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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Conditioning and Requalification of Legacy Waste Packages - 18648

In most German interim storage facilities, hundreds of waste packages are waiting for re-conditioning and qualification to comply with the waste acceptance criteria (WAC) for the Konrad repository. Especially the waste classification as well as some characteristics and the description of the material composition in its current state do not comply with the Konrad requirements since the conditioning of these waste packages took place up to 40 years ago. The German name for these waste packages is 'Altgebinde', hereafter referred to as 'legacy waste packages'. The present article will provide an overview of the measures for the qualification of these legacy waste packages to fulfill the Konrad requirements. The following measures, when necessary, are carried out for the qualification of legacy waste packages: - Collection and evaluation of all documents related to the primary conditioning process, e. g. the used equipment, documented waste and packaging properties, conditioning and radiology, if available (hereafter 'old documents'). - Preparation of a conceptual report and a process control quality plan based on existing data, with appropriate measures to qualify the waste packages. Assessment and approval of the process control quality plan by the competent authority. - Characterization of the waste and packaging properties by visual inspection or tomography, recording or updating of dose rate, contamination and mass, verification of product stability, for example by gas analysis of the interior atmosphere of the container. - Characterization of the nuclide inventory by gammas scans, sample taking to analyze nuclides which can not be measured by means of in-situ gamma spectrometry, and for material declaration. - Qualification or post-conditioning of the legacy waste packages if necessary, e. g. by drying the waste packages and packaging them in containers approved for disposal. - Compilation of all results of the above mentioned tests and steps to create a waste package documentation, including the calculations for the nuclide inventory and the description of the material composition of the waste packages. - Receiving approval by the competent authority after evaluation of the final documentation. The aim behind this structured approach is to prevent additional, unnecessary measures and treatment of the waste. Consequently, the additional characterization is preferably carried out on the basis of the available documents. (authors)

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DOE OSTI2017

Practical Considerations for the Implementation of a World Information Library (WIL) - 17112

The U.S. Department of Energy's (DOE) Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for the permanent disposal of defense-generated transuranic waste. Title 40 Code of Federal Regulations (CFR) Part 194.43, Passive institutional controls, requires that measures be employed, '...to preserve knowledge about the location, design, and contents of the [WIPP] disposal system'. These measures include the, 'Placement of records in the archives and land record systems of local, State, and Federal governments, as well as international archives, that would likely be consulted by individuals in search of unexploited resources'. One possibility for allowing individuals to quickly and easily access information on the location, design, and contents of a disposal system and its relation to unexploited resources is the development of an electronic geographic information system (GIS) that would be used to interpret spatial data. In 2016, R. Patterson proposed such a system, known as the 'World Information Library (WIL)'. Information in this database system would be designed to be publicly accessible through a simple web-based user interface that would focus on spatial databases with categories such as; waste repositories, resources, historical sites, wildlife, infrastructure, medicine, agriculture, and technology. The conceptual WIL would be a compilation of searchable libraries and databases. A category in the WIL would consist of spatial databases created for the purpose of preserving knowledge and providing a means of communicating location information in a GIS-based Land Use Information System (LUIS) user interface. The successful implementation of the WIL would require the robust use of project controls, an effective Quality Assurance program, and the development of international standards in collaboration with industry experts. The purpose of this paper is to identify and discuss known practices and industry standards that could be utilized for the implementation of a WIL as was introduced by Patterson et al. in 'A Conceptual World Information Library (WIL) and Land Use Information System (LUIS)'. (authors)

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DOE OSTI2020

Summary of Information Available and Recommendations for a Doe Complex-Wide Waste Glass Database

The purpose of this document is to summarize the existing repositories of nuclear waste glass property-composition data produced for DOE. A framework is described for consolidation of the data into a single database. Recommended attributes of the consolidated database are provided, including quality assurance, the user interface, and means for updates and maintenance. Future work will utilize the framework described herein to produce the complex-wide database and establish the user interface.

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DOE OSTI2018

Qualification of Numerical and Semi-Analytical Modeling Codes for Thermal Design of HLW Repository - 18455

Disposal in a geological repository is a viable option for safely isolating high-level radioactive waste. Thermal processes induced by decay heat from the radioactive waste may impact the integrity of the engineered barrier system, and may lead to fluid flow and transport; and the associated migration of radionuclides. This study looks at an analysis of thermal conduction for the emplacement of high-level waste in a repository located in bedded salt. The simulations used analytical and numerical methods to study conductive heat transport in a repository. This study concentrated on heat flow in a central waste package with contributions from adjacent waste packages and drifts. Of particular importance to this work is comparison of predictions of peak temperature at the drift wall and at the waste package surface using different solution methods. Simulation results of analytical and numerical modeling methods are compared. The results show that predictions of the codes were comparable under identical initial and boundary conditions. (authors)

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DOE OSTI2018

Coupled Processes Modeling in Rock Salt and Crushed Salt Including Halite Solubility Constraints: Application to Disposal of Heat-Generating Nuclear Waste

This work presents numerical modeling of coupled thermal, hydraulic and mechanical processes in rock salt and crushed salt considering halite solubility constraints. The TOUGH-FLAC simulator is used, with a recently enhanced Equation-Of-State module that includes the thermodynamic properties of aqueous fluids of variable salinity. Laboratory and field scale tests performed on rock salt and crushed salt under temperature gradients are modeled first to evaluate the capabilities of the simulator to reproduce important features, such as porosity changes induced by halite dissolution/precipitation, and brine and heat migration. Since the results are quite satisfactory, the simulator is used to predict the long-term response of a generic salt repository for heat-generating nuclear waste. To evaluate the impacts of halite solubility on the predictions, two simulations that respectively consider or neglect solubility constraints are performed. In the scenario studied, the results are not significantly affected by dissolution/precipitation, and only some differences are observed due to changes in porosity, but the dominating processes remain the same. Finally, with the new provisions, TOUGH-FLAC is more complete in terms of processes occurring around a heat-releasing nuclear waste package and can therefore provide more accurate predictions of the long-term performance of a nuclear waste repository in salt formations.

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DOE OSTI2024

KOMPASS-II: Compaction of Crushed salt for Safe Containment – Phase 2

Long-term stable sealing elements are a basic component in the safety concept for a possible repository for heat-emitting radioactive waste in rock salt. The sealing elements will be part of the closure concept for drifts and shafts. They will be made from a welldefinied crushed salt in employ a specific manufacturing process. The use of crushed salt as geotechnical barrier as required by the German Site Selection Act from 2017 /STA 17/ represents a paradigm change in the safety function of crushed salt, since this material was formerly only considered as stabilizing backfill for the host rock. The demonstration of the long-term stability and impermeability of crushed salt is crucial for its use as a geotechnical barrier. The KOMPASS-II project, is a follow-up of the KOMPASS-I project and continues the work with focus on improving the understanding of the thermal-hydraulic-mechanical (THM) coupled processes in crushed salt compaction with the objective to enhance the scientific competence for using crushed salt for the long-term isolation of high-level nuclear waste within rock salt repositories. The project strives for an adequate characterization of the compaction process and the essential influencing parameters, as well as a robust and reliable long-term prognosis using validated constitutive models. For this purpose, experimental studies on long-term compaction tests are combined with microstructural investigations and numerical modeling. The long-term compaction tests in this project focused on the effect of mean stress, deviatoric stress and temperature on the compaction behavior of crushed salt. A laboratory benchmark was performed identifying a variability in compaction behavior. Microstructural investigations were executed with the objective to characterize the influence of pre-compaction procedure, humidity content and grain size/grain size distribution on the overall compaction process of crushed salt with respect to the deformation mechanisms. The created database was used for benchmark calculations aiming for improvement and optimization of a large number of constitutive models available for crushed salt. The models were calibrated, and the improvement process was made visible applying the virtual demonstrator.

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DOE OSTI2017

Thermal Analysis of Disposal of High-Level Nuclear Waste in a Generic Bedded Salt repository using the Semi-Analytical Method.

An example case is presented for testing analytical thermal models. The example case represents thermal analysis of a generic repository in bedded salt at 500 m depth. The analysis is part of the study reported in Matteo et al. (2016). Ambient average ground surface temperature of 15°C, and a natural geothermal gradient of 25°C/km, were assumed to calculate temperature at the near field. For generic salt repository concept crushed salt backfill is assumed. For the semi-analytical analysis crushed salt thermal conductivity of 0.57 W/m-K was used. With time the crushed salt is expected to consolidate into intact salt. In this study a backfill thermal conductivity of 3.2 W/m-K (same as intact) is used for sensitivity analysis. Decay heat data for SRS glass is given in Table 1. The rest of the parameter values are shown below. Results of peak temperatures at the waste package surface are given in Table 2.

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DOE OSTI2018

Advanced Experimental Technique for Radiation Damage Effects in Nuclear Waste Forms: Neutron Total Scattering Analysis

For the past 30 years, the development of durable materials for radionuclide immobilization has been driven by efforts to dispose of wastes generated by the nuclear fuel cycle [National Research Council, ‘Waste Forms Technology and Performance: Final Report’, the National Academies Press, Washington D.C., 2011]. Many materials have been developed, but there still exist large gaps in the knowledge of fundamental modes of waste form degradation in repository environments. An important aspect of waste form science is the behavior of the materials under intense irradiation from decaying actinides and fission products. This irradiation induces a wide range of defects and disorder, the details of which depend on the specific waste form material. At the present time, it is not fully explained how radiation effects will influence the performance of nuclear waste forms and their long-term retention of fission products and actinides under operational conditions. The complex defect behavior and radiation damage must be understood over a range of length scales, from the initial atomic-scale defect structure to the long-range observable material modification. This is particularly challenging and requires advanced characterization techniques. This contribution describes how pair distribution function (PDF) analysis obtained from neutron total scattering experiments can be applied in the research field of waste form science to uniquely characterize radiation effects in a wide range of materials, including crystalline complex oxides and waste glasses. Neutron scattering strength does not have an explicit Z-dependence; this allows access to many low-Z elements, such as oxygen, that cannot be accurately studied with X-rays. In many cases, this can permit a detailed analysis of both cation (often high-Z) and anion (often low-Z) defect behavior. In contrast to traditional crystallography, which relies on long-range order, PDF analysis probes the local defect structure, including changes in site occupation, coordination, and bond distance. This is particularly important when characterizing aperiodic waste glasses with no long-range order at all. In contrast to X-ray characterization which requires very little sample mass (~0.1 mg), neutron characterization (even at state-of-the-art spallation facilities) requires relatively large sample mass (~50 - 100 mg). In conclusion, obtaining this quantity is challenging for studies of irradiated materials, but by tailoring our experimental approach to use high-energy ions (GeV) with very high penetration depth, we are able to produce the required mass.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Broad Scope Analysis of the MOX Fuel Fabrication Facility and Disposition of Plutonium

The Plutonium Management and Disposition Agreement (PMDA), signed into effect in 2000, between the United States and Russia, was made in a post-Cold War effort to dispose of military-useable material, a combined 68 metric tons of weapon-grade plutonium. The United States has been pursuing a disposition path through conversion to MOX fuel for use in current commercial nuclear reactors. The Savannah River Site’s (SRS) MOX Fuel Fabrication Facility (MFFF, the MOX facility) began construction in 2007 with intent to uphold the United States’ half of the agreement. In the FY2017 budget, the Obama administration proposed a termination of the MFFF after lack of sufficient funding to the project, which caused paralyzing delays and drastically inflated costs for completion. Prolonged construction, piled costs, and divergent policies between shifting administrations have resulted in turmoil between the United States and Russia, culminating with the proposed abandonment of the MOX disposition path and being used as a justification by Russia to withdraw from the agreement until the United States is able to identify a clear plan, fulfilling the requirements set forth by the PMDA. The leading expedient to replace MOX fuel is the dilute and dispose method, which leaves uncertainty for the SRS MFFF and places reliance on New Mexico’s Waste Isolation Pilot Plant (WIPP) to legally accommodate the additional and unaccounted waste. If the current capacity of WIPP proves insufficient, construction of a new repository for transuranic waste would only incur additional costs and regulatory burdens, along with extreme schedule delays. Future proposed adaptations of the SRS MFFF include pit production, which would require complex and costly alterations to be operational at the current site. Restrictions from Congress, such as mandating the proposed solution to abandoning MFFF must equal half of the costs required for completing the facility, have fueled severely underestimated quotations of these alternatives. The increasing pressure from both Russia and South Carolina government officials to make a decision could result in the pursuit of another doomed project, leaving the United States with even further sunk costs and strained international relations. This paper will summarize the involved history of the MFFF, explicate the recourses to salvage the facility for pit production, and provide postulation on the dilute and dispose method.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Investigation of Coupled Processes and Impact of High Temperature Limits in Argillite Rock: FY17 Progress. Predecisional Draft

The focus of research within the Spent Fuel and Waste Science and Technology (SFWST) (formerly called Used Fuel Disposal) Campaign is on repository-induced interactions that may affect the key safety characteristics of EBS bentonite and an argillaceous rock. These include thermal-hydrologicalmechanical- chemical (THMC) process interactions that occur as a result of repository construction and waste emplacement. Some of the key questions addressed in this report include the development of fracturing in the excavation damaged zone (EDZ) and THMC effects on the near-field argillaceous rock and buffer materials and petrophysical characteristics, particularly the impacts of temperature rise caused by waste heat. This report documents the following research activities. Section 2 presents THM model developments and validation, including modeling of underground heater experiments at Mont Terri and Bure underground research laboratories (URLs). The heater experiments modeled are the Mont Terri FE (Full-scale Emplacement) Experiment, conducted as part of the Mont Terri Project, and the TED in heater test conducted in Callovo-Oxfordian claystone (COx) at the Meuse/Haute-Marne (MHM) underground research laboratory in France. The modeling of the TED heater test is one of the Tasks of the DEvelopment of COupled Models and their VAlidation against EXperiments (DECOVALEX)-2019 project. Section 3 presents the development and application of thermal-hydrological-mechanical-chemical (THMC) modeling to evaluate EBS bentonite and argillite rock responses under different temperatures (100 °C and 200 °C). Model results are presented to help to understand the impact of high temperatures on the properties and behavior of bentonite and argillite rock. Eventually the process model will support a robust GDSA model for repository performance assessments. Section 4 presents coupled THMC modeling for an in situ test conducted at Grimsel underground laboratory in Switzerland in the Full-Scale Engineered Barrier Experiment Dismantling Project (FEBEX-DP). The data collected in the test after almost two decades of heating and two dismantling events provide a unique opportunity of validating coupled THMC models and enhancing our understanding of coupled THMC process in EBS bentonite. Section 5 presents a planned large in-situ test, “HotBENT,” at Grimsel Test Site, Switzerland. In this test, bentonite backfilled EBS in granite will be heated up to 200 °C, where the most relevant features of future emplacement conditions can be adequately reproduced. Lawrence Berkeley National Laboratory (LBNL) has very actively participated in the project since the very beginning and have conducted scoping calculations in FY17 to facilitate the final design of the experiment. Section 6 presents present LBNL’s activities for modeling gas migration in clay related to Task A of the international DECOVALEX-2019 project. This is an international collaborative activity in which DOE and LBNL gain access to unique laboratory and field data of gas migration that are studied with numerical modeling to better understand the processes, to improve numerical models that could eventually be applied in the performance assessment for nuclear waste disposal in clay host rocks and bentonite backfill. Section 7 summarizes the main research accomplishments for FY17 and proposes future work activities.

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DOE OSTI2019

Fault Stability Perturbation by Thermal Pressurization and Stress Transfer Around a Deep Geological Repository in a Clay Formation

The increase of temperature of a low-permeability, fluid-saturated media may trigger significant thermal pressurization, where the expansion of pore fluid cannot be accommodated by the thermal expansion of the pore space. With the aid of a coupled thermohydromechanical numerical simulator, we investigate the possible impact of thermal pressurization during the life of a deep geological repository (DGR) for high-level radioactive waste, characterized by the emplacement of radioactive material-filled canisters in a series of parallel tunnels, excavated in a low-permeability clay formation. We represent the fault as a planar structure embedded in a thermoporoelastic material and shear activation evaluated by a strain-softening Mohr-Coulomb failure criterion. The results show that stress changes caused by temperature and thermal pressurization of a rock mass around the emplacement tunnels may trigger a slip event on a fault plane in proximity of the geological disposal site: Rupture nucleates at depth, hundreds of meters below the DGR. Stress transfer plays a key role, while a direct hydraulic connection between the repository and the nucleation zone is not necessary in order to trigger rupture. A low stress ratio may favor the occurrence of slip up to a distance of 600 m of the fault from the outermost tunnel. Furthermore, these results highlight the need of investigating hydromechanical properties and local stress conditions at depth to characterize the geomechanical response of weak planes located in the surroundings of a high-level radioactive waste repository and to provide sufficient knowledge for the safe development of the DGR site.

58 GEOSCIENCES↗
DOE OSTI2020

Quest for Continuous Design and Safety Optimization: A Look at Cires, the French Very Low Level Waste Disposal - 20010

Cires, the French repository for very low-level waste (VLLW) managed by Andra, has a fifteen-year history of enhancing design and disposal operations, responding to producers' needs by optimizing existing facilities and providing new facilities as needed. Its flexibility allows the facility to adapt to changes in waste delivery flows, to dispose of larger waste volumes without using more space, to integrate new concepts and even envisage disposal capacity extension. The article traces the ever evolving history of technical and operational optimization of this unique disposal facility situated in Eastern France. (authors)

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DOE OSTI2017

Waste Minimization of HyBRID Decontamination Process - 17060

The Korea Atomic Energy Research Institute (KAERI) developed a Hydrazine base reductive metal ion decontamination (HyBRID) and related pre-oxidation techniques for use in decommissioning a nuclear facility. Inorganic chemicals were used to increase the integrity of the radioactive waste in a repository. Decomposition and precipitation of residual reagents with inorganic chemicals were introduced to minimize the final waste volume. The waste volume was calculated theoretically and demonstrated in a small-scale waste treatment system and bench-scale candle filtration system. The total waste volume of the HyBRID process was less than 1/10 of that of commercial processes, and spent ion-exchange resins were significantly reduced to 5% of that of commercial processes. (authors)

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DOE OSTI2017

Advanced Fuel Cycle Cost Basis – 2017 Edition

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Enthalpies of formation of polyhalite: A mineral relevant to salt repository

Polyhalite is an important coexisting mineral with halite in salt repositories for nuclear waste disposal, such as Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. The thermal stability of this mineral is a key knowledge in evaluating the integrity of a salt repository in the long term, as water may release due to thermal decomposition of polyhalite. Previous studies on structural evolution of polyhalite at elevated temperatures laid the basis for detailed calorimetric measurements. Using high-temperature oxide-melt drop-solution calorimetry at 975 K with sodium molybdate as the solvent, we have determined the standard enthalpies of formation from constituent sulfates (ΔH° f,sul ), oxides (ΔH° f,ox ) and elements (ΔH° f,ele ) of a polyhalite sample with the composition of K 2 Ca 2 Mg(SO 4 ) 4 ·1.95H 2 O from the Salado formation at the WIPP site. The obtained results are: ΔH° f,sul = -152.5 ± 5.3 kJ/mol, ΔH° f,ox = -1926.1 ± 10.5 kJ/mol, and ΔH° f,ele = -6301.2 ± 9.9 kJ/mol. Furthermore, based on the estimated formation entropies of polyhalite, its standard Gibbs free energy of formation has been derived to be in the range of -5715.3 ± 9.9 kJ/mol to -5739.3 ± 9.9 kJ/mol. In conclusion, these determined thermodynamic properties provide fundamental parameters for modeling the stability behavior of polyhalite in salt repositories.

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DOE OSTI2018

Buffer Material Degradation in High Level Radioactive Waste Disposal - 18454

The concept of geological disposal for high-level radioactive waste (HLW) is based on a multi-barrier system comprising the natural geological barrier that repository host rock provides and its surroundings, as well as an engineered barrier system (EBS) that includes waste form, waste canisters, buffer materials, and backfill. The multi-barrier system is expected to perform its desired functions to isolate waste from the biosphere and achieve HLW disposal. Because of the interaction processes of radionuclide decay heat with hydraulic and water-mineral reactions in the radioactive waste repository, the bentonite buffer degradation resulting from the external temperature and pressure warrants investigation. However, seldom addressed in Taiwan is the influence of thermo-hydro-chemical (T-H-C) processes on buffer material degradation in the EBS of HLW disposal repositories as related to smectite clay dehydration. Therefore, we adopted the chemical kinetic model of smectite dehydration to calculate the amount of water expelled from smectite clay minerals caused by higher temperatures of waste decay heat. We discovered that smectite clay could cause volume shrinkage because of a loss of interlayer water, which can lead to bentonite buffer compression. The results of this study may be used in advanced research on the evolution of bentonite degradation for both performance assessments and safety analyses of final HLW disposal in Taiwan. (authors)

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