositories.

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DOE OSTI2024

Long-length Scintillating Fibers for Nuclear Geological Repositories

This research focuses on the usage of plastic scintillating fibers as ruggedized detectors in long-length sections coupled to long-length standard optical fiber segments for the purpose of radiation detection and monitoring. This technology addresses a need for monitoring radiation in difficult-to-access or inaccessible areas, such as monitoring: radioactive waste within geological repositories, nuclear material storage facilities, spent fuel dry storage areas, inaccessible areas within nuclear facilities including reactors and reprocessing facilities, decommissioning of nuclear facilities, and other similar areas. The fiber can be placed along the tunnels of a geological repository to monitor the radiation field along its length. This can detect an increase of radiation due to movement, leakage, or unexpected opening of the containers, as well as decrease in the radiation due to removal of material. These capabilities have been demonstrated at INL’s Radioactive Scrap Waste Facility by successfully detecting the removal of an interim storage containers from the storage area. The poster focuses on explaining the technology and its capability in monitoring deep geological repositories (DGRs), especially in the domain of safety and international safeguards. As more countries consider the development of DGRs, technologies such as the one being presented can help establish continuity of knowledge of the material, maintain public confidence, and ensure safe operation of the facility. Furthermore, the poster looks at the future steps of the technology and its potential role in the international safeguards regime.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2024

DECOVALEX-2023: Task F1 Final Report

DECOVALEX-2023 Task F is a comparison of models and methods for post-closure performance assessment (PA) of a deep geologic repository for radioactive waste. The general aims of Task F are to build confidence in the models, methods, and software used for PA and to stimulate additional research and development in PA methodologies. The task objectives are to motivate development of PA modelling skills and capabilities, to examine the influence of model choices on calculated repository performance, and to compare the uncertainties introduced by model choices to other sources of uncertainty. Task F involves no actual experiment or site. It is a PA modelling exercise that requires the conceptual development of hypothetical repository designs and geologic settings. Because three of the teams were interested in salt and the rest of the teams were interested in crystalline rock, Task F was split into two branches: Task F1 for crystalline rock and Task F2 for salt. This report is for Task F1, crystalline rock. Teams from seven countries (Canada, Czech Republic, Germany, Korea, Sweden, Taiwan, and United States) participated in Task F1. The teams worked together to define the features, events, and processes of the reference case repository and established a set of performance measures. In addition, they defined a set of benchmark problems designed to test and compare modelling capabilities for fracture flow and transport at different scales. The repository design and benchmark problems are documented in a Task Specification that evolved over time as the group honed the specifications. The benchmark problems verified that each team can aptly model flow and transport in fractured media in 1-, 2-, and 3-dimensions. Two general approaches were used for the 3-dimensional benchmarks: discrete fracture network (DFN) and equivalent continuous porous medium (ECPM). DFN modelling involves explicit meshing of each fracture while ECPM modelling aims to capture the effective porosity and directional permeability of each cell in a space-filling mesh as affected by intersecting fractures. In some models, a combination of the two is used, i.e., DFN for large known fractures and ECPM for the rest of the domain. Transport is solved by using either the advection-dispersion equation or particle tracking. Although some variation is observed among model breakthrough curves in the benchmark problems, there is strong agreement in breakthrough behaviour up to at least the 75 th percentile for all benchmarks. At the 90 th percentile, breakthrough results show larger differences, suggesting several models retain substantially higher fractions of tracer in regions of slower moving water. In addition to the flow and transport benchmarks, several teams completed the source term benchmark, verifying capabilities for modelling radionuclide decay and ingrowth, waste package breach, instant release fractions, fuel matrix degradation rates, and radionuclide solubility limitations. The reference case is conceptualized as a generic spent fuel repository at a depth of 450 m in fractured crystalline rock. The repository has 50 parallel backfilled drifts, each with 50 deposition holes 6 m apart. Each deposition hole contains a 4-PWR waste package and bentonite buffer. The rock domain is 5 km in length, 2 km in width, and 1 km in depth. It has 6 deterministic fractured deformation zones and a multitude of stochastic fractures. Teams generally used the ECPM approach for the entire rock or a hybrid approach in which the deterministic fracture zones are modelled with a DFN and the rest of the rock is modelled by ECPM. Of the reference case problems specified, only the results of the initial reference case problem are compared in this report. The initial problem focuses on transport from the deposition holes to the surface, i.e., it neglects waste package performance. Tracers are released at all waste package locations at time zero and tracked for their releases to the near field and ground surface. The water fluxes calculated at the ground surface entry and exit regions of the domain are similar for all models except for two that have considerably lower fluxes. For tracer transport, large differences are observed among models in the magnitude of tracer transported. Much of the difference appears to be due to how the repository is implemented and hence the different degrees of repository simplification. Models that exclude the drifts, buffer, and backfill from the domain tend to show greater release of tracers and radionuclides from the repository. The initial study presented here indicates that major differences in modelling important processes within the repository (e.g., diffusion through buffer and backfill) can produce broadly different release and transport results, especially when those processes are excluded. Even for the models that included all specified features, events, and processes, the results show significant differences and demonstrate the importance of examining multiple modelling approaches in performance assessment. The differences in results observed in this study are expected to motivate teams to either increase complexity in future versions of the reference case models or to improve methods to account for the effects of simplified features and processes. Either way, future improvements in these models are expected to produce results that more closely agree.

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DOE OSTI2018

Potential for Salt Repositories - 18693

Progress toward nuclear waste disposal in the US has been badly corrupted. Since passing and amending the Nuclear Waste Policy Act, the US is further away from geologic disposal than it was in 1987, which was 30 years ago. The waste inventory, liability costs, and elements for dealing with the burgeoning problem continue to increase in complexity - largely because of inaction. The Nuclear Waste Policy Act, as amended, remains the national law in the US. Even if a repository at Yucca Mountain can be licensed, constructed, operated, and closed, the national nuclear waste imperative will not be solved by these actions. Our national challenge includes broad technical problems that are exacerbated by national procrastination, while other countries suffer similarly. This paper discusses several roadblocks that deny progress in our nuclear waste-disposal program, many of which are shared internationally. Recommendations for moving this seemingly overwhelming problem toward solutions are also included. (author)

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DOE OSTI2025

Guiding Principles for Geochemical/Thermodynamic Model Development and Validation in Nuclear Waste Disposal: A Close Examination of Recent Thermodynamic Models for H + —Nd 3+ —NO 3 - (—Oxalate) Systems

Development of a defensible source-term model (STM), usually a thermodynamical model for radionuclide solubility calculations, is critical to a performance assessment (PA) of a geologic repository for nuclear waste disposal. Such a model is generally subjected to rigorous regulatory scrutiny. In this article, we highlight key guiding principles for STM model development and validation in nuclear waste management. We illustrate these principles by closely examining three recently developed thermodynamic models with the Pitzer formulism for aqueous H + —Nd 3+ —NO 3 - (—oxalate) systems in a reverse alphabetical order of the authors: the XW model developed by Xiong and Wang, the OWC model developed by Oakes et al., and the GLC model developed by Guignot et al., among which the XW model deals with trace activity coefficients for Nd(III), while the OWC and GLC models are for concentrated Nd(NO 3 ) 3 electrolyte solutions. The principles highlighted include the following: (1) Principle 1. Validation against independent experimental data: A model should be validated against experimental data or field observations that have not been used in the original model parameterization. We tested the XW model against multiple independent experimental data sets including electromotive force (EMF), solubility, water vapor, and water activity measurements. The results show that the XW model is accurate and valid for its intended use for predicting trace activity coefficients and therefore Nd solubility in repository environments. (2) Principle 2. Testing for relevant and sensitive variables: Solution pH is such a variable for an STM and easily acquirable. All three models are checked for their ability to predict pH conditions in Nd(NO 3 ) 3 electrolyte solutions. The OWC model fails to provide a reasonable estimate for solution pH conditions, thus casting serious doubt on its validity for a source-term calculation. In contrast, both the XW and GLC models predict close-to-neutral pH values, in agreement with experimental measurements. (3) Principle 3. Honoring physical constraints: Upon close examination, it is found that the Nd(III)-NO 3 association schema in the OWC model suffers from two shortcomings. Firstly, its second stepwise stability constant for Nd(NO 3 ) 2+ (log K 2 ) is much higher than the first stepwise stability constant for NdNO 3 2+ (log K 1 ), thus violating the general rule of (log K 2 –log K 1 ) < 0, or $\frac{K1}{K2}$>1. Secondly, the OWC model predicts abnormally high activity coefficients for Nd(NO 3 ) 2 + (up to ~900) as the concentration increases. (4) Principle 4. Minimizing degrees of freedom for model fitting: The OWC model with nine fitted parameters is compared with the GLC model with five fitted parameters, as both models apply to the concentrated region for Nd(NO 3 ) 3 electrolyte solutions. The latter appears superior to the former because the latter can fit osmotic coefficient data equally well with fewer model parameters. The work presented here thus illustrates the salient points of geochemical model development, selection, and validation in nuclear waste management.

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DOE OSTI2019

Spatiotemporal Distribution of {sup 241}Am in the Vicinity of the Waste Isolation Pilot Plant - 19543

Environmental monitoring around nuclear facilities is essential for characterizing the baseline conditions that are critical for identifying releases and quantifying their effects, should they occur, and for assessing post-closure performance. The spatiotemporal distribution and transport behavior of {sup 241}Am in the vicinity of a deep geologic repository for transuranic waste were assessed using data spanning the pre-operational phase to the present, including an accidental radionuclide release in 2014. The {sup 241}Am background values of 0.027±0.109 μBq/m{sup 3} (aerosols), and 0.061±0.065 Bq/kg (soil), as well as the {sup 241}Am/{sup 239+240}Pu ratios of 0.41±0.06 (aerosols) and 0.38±0.04 (soil), are similar to global fallout values. Pre-operations concentrations of {sup 241}Am show a strong seasonal dependence, attributed to resuspension of soils contaminated by global fallout, with no contribution from the upwind Project Gnome site ({sup 241}Am/{sup 239+240}Pu =0.24±0.08). During the 2014 accidental release, aerosol concentrations of {sup 241}Am briefly increased to 115.2 μBq/m{sup 3} at an onsite monitoring station, and 81.4 μBq/m{sup 3} at a station ∼1 km downwind of the exhaust shaft, changing the {sup 241}Am/{sup 239+240}Pu from ∼0.37 to ∼10. However, some four years later, conditions have returned to background. Almost 90% of the {sup 241}Am inventory is the 0-5 cm soil layer and there is no evidence of concentrations above regulatory standards, or that can be attributed to waste disposal operations. (authors)

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DOE OSTI2022

Examining Autonomous Inspection of Geologic Repositories

Geological repositories for nuclear waste, including spent nuclear fuel, present a significant challenge for traditional International Atomic Energy Agency (IAEA) safeguards tools due to their inaccessibility and demanding operational conditions. The IAEA has been working closely with Member State organizations currently involved in repository construction and planning including Euratom, the Finnish and Swedish regulatory authorities, and relevant facility operators. However, the verification challenge remains unsolved, and there persists an out-standing need for tools and approaches that will help the IAEA verify that no nuclear material is diverted from a repository environment. The challenge is also not static as activities must encompass verification of the design prior to and during the construction/operation phase, and post backfill. Throughout these various phases, it is imperative that the IAEA maintains a continuity of knowledge (CoK) of all material, including information on material inventory and flow. This paper highlights these challenges and outlines how they might be addressed by using remote or autonomous vehicles. Specifically, it discusses the current state of the art in robotic autonomy for known or partially known environment mapping and patrolling, as well as shared autonomy, where humans collaborate with closed loop autonomation to complete tasks. The feasibility of using rovers for these verification tasks is explored, along with the challenges associated with system implementation. Hardware and software suggestions are provided based on the adoption of similar technologies in other comparable areas and ability to close technical gaps. Finally, human-robotic interactions are considered based on the challenges of the environment of the repository and effective deployment and continued operation of the robot system

autonomous monitoring↗
DOE OSTI2020

Autonomous Monitoring of Nuclear Repositories: Current State of the Art and Future Directions

Geological repositories for nuclear waste, including spent nuclear fuel, present a significant challenge for traditional International Atomic Energy Agency (IAEA) safeguards tools due to their inaccessibility and demanding operational conditions. The IAEA has been working closely with Member State organizations currently involved in repository construction and planning, including Euratom, the Finnish and Swedish regulatory authorities, and relevant facility operators. However, the verification challenge remains unsolved, and there persists an out- standing need for tools and approaches that will help the IAEA verify that no nuclear material is diverted from a repository environment. The challenge is also not static. Activities must encompass verification of the design, prior to and during the construction/operation phase, and post backfill. Throughout these various phases, it is imperative that the IAEA maintains a continuity of knowledge (CoK) of all material including information on material inventory and flow. This paper highlights these challenges and outlines how they might be addressed by using remote or autonomous vehicles. Specifically, we discuss the current state of the art in robotic autonomy for known or partially known environment mapping and patrolling, as well as shared autonomy, where humans collaborate with closed-loop autonomation to complete tasks. We explore the feasibility of using rovers for these verification tasks, along with the challenges associated with system implementation. Hardware and software suggestions are provided based on the adoption of similar technologies in other comparable areas and ability to close technical gaps. Lastly, human-robotic interactions are considered based on the challenges of the environment of the repository and effective deployment and continued operation of the robot system.

geologic repository, safeguards, IAEA, Robotics↗
DOE OSTI2018

Neutron Measurements from Boron Carbide Addition to a Plutonium Waste Form

The Department of Energy (DOE) has identified the Dilute and Dispose approach as the preferred method for the disposition of surplus plutonium materials. Dilute and Dispose requires surplus plutonium to be converted to an oxide product and then dry blended with a multi-component adulterant mixture to reduce the Safeguards and Security attractiveness level of the materials. The resulting product will be discarded in a geologic repository at the Waste Isolation Pilot Plant (WIPP).

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DOE OSTI2024

Value of abstraction in performance assessment – When is a higher level of detail necessary?

In this study, different approaches in performance assessment (PA) of the long-term safety of a repository for radioactive waste were examined. This investigation was carried out as part of the DECOVALEX-2023 project, an international collaborative effort for research and model comparison. One specific task of the DECOVALEX-2023 project was the Salt Performance Assessment Modelling task (Salt PA), which aimed at comparing various models and methods employed in the performance assessment of deep geological repositories in salt. In the context of the Salt PA task, three distinct teams from SNL (United States), Quintessa Ltd (United Kingdom), and GRS (Germany) examined the consequences of employing different levels of abstractions when modelling the repository's geometry and implementing various features and processes, using the example of a simple hypothetical repository structure in domal salt. Each team applied their own tools: PFLOTRAN (SNL), QPAC (Quintessa) and LOPOS (GRS). These differ essentially regarding numerical concept and degree of detail in the representation of the underlying physical processes. The discussion focused on when simplifications can be appropriately applied and what consequences result from them. Furthermore, it was explored when and if a higher level of fidelity in geometry or physical processes is required.

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DOE OSTI2019

Future Transition of WIPP from EM to Another DOE Office - 19256

The U. S. Department of Energy (DOE) Carlsbad Field Office (CBFO) has a primary mission to assist federal government entities disposition transuranic (TRU) radioactive waste. The main resource for disposition is the deep geologic repository known as the Waste Isolation Pilot Plant (WIPP). Other resources include a complete transportation system and a program that assists in the certification and shipment of TRU waste to WIPP known as the National TRU Program (NTP). All of these resources are managed by CBFO, which currently is part of the DOE Environmental Management (EM) program. EM's primary mission for DOE is to clean up the legacy remaining from many years of nuclear weapons research and production. But, CBFO estimates the need for WIPP and NTP will continue beyond the mission of EM and will assist other entities such as the National Nuclear Security Administration (NNSA), the Office of Science (OS) and the Navy. As the inventory of legacy TRU waste is reduced, the inventory of new TRU waste will continue from ongoing missions. So, over the life span of WIPP the primary customers of CBFO will shift from legacy EM sites to newly-generated TRU waste sites. And, somewhere along the way CBFO estimates that it will transition, primarily from an EM support (legacy waste) entity to a primarily a NNSA support (newly-generated waste) entity to facilitate ongoing DOE missions. This paper will address the need for that change at a programmatic level and will include some ideas for conceptual planning for such a transition. (author)

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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 OSTI2019

Why Cleanup Acceptance Is Running into Conflicts - Off-site Disposal Versus On Site Disposal - 19068

For years the U.S. Department of Energy (DOE) identified several sites to receive shipments of various types of wastes. However, with the backlog at the Waste Isolation Pilot Plant (WIPP), the inability to license Yucca Mountain, and the agency's desire to expedite cleanup, DOE is trying to 'dispose of' or 'store' more waste belonging to the Federal government on-site. Some states and communities that were 'promised' disposal of DoE's waste in geologic repositories and licensed landfills in arid locations, are now being told that on-site disposal is the best path forward. Competition for resources is forcing more sites to focus on near term cost savings of on-site disposal, rather than develop comprehensive life-cycle estimates representing a more realistic picture of costs associated with isolating radioactive and other hazardous materials on-site in perpetuity. The paper explores the issues and possible solutions in a case study on proposed on-site disposal of DOE waste, and concludes that DOE needs community support and community acceptance to successfully implement the agency's cleanup mission. Ongoing dialogue with the community saves both time and money but more importantly, cooperation helps ensure that effective long-term cleanup occurs. (author)

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