Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Waste Repository”

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

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

At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

CEA decommissioning and clean-up activities legacy waste management and lessons learned - 15506

The CEA has generated radioactive waste in both liquid and solid form for more than 50 years within the scope of its R and D, clean-up and dismantling programs. The vast majority of the waste produced is transferred to disposal centers in approved waste packages. Nonetheless, a large quantity of legacy waste still remains in ageing facilities in which the storage conditions are sometimes no longer satisfactory with respect to the safety requirements that are constantly evolving. As part of a broad clean-up and dismantling process focusing on these ageing facilities managed by the CEA, the top priority is to shut down nuclear activities on two symbolic sites: Grenoble and Fontenay-aux-Roses, which are both located in a dense urban environment. After about ten years of dismantling, the nuclear facilities on the Grenoble site are on their way to be completely decommissioned in the regulatory sense of the term. Furthermore, the radwaste issue on this site has raised no specific problems, though it did require developing innovative and sometimes costly waste outlets. The older Fontenay-aux-Roses site is where the first experimental research on minor actinide separation took place. The facilities on this site have reached a stage in dismantling which necessitates the recovery of irradiating legacy waste containing alpha, beta and gamma emitters. The transfer of this waste to interim storage centers more than 600 kilometers away is a decisive factor in reducing the site's nuclear source term and in ensuring the satisfactory progress of the dismantling operations. This paper details the issues raised by the recovery of legacy waste, which requires greater characterization to better understand its radiological and chemical status. This waste also requires better characterization to meet the storage requirements of new sites designed for storage periods exceeding 50 years prior to its final disposal in a deep geological repository. The issue of transporting waste on public roads is specifically covered, along with the strategy defined to securely transfer all legacy waste resulting from the CEA's very diverse R and D programs. Emphasis is placed on finding the right balance between collecting the necessary data on the waste to be transported, ensuring the reconditioning of this waste so it can be transported safely, and maintaining an efficient industrial-scale waste transfer process to limit the lead times and costs related to legacy waste recovery operations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Post-closure Nuclear Criticality Safety Evaluations for Disposition of Criticality Control Overpacks at the Waste Isolation Pilot Plant

The Waste Isolation Pilot Plant (WIPP) is a geological repository in southern New Mexico that provides for disposal of transuranic (TRU) wastes from atomic energy defense activities. The Sandia National Laboratories (Sandia) Report, Consideration of Nuclear Criticality When Disposing of Transuranic Waste at the Waste Isolation Pilot Plant, addresses nuclear criticality safety based on the projected inventory characteristics for the initial compliance certification application of WIPP in 1996. As the inventory, waste forms, and disposal package designs change, revised or new analyses are necessary to demonstrate acceptability for these configurations within the WIPP safety basis and compliance with 10,000-year post-closure standards of the US Environmental Protection Agency (EPA). Saylor and Scaglione evaluated criticality control overpacks (CCOs) in 2017 based on conservative assumptions for post-closure repository structural conditions with resulting effects on containers and container spacing, The Saylor and Scaglione evaluation of CCOs addressed a single waste configuration that represents the Surplus Plutonium Disposition Program’s dilute and dispose waste form and composition. This initial CCO study demonstrated that 50 grams of boron carbide (B 4 C) per CCO is sufficient to ensure post-closure criticality safety based on a well-mixed waste composition, and Oak Ridge National Laboratory (ORNL) subsequently determined that this amount of B 4 C does not require constraints on moisture or plastic present as moderator. The Saylor and Scaglione analysis conservatively assumes repository room closure that eliminates all space between fissile gram equivalent (FGE) 239 Pu masses. The close-packed array was selected based on limited availability of repository salt creep modeling results at that time. In 2019, Brickner provided additional evaluations for pipe overpack containers (POCs), building on the conservative basis provided by Saylor and Scaglione. Brickner’s 2019 analysis made use of new geomechanical data for post-closure spacing that rely on advances in repository modeling as documented in the work by Reedlunn and Bean. This current CCO evaluation for generic waste materials expands on earlier work performed at ORNL and includes evaluation of CCOs across a much broader range of possible waste compositions and geometries. This evaluation is intended to provide input for the required feature, event and process (FEP) screening to determine if post-closure criticality must be included as an event in the 10,000-year regulatory evaluation. As such, the approach to modeling post-closure criticality presented in this report has been coordinated with the Sandia team responsible for FEP screening. The resulting analysis supports disposition of fissile materials in the CCO containing up to 380 FGE 239 Pu and expands conditions acceptable for disposal of fissile material in CCOs. This evaluation builds on the methodology of Saylor and Scaglione and Brickner, using the most recently available geomechanical data for CCO spacing under salt creep compaction scenarios provided by Reedlunn and Bean. The broad range of fissile material configurations analyzed in this report are intended to account for configurations that may occur during the post-closure disposal time period, and it also includes waste configurations that are not physically possible to support analysis of conditions that influence neutron fluence.

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

International Collaboration Activities on Engineered Barrier Systems

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Why accredit ISO/CEI 17025 a nuclear waste management laboratory? - 15112

When there is no storage, nor repository, available on a site, the entity in charge of nuclear waste management should find outlets in order to evacuate outside the site. Each outlet has his rules and waste acceptance criteria (packaging rules, physical/chemical/radiological criteria for example). Before you are able to evacuate your waste, an acceptance's process (or contract review) takes place. This process scans the entire measures implemented by the site in order to satisfy outlet's criteria. When the acceptance is obtained, shipments could be organized, and the process gets into an observation's phase. During this phase, the outlet could choose to inspect a finished nuclear waste package or to audit regularly the waste management process. In this last case, the outlet re-scans the entire documents and implemented measures. This audit lasts from one to several days. A way to tackle/approach the audit is to acknowledge, by a trusted third party (i.e. independent from the outlet), all or part of waste management system, in order to base one's reputation on this organism and win outlet's trust. For this purpose, the entity could choose certification or accreditation. Certification (ISO 9001 as often as not) aims at evaluating a system for managing the quality of one product or service. Accreditation (ISO 17025, for example) provides formal recognition for technical skills within a quality management system (clauses 5 and 4 of the ISO 17025 standard). In 2006, the LDM laboratory is the first French laboratory accredited, in accordance with ISO 17025 standard, for the characterization by gamma spectrometry on 100 L- and 200 L- radioactive waste drums (accreditation No 1-1712 - scope available on www.cofrac.fr). This accreditation allows the lab: - to update the quality management system, - to compile the entire technical file of the gamma-ray assay (description, evaluation of the uncertainties, calibration, proficiency testing programs, measurement procedures,...), - to pass previous identified constraints on to producers and nuclear waste package production, - to maintain permanent quality improvement loop, - to lighten the programs of outlets audits, through the acknowledgement of the technical competence of the laboratory to evaluate by gamma spectrometry the activity contained in our nuclear waste drums. On the other hand, the lab is re-evaluated periodically to ensure its continued compliance with requirements: within a 12 month's interval for an internal audit, with a 18 month's period for COFRAC (French sole official national accreditation body). Since 2006, the lab is accredited. His accreditation renewed in 2010. A new entire renewal audit is planned in 2015. (authors)

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

Transport of French PWR steam generators to very low level waste disposal - 15203

When considering the complexity of IAEA nuclear transport regulations, the transport of a Steam Generator (SG) in one un-packaged block from a dismantling plant to a nuclear waste disposal is a challenge. The choices made, the difficulties encountered and the feedback that has been learnt from it are described. The disposal of decommissioning waste requires taking into account not only the constraints of acceptance of the repository but also the specific constraints of transportation from the dismantling plant to the final disposal site. In France, these criteria and regulations are respectively under the supervision of the French National Radioactive Waste Management Agency (ANDRA) and the French Nuclear Safety Authority (ASN). In the course of the dismantling of the EDF PWR plant Chooz A, the four decontaminated SGs were shipped to the ANDRA Very Low Level Waste (VLLW) repository. After removal, the SGs were decontaminated, allowing them to attain the criteria of VLLW and thus be accepted by the ANDRA. However, to transport these decontaminated SGs as single un-packaged blocks, it was necessary to ensure compliance with the Surface Contaminated Object SCO-I regulation criteria. For this, transport safety files were written and surface contamination and radiation criteria checked. One of the four SGs had to be decontaminated again and another one had to be transported under special arrangement. Due to the exceptional nature of the transportation, specialized means of transport and stowage were also studied. (authors)

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

DECOVALEX-2023: Task E Final Report

This is the Task E final report for DECOVALEX-2023. Task E is focused on understanding thermal, two-phase hydrological, and mechanical (TH2M) processes, especially related to predicting brine migration in the excavation damaged zone around a heated excavation in salt. Salt is attractive as a disposal medium for radioactive waste because it is self-healing and is essentially impermeable and essentially non-porous in the far field (away from excavations). Investigation of the short-term (days to years) near-field (centimeters to tens of meters) behavior of salt is important for radioactive waste disposal because this early period strongly controls the amount of brine in a salt repository. Brine leads to corrosion of waste forms and waste packages, and possible dissolution of radionuclides with brine transport being a potential transport vector to the accessible environment. The main test case used in Task E is the ongoing Brine Availability Test in Salt (BATS) heater test located underground at the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico, USA. The Task was divided into a series of Steps. Step 0 was an introduction to processes in salt, that included matching historical unheated brine inflow data from boreholes at WIPP and matching temperature observations during BATS heater test 1a. Step 1 included validation of models against a thermo-poroelastic analytical solution relevant to heated boreholes in salt, and two-phase flow around an excavation in salt. Step 2 required all the individual components covered in steps 0 and 1 to come together to match observed brine inflow behavior during the BATS 1a heater test. There were a range of approaches from the teams, from mechanistic to prescriptive. Given the uncertainties in the problem, some teams used one- or two-dimensional models of the processes, while other teams included more geometrical complexity in three-dimensional models. The key learning points from Task E have been: • Heat conduction through salt typically requires non-linear thermal conductivity (as a function of temperature), but most models do a good job matching observations, given appropriate adjustments to the applied power and some thermocouple locations. • Thermal pressurization requires coupled thermal-hydrological-mechanical (THM) responses that consider the thermal expansion of the fluid and solid phases. • Initialization of two-phase flow models around a borehole or excavation in salt are more realistically represented as “wetting up”, rather than “drying down” (i.e., the initial state after excavation is mostly dry, rather than mostly wet). • The BATS 1a heater test includes a significant release of brine after the end of heating, which requires a large increase in permeability to recreate. Task E has been a great learning experience for all the teams involved, and feedback from the modeling teams has led to changes in the design of follow-on BATS experiments, which are now ongoing underground at WIPP. There was a balance throughout the task between freedom to model phenomena how each team saw fit, and prescriptiveness in problem design to bring the modeling teams closer together to allow attribution of smaller differences between models to different modeling choices. The modeling approaches seem to go through two phases: an early phase of discovery or testing, and a later phase of refinement and improvement. In future modeling efforts, different field data could be used (e.g., BATS 2) and more time should be included in the processes for teams to make multiple model refinement or even significant changes to their conceptual model or setup, based on lessons learned from the modeling exercise.

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

Final Report: Vitrification of Inorganic Ion-Exchange Media, VSL-16R3710-1

One of the primary roles of waste pretreatment at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) is to separate the majority of the radioactive components from the majority of the nonradioactive components in retrieved tank wastes, producing a high level waste (HLW) stream and a low activity waste (LAW) stream. This separation process is a key element in the overall strategy to reduce the volume of HLW that requires vitrification and subsequent disposal in a national deep geological repository for high level nuclear waste. After removal of the radioactive constituents, the LAW stream, which has a much larger volume but smaller fraction of radioactivity than the HLW stream, will be immobilized and disposed of in near surface facilities at the Hanford site.

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

Potential for Transport of Cesium as Bio-colloids in a High Ionic Strength System

The Waste Isolation Pilot Plant (WIPP) is a deep geologic repository for long-term disposal of transuranic, radioactive waste that is a byproduct from the nation's nuclear defense program. Most likely release scenario → human intrusion (potentially through drilling) (US DOE 1995; 1996). Brine may proceed through the Rustler formation (the most transmissive layer) (Perkins et al., 1999). There is a need to investigate halophilic microbes including their mobility in the Rustler formation and the potential for enhancing transport of cesium as a bio-colloid. Objectives: To quantify the change in mobility of Cs by interaction with microbes (Chromohalobacter) that may be present near Waste Isolation Pilot Plant. What is Chromohalobacter?: a halophilic microbe, isolated from near the WIPP site, able to thrive in the high salt concentrations relevant to the repository. Does the microbe growth phase effect uptake of Cs? Experiments conducted with actively growing and stationary phase Chromohalobacter. Uptake may differ if Cs{sup +} is mistaken for K{sup +} in an actively growing (Log) phase versus stationary phase microbes (not actively reproducing). Materials: 1'' Teflon column packed with 1 gram of dolomite [355-500 μm, CaMg(CO{sub 3}){sub 2}] connected to a syringe pump and fraction collector via Teflon tubing. Chromohalobacter (100 million cells/mL) and dolomite collected near the WIPP. Synthetic brine - 15% NaCl (w/v ∼2.78 M) + 3 mM NaHCO{sub 3}. Methodology: Negative Control 5,000 ppb Cs initially injected into mini-columns with brine. Injection solution was then switched to only brine. Cs Experiments with Stationary Phase Chromo: Set 1: Viable stationary Chromo injected into minicolumns after reaction with 200 ppb Cs. Injection solution was then switched to only brine (without Cs or microbes). Set 2: Stressed stationary Chromo initially injected into mini-columns after reaction with 200 ppb Cs. Then, switched to brine only. Cs Experiments with Log phase Chromo: Set 1: Chromo injected into mini-columns after reaction with 5,000 ppb Cs. Injection then switched to only brine (without Cs or microbes). Set 2: Chromo grown with 5,000 ppb Cs and spiked with additional Cs before injection into columns. Preliminary Conclusions: Cs was not taken up or mobilized by Chromohalobacter in the actively growing or stationary phases or under stressed conditions. Competition may have occurred between Cs and K{sup +} at lower ionic strength, significantly greater concentrations of K{sup +} were present. Results suggest that Chromohalobacter can selectively uptake K{sup +} in the presence of Cs{sup +}. Future Research: K{sup +} and Cs{sup +} uptake will be monitored at high ionic strength in K{sup +} depleted media to consider uptake of Cs in the absence of K.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Designing a repository in domal salt: The influence of design variants in different modelling environments

To understand the long-term environmental impact of disposing radioactive waste of in a deep geological repository and to optimise its design, performance assessments are used. In this study, four teams (COVRA, GRS, Quintessa, and DOE) modified the previously developed generic repository of DECOVALEX task F2 to identify commonalities and differences between the teams for specific changes in repository design. The teams tested six design modifications: (1) Replacing concrete abutments with run-of-mine salt; (2) Replacing the salt seal with a concrete abutment and using run-of-mine salt instead for the two concrete abutments in each drift seal; (3) Halving the size of the infrastructure area; (4) Using run-of-mine salt instead of gravel for backfilling the infrastructure area; (5) Disposal of spent nuclear fuel without the POLLUX-10 containers (6); Lower initial saturation of the spent nuclear fuel and vitrified waste disposal drifts. Despite different modelling strategies used, models agreed that a smaller infrastructure area has a limited effect on radionuclide transport. Responses to the absence of the two concrete abutments in each seal, the use of single large concrete abutments (200 m each), or the use of run-of-mine salt in the infrastructure area differ between teams due to differing modelling assumptions. Based on these results, the estimated efficacy of containment depends strongly on the model assumptions of each team. More specifically, it appears to depend on the compaction model used and therefore on the backfill material used in different areas of the repository. However, the drift seal appears to be a critical design element in all models, effectively limiting radionuclide transport by hydrologically disconnecting sections of the repository. Additional beneficial design choices include the use of dry salt in disposal drifts to limit radionuclide transport and reducing the infrastructure area costs and minimizing host rock damage.

DECOVALEX↗
DOE OSTI2015

Development and Durability Testing of a Low‐Temperature Sintering Bi–Si–Zn Oxide Glass Composite Material (GCM) 129 I Waste Form

The capture and safe storage of radiological iodine ( 129 I) from nuclear fuel reprocessing is of concern due to its long half‐life and potential mobility in the environment. The development of durable waste forms in which to store captured iodine requires materials that are both compatible with the iodine capture phases and durable to repository environments. To that end, Sandia is developing a low‐temperature sintering Bi–Si–Zn oxide glass composite material (GCM) waste form and herein presents results of durability testing. Furthermore, durability studies were extended to both the occluded iodine capture material Ag‐Zeolite (Mordenite, MOR) as well as the GCM, synthesized with compositional variations including: amount of Ag flake added, AgI‐MOR particle size in the GCM, and mass loading of I within the AgI‐MOR. Product consistency test (PCT‐B), chemical durability (MCC‐1) tests, and single‐pass flow‐through (SPFT) tests, were performed on both the individual components of the GCM and the completed GCMs. Durability tests indicate low GCM dissolution rates (<10 −3 g/m 2 ·d) across wide variable ranges including: pH, AgI‐MOR loading, I loading, and AgI‐MOR particle size. Results indicate that the Bi–Si–Zn oxide glass matrix sharply limits the release of iodine from the otherwise relatively fast degrading AgI‐MOR getter material. Furthermore, the formation of an amorphous AgI phase during sintering of the GCM results in the limitation of iodine release during waste form degradation. Durability of GCM and release rates approximate those of established nuclear waste glasses, or analogues such as basaltic glass. This suggests that the Bi–Si–Zn GCM is a viable candidate as a repository iodine waste form.

Mowry, Curtis D.↗
DOE OSTI2020

Reduction of the Waste Occupied Area by Nuclide Separation and Horizontal Emplacement of the Waste Packages - 20422

Two of the key objectives for realizing the advanced nuclear fuel cycle are to reduce the volume of high-level radioactive waste and to reduce the foot-print of the geological repository. To achieve these objectives, research and development on nuclide separation technologies is a prerequisite. In this context, it is necessary to consider the vitrification conditions and the design of the geological repository where separation technologies would be implemented to achieve effective reductions in the volume of vitrified waste and the foot-print of the geological repository. In this study, the effects of the vitrification conditions on the number of vitrified waste units and the disposal area were evaluated, based on the assumption that separation of Cs and Sr from the spent nuclear fuel had been performed. In addition, with reference to the design of the geological repository, the impact on the foot-print of the repository for different emplacement methods of the waste packages was examined. As a result, it was clarified that the waste-occupied area could be reduced by horizontal emplacement of the vitrified waste with low heat generation. In addition, it was demonstrated that a low waste-loaded vitrified waste contributed to a reduction in the foot-print of the geological repository in the case of horizontal emplacement, although the number of vitrified waste units increased. (authors)

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

French Atomic and alternative energies commission decommissioning policy and strategies: What is at stake lessons learned and feedback experience - 15013

Since its creation in 1945, the French Atomic Energy and Alternative Energies Commission has carried out research programmes on civil use of nuclear, and its application. CEA has set up and run various types of installations from research (accelerators, pond reactors, fast reactors, laboratories) to industry (power reactor, processing facilities). There are currently more than several tens of facilities which are presently closed down pending dismantling or in the process of being dismantled by CEA, the governmental body for research and development in nuclear and renewable energies. When operating, all these installations have generated radioactive waste. When they are shutdown, either as a result of the programmes for which they were set up, or due to equipment wearing out or being obsolete, these installations must be cleaned up, dismantled and decommissioned, taking into account their nuclear nature and requirements related to safety and environment. During the decommissioning process, the implementer has to deal with safety and site remediation issues. Dealing with these installations is a long-term programme which must be carefully planned and financed in order to optimise its management. The paper will describe the CEA's decommissioning policy, strategies adopted, and draw up lessons learned and feedback experience, showing that: - Safety is guaranteed, - Dismantling is a technically controllable operation, but R and D actions are to be developed for specific problems as characterization or operation in hostile environment. Moreover all the methods and techniques improve as a result of feedback experience, - The amount of material and waste produced by the dismantling are manageable. The very low, low and intermediate level waste can be stored on existing shallow land burial centres, some new repositories have to be created for specific waste (graphite, radium,...) and for high and intermediate level long live waste, which are stored in interim storage , - The dismantling must be optimized, especially by appliance of a good strategy based on rational industrial choices. - Communication and transparency with stakeholders is essential for global process acceptance, - The lessons learned and how feedback experience could benefit for future decommissioning projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Panel Session 26: Decommissioning, Issues and Possibilities/ Opportunities

The panel focused on the latest Issues and Opportunities in D and D around the world. The format of this panel was aiming at an open exchange of experience and ideas based on actual decommissioning and gave the panelists and the audience an opportunity to share lessons learned as well as networking during and after the session. For the session, the panelists with experience from different fields in decommissioning such as dismantling, waste management, regulatory aspects and repository management participated. Interesting topics to discuss with the audience were: 1) Logistics (from cradle to grave); 2) Organizational aspects of decommissioning; 3) Regulatory aspects and issues; 4) Definition of the back-end with regard to packages and Waste Acceptance Criteria (WAC); and 5) Identification and management of difficult-to-handle D and D waste (e.g. damaged fuel, legacy/forgotten waste inside Nuclear Power Plants, etc. Presentation proposed: Decommissioning - Issues, Possibilities and Opportunities (David Peattie, Nuclear Decommissioning Authority (United Kingdom))

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

Response of bentonite microbial communities to stresses relevant to geodisposal of radioactive waste

Microbes have been isolated previously from bentonite materials that may be used as barriers for the disposal of radioactive waste. Actively respiring microbes in such barrier materials, within a repository environment, have the potential to adversely affect waste container corrosion rates. Additionally, they could potentially alter the properties of the bentonite barrier itself. This is of significance, since the integrity of the waste container and properties of the bentonite barrier are required to fulfil defined safety functions. Furthermore to help identify the critical factors that affect microbial activity in bentonite materials, this study examines the impact of a range of parameters that could affect microbial metabolism in a geodisposal environment. Several bentonites from different sources (bentonite mined from locations in Spain and the USA, along with commercially-sourced bentonite) were subjected to increased pressure (74 MPa, 30 s), heat (90 °C, 24 h), and irradiation (1000 Gy, 24.17 Gy min –1 ), before incubation in growth media selective for sulfate-reducing bacteria (SRB) or iron-reducing bacteria (IRB).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Reconsolidated Salt as a Geotechnical Barrier

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt repositories have the potential to isolate permanently vast inventories of radioactive and hazardous wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Reconsolidated Salt as a Geotechnical Barrier - 16535

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt repositories have the potential to isolate permanently vast inventories of radioactive and hazardous wastes. (authors)

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

Legacy Waste Containers for Low and Intermediate Level Radioactive Waste in Germany - Actual Situation and Challenges - 17070

Low and intermediate level radioactive waste has been produced in Germany for several decades. While a part of the waste had been deposited in the Morleben repository and the Asse mine which have been operated for just a limited period of time, the major amount is conditioned and packaged in temporarily stored waste containers. Those containers are designated for final disposal in the KONRAD repository being currently under construction and expected to start operation in 2022. As a consequence, an assessment of the waste container documentation and safety is strongly needed to receive an approval for the Konrad repository. This requalification is challenging because those old containers have to meet the same safety requirements as new containers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

The isotope 99 Tc is an important fission product generated from nuclear power production. Because of its long half-life (t 1/2 = 2.13 ∙ 10 5 years) and beta-radiotoxicity (β⁻ = 292 keV), it is a major concern in the long-term management of spent nuclear fuel. In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up. In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form. Technetium waste forms under consideration include metallic alloys, oxide ceramics and borosilicate glass. In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository. Metallic alloys under consideration include Tc–Zr alloys, Tc–stainless steel alloys and Tc–Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long-term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed, but efforts to model the behavior of Tc metallic alloys are limited. One parameter that should also be considered in predicting the long-term behavior of the Tc waste form is the ingrowth of stable Ru that occurs from the radioactive decay of 99 Tc ( 99 Tc → 99 Ru + β⁻). After a geological period of time, significant amounts of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance. In this context, we studied the electrochemical behavior of Tc metal, Tc-Ni alloys (to model Tc-Inconel alloy) and Tc-Ru alloys in acidic media. The study of Tc-U alloys has also been performed in order to better understand the nature of Tc in metallic spent fuel. Computational modeling and simulations were performed to shed light on experimental results and explain structural and kinetics trends.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗