Development of Modern Performance Assessment Tools and Capabilities for Underground Disposal of Transuranic Waste at WIPP.
Abstract not provided.
Zeitler, Todd↗
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Field-scale Thermal Testing in a Generic Salt Disposal Environment Underground Research Laboratory (URL): Delineation of Principal Purpose Objectives and Hypotheses
The amount of brine present in domal salt formation is far less than in bedded salts (e.g., 0.01 to 0.1% compared with 1 to 3%). In salt domes, shear deformation associated with diapirism has caused existing brine to coalesce, leading to flow and expulsion. Brine migration behavior was investigated in bedded salt at WIPP (Nowak and McTigue 1987, SAND87-0880), and in domal salt at Asse (Coyle et al. 1987, BMI/ONWI-624). Test methods were not standardized, and the tests involved large diameter boreholes (17 to 36 in. diameter) and large apparatus. The tested intervals were proximal to mined openings (within approximately 1 diameter) where in situ stresses are redistributed due to excavation. The tests showed that (1) brine inflow rates can range over at least 2 orders of magnitude for domal vs. bedded salt, (2) that brine inflow is strongly associated with clay and interbedded permeable layers in bedded salt, and (3) that measurement systems can readily collect very small quantities of moisture over time frames of 2 years or longer. Brine inflow rates declined slightly with time in both test series, but neither series approached a state of apparent depletion. This range of flow magnitude could be significant to repository design and performance assessment, especially if inflow rates can be predicted using stratigraphic and geomechanical inputs, and can be shown to approach zero in a predictable manner.
58 GEOSCIENCES↗
Simulating the Effect of Fracture Connectivity on Repository Performance with GDSA Framework - 18589
This work examines the possible migration of radionuclides from a deep geologic nuclear waste repository sited in fractured crystalline host rock. The key safety concern is the potential effect on waste isolation of the inter-connectivity of the fracture network, which is primarily established by the temporal evolution of the temperature and stress fields at the time of original rock intrusion and cooling. Two end members are considered in these simulations, one with a high degree of connectivity to the biosphere, such that advective transport through the fracture and fault network controls radionuclide migration, and the other with a low degree of connectivity, such that slow diffusive transport through the crystalline rock matrix is the controlling process for radionuclide migration to the biosphere. Both end members have several areally extensive, high-transmissivity deformation zones, whose distance to the repository is an important factor for waste isolation capability. Uncertainties in fracture properties (transmissivity, orientation, radius) and fracture distribution also give rise to uncertainty in waste isolation capability (i.e., the overall connectivity of the repository with the surface biosphere). These types of natural system uncertainties are generally present in any geologic site-characterization program and thus represent an important factor in assessing repository performance in hard rock environments. The waste considered in these simulations are spent fuel rod assemblies from the U.S. commercial nuclear reactor fleet. Thermal output of the waste must be considered in the simulations and can influence the rate and timing of waste package failure, waste form degradation, and fluid flux due to thermal expansion around the repository horizon. To achieve a high-fidelity representation of radionuclide transport in fractures and rock matrix, combined with thermal energy transport and fluid flow in fractures and matrix, the mathematical model is solved numerically in a parallel high-performance computing (HPC) environment on a finite volume unstructured grid consisting of approximately 5 million cells, using Geologic Disposal Safety Assessment (GDSA) Framework (https://pa.sandia.gov), an open-source performance assessment tool for deep underground disposal of nuclear waste. GDSA Framework uses PFLOTRAN to solve the balance equations on a three-dimensional grid with heterogeneous properties, using multiple processors in a parallel configuration based on domain decomposition. The fracture networks in these simulations are originally generated as discrete fracture networks (DFNs), which are sets of two-dimensional planes distributed in a three-dimensional domain. The method used in GDSA Framework maps the stochastically generated DFN to an equivalent continuous porous medium (ECPM) domain that allows for the simulation of coupled heat flow, fluid flow, and radionuclide transport, including heat conduction through the matrix of the fractured rock, which is a process not easily modeled in a DFN representation. Computational efficiency is also greatly enhanced using the ECPM method, allowing for a realistic representation and analysis of uncertainties in a multi-realization performance assessment of a deep geologic repository. The effect of fracture connectivity on the waste isolation safety function, as brought to light by these GDSA Framework simulations, points to the importance of including a realistic representation of uncertainties in fracture properties and distribution (effectively, uncertainty in spatial heterogeneity) in repository safety assessment simulations. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Geochemical and microbial characteristics of seepage water and mineral precipitates in a radwaste disposal facility impacted by seawater intrusion and high alkalinity
The construction of an underground facility can dramatically change the quality, flow direction, and level of groundwater. It may also impact subsurface microbial composition and activity. Groundwater quality was monitored over eight years in two observational wells near an underground disposal facility on the east coast of South Korea. The results showed dramatic increases in dissolved ions such as O 2 , Na, Ca, Mg, and SO 4 during facility construction. Seepage water samples downgradient from the silos and tunnels, and precipitates deposited along the seepage water flow path were collected to determine the impact inside the disposal facility. X-ray analysis (powder X-ray diffraction (pXRD) and X-ray absorption fine structure (XAFS)) were used to characterize the mineral precipitates. Microbial community composition was determined by 16S rRNA gene sequencing. The seepage water composition was of two types: Ca-Cl and Ca-Na-HCO 3 . The ratio of Cl and δ 18 O showed that the Ca-Cl type seepage water was influenced by groundwater mixed with seawater ranging from 2.7% to 15.1%. Various sulfate-reducing bacteria were identified in the Ca-Cl type seepage water, exhibiting relatively high sulfate content from seawater intrusion. Samples from the Ca-Na-HCO 3 type seepage water had an extremely high pH (>10) and abundance of Hydrogenophaga. The precipitates observed along the flow path of the seepage water included calcite, ferrihydrite, green rust, and siderite, depending on seepage water chemistry and microbial activity. This study suggests that the construction of underground structures creates distinct, localized geochemical conditions (e.g., high alkalinity, high salinity, and oxic conditions), which may impact microbial communities. These biogeochemical changes may have undesirable large-scale impacts such as water pump clogging. Overall, an understanding of the process and long-term monitoring are essential to assess the safety of underground facilities.
54 ENVIRONMENTAL SCIENCES↗
Site Selection Criteria for Deep Borehole Disposal of Intermediate Level Waste in Australia - 19615
Disposal of long-lived intermediate-level waste (ILW) and long-lived spent sealed sources (SSS) requires a high degree of containment and isolation deep underground. Disposal in medium-depth (several hundreds of metres) boreholes in suitable hard rock formations can provide adequate isolation and containment for safe and cost-effective disposal of relatively small volumes of ILW and SSS. In this paper we outline recommended site selection criteria for the purpose of (i) identifying regions that would be excluded from further consideration, and subsequently for the non-excluded regions, (ii) to identify sub-regions that could be developed into a deep borehole disposal site. The CSIRO approach builds on siting guidelines previously developed by the Sandia National Laboratory which have been revised to suit the Australian context. (authors)
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Impact of a 70 °C temperature on an ordinary Portland cement paste/claystone interface: An in situ experiment
Radioactive wastes in future underground disposal sites will induce a temperature increase at the interface between the cementitious materials and the host rock. To understand the evolution of Portland cement in this environment, an in situ specific device was developed in the Underground Research Laboratory in Tournemire (France). OPC cement paste was put into contact with clayey rock under water-saturated conditions at 70 °C. The initial temperature increase led to ettringite dissolution and siliceous katoite precipitation, without monosulfoaluminate formation. After 1 year of interaction, partial decalcification and diffuse carbonation (calcite precipitation) was observed over 800 μm in the cement paste. At the interface, a layer constituted of phillipsite (zeolite), tobermorite (well-crystallised C-S-H) and C-(A)-S-H had formed. Globally, porosity decreased at both sides of the interface. Geochemical modelling supports the experimental results, especially the coexistence of tobermorite and phillipsite at 70 °C, minerals never observed before in concrete/clay interface experiments.
36 MATERIALS SCIENCE↗
WIPP PA Sensitivity Studies in Support of CRA-2014 Completeness Determination and Compliance Decision - 18209
The Waste Isolation Pilot Plant (WIPP) has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. Containment of TRU waste at the WIPP is regulated by the U.S. Environmental Protection Agency (EPA). The DOE demonstrates compliance with the containment requirements by means of performance assessment (PA) calculations. WIPP PA calculations estimate the probability and consequence of potential radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. Since TRU waste was first received at the WIPP in 1999, compliance recertification applications (CRAs) have been required by the EPA on five-year cycle. The latest CRA (CRA-2014) was submitted by the DOE to the EPA in 2014. As part of the application, a formal set of PA calculations (CRA-2014 PA) was performed and results were documented in detail. The CRA-2014 PA incorporated several parameter updates based on new data, as well as the implementation of a different model for panel closures, barriers planned for the opening of each waste panel. The results presented as part of the CRA-2014 showed via the required complementary cumulative distribution function (CCDF) release curves that the estimated releases were within compliance limits. As part of their process for reviewing the completeness of the CRA-2014 [1], the EPA formally requested that the DOE perform sensitivity studies based on EPA-requested changes to model parameters. The DOE performed and documented the results of three sensitivity studies described below: CRA14-SEN2 [2], CRA14-SEN3 [3], and CRA14-SEN4 [4]. In each case, the new analysis results were compared to CRA- 2014 PA (CRA14) results to determine the impact of the EPA-requested changes. CRA14-SEN2: the application of EPA-requested modified parameters in the operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model resulted in increased pressures and decreased brine saturations in waste areas and increased pressures and brine saturations in the operations and experimental areas. However, the overall effects on total high-probability mean releases from the repository were entirely insignificant. It was concluded that the modeling assumptions associated with the operations and experimental areas of the repository have an insignificant effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN3: the application of EPA-requested modified parameters in the panel closure areas and operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model predominantly result in increased pressures and decreased brine saturations in waste areas, increased pressures and brine saturations in the operations and experimental areas, and decreased brine and gas flows across all panel closures. The pressure increases in repository waste regions yielded slightly decreased brine saturations (on average) in those areas. However, the overall effects on total high-probability mean releases from the repository are entirely insignificant. It was concluded that the modeling assumptions made for this sensitivity study which were associated with the panel closures and north end areas of the repository have only a small effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN4: the application of several changes to the current WIPP PA model (CRA14), including: parameter changes, use of an updated code, and a revised computational grid for the BRAGFLO code. The modifications to the repository model resulted in increased releases in all primary release mechanisms. The impacts of each EPA-requested change to CRA14 were analyzed with respect to each release mechanism in the CRA14-SEN4 study. Overall, total high-probability predicted mean releases from the repository were increased by about 15%. Total low-probability predicted mean releases were increased by about 107%. It was concluded that the EPA-requested changes to the CRA14 result in increases to the predicted total releases from the repository. However, releases calculated in the CRA14-SEN4 analysis remained below regulatory limits, demonstrating continued compliance of the WIPP. In January 2017, the EPA determined that CRA-2014 was complete and then began a technical evaluation of CRA-2014 to determine whether the WIPP continued to comply with the radiation protection standards for disposal. During their evaluation, the EPA utilized the sensitivity study results to inform their ultimate determination in July 2017 that the WIPP continues to comply with regulatory standards. (authors)
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Leaching of Degraded Preservative Matrices, Based on Sodium Aluminophosphate Glasses, for High-Level Wastes
The chemical stability of vitrified high-level nuclear industry wastes in contact with water was studied experimentally in application to underground disposal of the wastes. The experiments were performed on samples of sodium aluminophosphate glass used in industrial solidification of wastes. Stable isotopes of strontium, cesium, uranium, and rare-earth elements as geochemical analogs of actinides were introduced into the glass to simulate radionuclides. It was observed that water vapor affects the degradation of the glass when vitrified wastes are placed in temporary repositories. It is shown that as a result of the degradation the rate of leaching of the glass in water increases significantly. In addition, a substantial portion of radionuclide simulators in colloidal form goes into the water medium.
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Impacts of an additional exhaust shaft on WIPP performance - 15170
The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. WIPP performance assessment (PA) demonstrates repository performance from facility closure to 10,000 years after closure. The recent radiological release event at the WIPP site has temporarily halted waste emplacement activities at the facility. A modified ventilation system is envisioned that will provide sufficient airflow necessary for the resumption of full-rate disposal operations in the future. A primary component of the modified ventilation system is an additional exhaust shaft in the north end of the repository. There are four shafts currently in the repository north end, namely a salt handling shaft, an exhaust shaft, a waste shaft, and an air intake shaft. These shafts are combined into a single composite shaft in WIPP PA that captures the combined impacts of all of them. The proposed additional exhaust shaft is combined with the four existing shafts in the repository model to determine its impacts on long-term repository performance. Moreover, mined volume in the repository north end is modified in the repository representation so as to include additional drifts created to access the new shaft. Computed gas and brine flow behaviors corresponding to the repository with the additional exhaust shaft are compared to those obtained in the PA executed for the 2014 Compliance Re-certification Application (the CRA-2014 PA). Results show that the repository with an additional exhaust shaft and drifts yields brine and gas flow behaviors similar to the current repository configuration. It is concluded that WIPP continues to satisfy regulatory compliance limits with the addition of an exhaust shaft and its access drifts, with compliance curves like those found in the CRA-2014 PA for total normalized releases. (authors)
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Basis of Knowledge for Oxidizing Chemicals in TRU Waste - 18654
A new Waste Isolation Pilot Plant (WIPP) document for implementing new requirements in the WIPP Waste Acceptance Criteria (WAC) titled Basis of Knowledge for Evaluating Oxidizing Chemicals in the TRU waste was issued in July, 2017 and has now been implemented on several waste streams. The February 14, 2014, airborne radiation release in Room 7 of the WIPP underground disposal unit 7 was caused by an exothermic chemical reaction of nitrate oxidizing chemicals and organic materials in a single 55-gallon container. That event exposed hazards not previously evaluated in the WIPP safety basis. A new WIPP Documented Safety Analysis (DSA), Revision 5b, was developed that addressed potential hazards related to the airborne radiation release and evaluated and addressed new fire hazard scenarios. The revised hazards analysis within the DSA specifically included potential chemical exothermic reactions in waste containers and propagating fires. Compliance with WAC was credited as initial conditions to the hazards analysis, and Chapter 18 was added making the WIPP WAC compliance program a safety management program within the DSA. The WIPP DSA, Revision 5b, Chapter 18, and the WIPP WAC require the use of this document when performing the enhanced acceptable knowledge (AK) process on transuranic (TRU) waste determined to contain oxidizing chemicals. The WIPP WAC was revised to address the conditions and requirements of DSA Revision 5b. The use of this Basis of Knowledge (BoK) by WIPP Certified Programs is specified in the WIPP WAC, Appendix H, Enhanced Acceptable Knowledge. This document establishes criteria that must be used by the WIPP Certified Programs to evaluate TRU waste containing one or more oxidizing chemicals to determine acceptability at the WIPP as-is and to identify when additional evaluation or treatment is required. This BoK also includes options and requirements to be used by TRU waste sites when further evaluation, testing, and/or treatment are required. The U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) will review requests, plans, and technical justifications and provide written decisions prior to implementation by the TRU waste sites. (authors)
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Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.
This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2009 of the performance of a repository for spent nuclear fuel and high - level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment - specific laboratory experiments, in - situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site - specific characterization . The current sixth period beyond 2010 represents a new effort to set waste management policy in the United States. Because the relationship is important to understanding the evolution of the Yucca Mountain Project , the tabulation also shows the interaction between the policy realm and technical realm using four broad categories of events : (a) Regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives, (c) technical milestones of implementing institutions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste. Preface The historical progression of technical milestones for the Yucca Mountain Project was originally developed for 10 journal articles in a special issue of Reliability Engineering System Safety on the performance assessment for the Yucca Mountain license application [1-10]. The listing of mile stones, a distinct feature of those articles, has been collected and tabulated here. A lthough a brief description is presented here (based on the summaries in the 10 journal articles), the emphasis remains on the tabulation because of its usefulness in providing a comprehensive but concise history of the Yucca Mountain Project. The tabulation presented here is more elaborate than originally presented in that many of the interactions that occurred between the technical realm and policy realm can be depicted in separate columns. The usefulness of the milestones table is due in part to L.A. Connolly, for editorial and reference support, and S.K. Best, Raytheon, and L. Mays, Sandia National Laboratories (SNL), for illustration support. Reviewers P.N. Swift, SNL, and K. Gupta, University of Oklahoma, helped improve the discussion. The historical perspective presented is that of the author and is not necessarily held by reviewers, Sandia National Laboratories , and the US Department of Energy. As a historic perspective, the author is reporting on the work of others; however, any interpretative error s of the documentation are those of the author alone. The characterization and modeling of the Yucca Mountain disposal system required numerous participants with expertise in many areas of science and technology, as evident from the extensive reference list. Their diligent efforts are generally acknowledged here and through the many references to their impressive work, but the 10 journal articles acknowledge by name many of the numerous participants that contributed to the Yucca Mountain Project .
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Andra's Underground Research Laboratory as a Tool to Prepare Monitoring System of Cigeo Project - 18070
Andra, as French radioactive waste management agency, plans to build an underground disposal facility, named Cigeo, dedicated to high and intermediate level, long lived radioactive wastes. To ensure the safety during the disposal operations that includes the principle of reversibility, the design of the associated monitoring system will be based on specific research and development program and feedback obtained over more than 15 years on tests performed in representative conditions. Andra's Underground Research Laboratory (URL), located at 500 m deep, is excavated in a clay-rich formation. It is dedicated to (i) assess the thermo-hydro-mechanical, chemical and radiologic behaviors of the rock and supporting layers (concrete and steel), (ii) test envisaged monitoring system, (iii) provide feedback on operational organization and safety management and (iv) qualify operating methods (digging, concrete pouring, monitoring system design, etc.) to be implemented in the future disposal. Data collected from (i) characteristics of samples retrieved from boreholes, (ii) experts' observations, (iii) sensors installed in situ and (iv) manual measurements, are integrated in a database. This database is dedicated to samples management (more than 73 600 classified samples), data measurements collection, and conversion to physical values and available to plots or exports. Each day, approximately 1.6 million of values are integrated in the database. Finally, a geo-referenced database is linked to the main database in order to allow the spatial localization, both on the surface or in the underground galleries. All these tools give confidence in Andra's capacity to monitor Cigeo and be ready for the future industrial pilot phase. The monitoring system will also allow, step by step, optimizing the disposal design, performing regular safety control, helping in decision making for reversibility and improving long term safety. (authors)
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Simulations of Criticality Control Overpack Container Compaction at the Waste Isolation Pilot Plant
Criticality Control Overpack (CCO) containers are being considered for the disposal of defense-related nuclear waste at the Waste Isolation Pilot Plant (WIPP). At WIPP, these containers would be placed in underground disposal rooms, which will naturally close and compact the containers closer to one another over several centuries. This report details simulations to predict the final container configuration as an input to nuclear criticality assessments. Each container was discretely modeled, including the plywood and stainless steel pipe inside the 55-gallon drum, in order to capture its complex mechanical behavior. Although these high-fidelity simulations were computationally intensive, several different material models were considered in an attempt to reasonably bound the horizontal and vertical compaction percentages. When exceptionally strong materials were used for the containers, the horizontal and vertical closure respectively stabilized at 43:9 % and 93:7 %. At the other extreme, when the containers completely degraded and the clay seams between the salt layers were glued, the horizontal and vertical closure reached respective final values of 48:6 % and 100 %.
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An integrated numerical model for coupled poro-hydro-mechanics and fracture propagation using embedded meshes
Integrated models for fluid-driven fracture propagation and general multiphase flow in porous media are valuable to the study and engineering of several systems, including hydraulic fracturing, underground disposal of waste, and geohazard mitigation across such applications. This work extends the coupled model multiphase flow and poromechanical model of Ren et al. (2018) to admit fracture propagation (FP). The coupled XFEM-EDFM scheme utilizes a separate fracture mesh that is embedded on a static background mesh. The onset and dynamics of fracture propagation are governed by the equivalent stress intensity factor (SIF) criterion. A domain-integral method (J integral) is applied to compute this information. Additionally, an adaptive time-marching scheme is proposed to rapidly restrict and grow temporal resolution to match the underlying time-scales. The proposed model is verified with analytical solutions, and shows the capability to accurately and adaptively co-simulate fluid transport and deformation as well as the propagation of multiple fractures.
42 ENGINEERING↗
Summary Report for the NEPA Impact Analysis. Revision 1
The Waste Isolation Pilot Plant (WIPP), located in southeastern New Mexico, has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of defense-related transuranic (TRU) waste. Containment of TRU waste at the WIPP facility is derived from standards set forth in Title 40 of the Code of Federal Regulations (CFR), Part 191. The DOE assesses compliance with the containment standards according to the Certification Criteria in Title 40 CFR Part 194 by means of Performance Assessment (PA) calculations performed by Sandia National Laboratories (SNL). WIPP PA calculations estimate the probability of radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. The DOE Carlsbad Field Office (CBFO) has initiated a National Environmental Policy Act (NEPA) action for a proposal to excavate and use additional transuranic (TRU) waste disposal panels at the WIPP facility. This report documents an analysis undertaken as part of an effort to evaluate the potential environmental consequences of the proposed action. Although not explicitly required for a NEPA analysis, evaluations of a dose indicator to hypothetical members of the public after final facility closure are presented in this report. The analysis is carried out in two stages: first, Performance Assessment (PA) calculations quantify the potential releases to the accessible environment over a 10,000-year post-closure period. Second, dose was evaluated for three hypothetical exposure pathways using the conservative radionuclide concentrations assumed to be released to the accessible environment.
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Development and Field Testing Novel Natural Gas Surface Process Equipment for Replacement of Water as Primary Hydraulic Fracturing Fluid (Final Report)
Southwest Research Institute® (SwRI ® ), Schlumberger Technology Corporation (SLB), and Chevron Corporation ® (Chevron) developed a novel, optimized, and lightweight process for natural gas (NG) to replace water as a low-cost fracturing medium with a low environmental impact. Hydraulic fracturing is used to increase oil and NG production by injecting high-pressure fluid, primarily water, into a rock formation, which fractures the rock and releases trapped oil and NG. This method was developed to increase yield and make feasible production areas that would not otherwise be viable for large-scale oil and NG extraction using traditional drilling technologies. Since the fracturing fluid is composed of approximately 90% water, one of the principal drawbacks to hydraulic fracturing is its excessive water use and associated large environmental footprint. According to recent data collected at fracture sites within the United States, fracturing applications in North America can consume as much as 9 million gallons of water per well. During the fracturing process, some of the fracturing fluid is permanently lost and the portion that is recovered is contaminated by both fracturing chemicals and dissolved solids from the formation. The recovered water or flow-back represents a significant environmental challenge, as it must be treated before it can be reintroduced into the natural water system. Although there is some recycling for future fracturing, the majority of the flow-back water is hauled from the well site to a treatment facility or to an injection well for permanent underground disposal. To mitigate these issues, an optimized, lightweight and modular surface process using NG to replace a majority of the water was developed as a cost-effective and environmentally clean fracturing fluid. Using NG will result in significantly less consumption since the gas that is injected as a fracturing fluid will be mixed with the formation gas and extracted as if it were from the formation itself. This process will minimize the collection, waste, and treatment of large amounts of water and will reduce the environmental impact of transporting and storing the fracturing fluid. There are two major steps involved in utilizing NG as the primary fracturing medium: (1) increasing the supply pressure of NG to wellhead pressures suitable for fracturing and (2) mixing the required chemicals and proppant needed for the fracturing process at these elevated pressures. The second step (NG-proppant mixing at elevated pressures) still requires demonstration; but very similar processes have been demonstrated in the field with other gases such as nitrogen (N 2 ) and carbon dioxide (CO 2 ). However, the first step (a compact, on-site unit for generating high-pressure NG at costs feasible for fracturing) has not been developed and is currently not commercially available. Due to the inherent compressibility of NG, more energy is required to compress the gas than what is required to pump water (or other incompressible liquids) to the very high pressure required for downhole injection. This project aimed to develop a novel, hybrid method to overcome this challenge. The project accomplishments and findings are discussed in this report. Ultimately, the research and development efforts described herein demonstrate that fracturing with NG foam is a feasible alternative to using water.
02 PETROLEUM↗
Compaction of crushed salt for safe containment – a summary of the KOMPASS projects
Abstract. For the underground disposal of high-level nuclear waste in rock salt formations, the safety concept includes the backfilling of open cavities with crushed salt. For the prognosis of the sealing function of the backfill for the safe containment of the radioactive waste, it is crucial to have a comprehensive process understanding of the crushed-salt compaction behavior. The crushed-salt compaction process is influenced by internal properties (e.g., grain size, mineralogy, and moisture content) and boundary conditions (e.g., temperature, stress state, and compaction rate) and, therefore, involves several coupled thermal–hydro–mechanical (THM) processes (Hansen et al., 2014; Kröhn et al., 2017). With the paradigm shift from the limited release of radionuclides to safe containment due to the German Repository Site Selection Act passed in 2017, the importance of crushed salt as geotechnical barrier has increased, with a focus on the evolution of its hydraulic properties. Based on the knowledge gaps in the current process understanding, the “Compaction of crushed salt for safe containment” (KOMPASS) projects were initiated to improve the scientific basis behind using crushed salt for the long-term isolation of high-level nuclear waste within rock salt repositories. The efforts to improve the prediction of crushed-salt compaction begun during the first phase of the KOMPASS projects (Czaikowski et al., 2020) and were followed up in a second phase ending in June 2023. The primary achievements of the projects are as follows (Czaikowski et al., 2020; Friedenberg et al., 2022): specification of the KOMPASS reference material, an easily available and reproducible synthetic crushed-salt material, for generic investigations; development of pre-compaction methods and successful production of samples in the short term and under in situ loading conditions; formulation of an extended laboratory program addressing the isolated investigation of known relevant factors influencing the compaction behavior of crushed salt (Düsterloh et al., 2022); execution of long-term compaction tests addressing isotropic and deviatoric load changes, temperature, and compaction state; construction of a backfill body using the KOMPASS reference material in the Sondershausen mine through collaboration with the SAVER (Entwicklung eines salzgrusbasierten Versatzkonzepts unter der Option Rückholbarkeit) project (Schaarschmidt and Friedenberg, 2022); advancement of the tools for microstructure investigation methods (Svensson and Laurich, 2022); generation (first stages) of a microphysical process list combining literature research with our own findings; benchmarking of long-term compaction test for model development and optimization of various existing models as well as the development of new models; application of a virtual demonstrator (2D model representing a backfilled drift in rock salt) for the visualization of developments and the quantification of the models (Rabbel, 2022). In summary, the KOMPASS projects contributed to the reduction of uncertainties and the strengthening of the safety case for using crushed salt within rock salt repositories.
Friedenberg, Larissa↗
French Cigeo Operational Safety Options: The General Approach and Challenging Issues Towards the Licensing - 18051
After around 25 years of In view of the licensing application, two main milestones for safety are identified for Cigeo: a Safety Options Dossier 'DOS' early 2016 and the safety case to support Future License Application of Cigeo. The Safety Options was an opportunity for Andra to send in advance a first safety case to the French Safety Authority in order to stabilize the safety strategy, the safety requirements, the safety methods, key safety and design options, the list of safety scenarios that are selected and a preliminary 'bounding' impact. The paper describes the approach applied for the operating safety options and the main concluding aspects as well as the challenging issues towards the licensing. (authors)
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