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At least 793 records · Page 44

Use of Robotics in Liquid Waste System - {sup 3}H Evaporator, Tank Closure and Defense Waste Processing Facility - 18525

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Salt-stone Production Facility (SPF). At Savannah River Remediation (SRR), the liquid waste contractor, robotic technologies are needed to enter areas and spaces that are otherwise inaccessible or prohibit direct access by workers due to unsafe, unstable, or unknown physical or structural conditions; or configurations that are hard to reach or beyond reach without taking extraordinary mechanical measures. As discussed below, SRR uses robotic technologies at various LWS facilities to reduce dose to the workers as well as provide access to unreachable locations due to the presence of radiological or chemical hazards. (author)

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Risk considerations for WIPP passive institutional controls - 15173

The regulatory assurance measures for the Waste Isolation Pilot Plant (WIPP) include such things as active and passive institutional controls, monitoring, waste retrieval/removal capabilities and engineered barriers. Passive Institutional Controls (PICs) are an assurance measure intended to reduce the likelihood of humans unintentionally intruding the WIPP geologic radioactive waste disposal system for thousands of years after it is closed. The waste repository is designed such that it should not pose a significant risk to future populations if it is left undisturbed. The design has also been demonstrated through a performance assessment (PA) analysis to not pose a significant risk to future generations even if it is intruded upon by an inadvertent drilling intrusion [1, Appendix PA]. Through PA, it has been determined that the overall risk of the disposal system to people and the environment is small. The intent of the regulatory Assurance Requirements and PICs is to provide additional confidence that the repository will perform as expected after closure. However, the actual implementation of the proposed PICs design may pose a greater risk to the current population than they would potentially reduce for future generations. Specifically, the risks due to industrial accidents, construction and transportation risks incurred during the fabrication and construction of the permanent markers system may be greater than the overall reduction in risk to future generations due to the existence of the permanent markers. If this were true, implementation of the current PICs design may be counter to the intention of the regulatory requirement to include such assurance measures. As such, provisions for assurance measures used in geologic radioactive waste disposal facilities, such as PICs, should consider risk impacts. A provision of any disposal program's Regulatory Assurance Requirements should address the impacts of the known risks from all the components of the assurance-related systems. The initial WIPP PICs design was developed as part of the regulatory certification process in 1996 with the intent of being revised when the WIPP closes several decades from now [2, Appendix PIC]. The revised design that will be developed at closure should address the design's potential reduction to long-term risks from the repository and the risks due to the implementation of the final design. (authors)

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Stability, Composition, and Core-Shell Particle Structure of Uranium(IV)-Silicate Colloids

Uranium is typically the most abundant radionuclide by mass in radioactive wastes and is a significant component of effluent streams at nuclear facilities. Actinide(IV) (An(IV)) colloids formed via various pathways, including corrosion of spent nuclear fuel, have the potential to greatly enhance the mobility of poorly soluble An(IV) forms, including uranium. This is particularly important in conditions relevant to decommissioning of nuclear facilities and the geological disposal of radioactive waste. Previous studies have suggested that silicate could stabilize U(IV) colloids. Here the formation, composition, and structure of U(IV)-silicate colloids under the alkaline conditions relevant to spent nuclear fuel storage and disposal were investigated using a range of state of the art techniques. The colloids are formed across a range of pH conditions (9-10.5) and silicate concentrations (2-4 mM) and have a primary particle size 1-10 nm, also forming suspended aggregates <220 nm. X-ray absorption spectroscopy, ultrafiltration, and scanning transmission electron microscopy confirm the particles are U(IV)-silicates. Additional evidence from X-ray diffraction and pair distribution function data suggests the primary particles are composed of a UO2-rich core and a U-silicate shell. U(IV)-silicate colloids formation correlates with the formation of U(OH)3(H3SiO4)3 2- complexes in solution indicating they are likely particle precursors. Finally, these colloids form under a range of conditions relevant to nuclear fuel storage and geological disposal of radioactive waste and represent a potential pathway for U mobility in these systems.

Neill, Thomas↗

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)

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Selection of Vadose Zone Flow and Transport Properties with Gravel Fraction Corrections for the Hanford Site Composite Analysis and Cumulative Impact Evaluation

This environmental calculation file (ECF) is a compilation of hydrologic and radiological/chemical properties to be used for the updated Hanford Site Composite Analysis (CA) and Cumulative Impact Evaluation (CIE) vadose zone (VZ) modeling. The CA will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1 Chg 1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. The CIE evaluates the effects of cleanup decisions regarding groundwater quality in the Hanford Site Central Plateau (DOE/RL-2018-69, Cumulative Impact Evaluation Technical Approach Document). Due to the complexity and large number of waste sites in source operable units (OUs), the computational tools used for the CIE must be capable of representing a range of site conditions and source terms in the VZ while also efficiently computing the impact that cleanup decisions have on the underlying aquifer. Waste-site proximity between and within source OUs has resulted in contaminants commingling in the vadose and saturated zones in complex ways. Plume commingling requires cleanup decisions to be evaluated considering the surrounding waste sites and existing groundwater contamination, therefore demonstrating the need to evaluate cumulative impacts from the VZ to groundwater.

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Proton Nuclear Magnetic Resonance (H NMR) of Glycolate in Real Waste: Developing and Testing Analytical Methods for the Savannah River Site Liquid Waste System

In preparation for implementing the NG acid flowsheet for the SRS LWS, analytical methods for determining glycolate at low concentration are desired to support system management and safety. Previous work documented that the IC method performed well for samples with low to moderate ionic strength. That work also included a scoping effort to determine if an alternate analytical strategy using H NMR would complement and extend the capabilities of the IC method. The H NMR scoping results indicated that the method had the potential to expand glycolate analysis in LWS samples to higher ionic strength tanks/solutions. The scope of this work is to develop innovative proton NMR techniques, including demonstrating ion exchange decontamination protocols needed for application of the technique to real waste samples. Part of radioactive waste processing at SRS uses formic acid to reduce oxidized (Hg 2+ ) to more volatile elemental Hg for steam striping, collecting, and disposal. Under acidic conditions found in the Chemical Processing Cell (CPC) at the Defense Waste Processing Facility (DWPF), formic acid has a much higher hydrogen generation rate than an alternative reductant, glycolic acid. Thus, an NG acid flowsheet has been developed utilizing glycolic acid with the benefit of easing the need for headspace monitoring requirements for hydrogen and ammonia at DWPF. Low concentrations of glycolate are conservatively assumed to be in the recycle stream, which will collect in the Recycle Collection Tank (RCT). The DWPF recycle stream collected in the RCT has a distinct pathway to the LWS waste tanks that feed the 2H and 3H Evaporator. This route involves transfer of the DWPF recycle to Tank 22 in the Concentration, Storage, and Transfer Facilities (CSTF) followed by transfer to the LWS tank farm/evaporators.

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Hanford Site Composite Analysis: Radionuclide Selection for Groundwater Pathway Evaluation.

The updated Hanford Site Composite Analysis (CA) will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1-1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. A key aspect of conducting a CA is selecting the radionuclides to be included in the quantitative analysis.

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Hot Isostatically Pressed Ceramic Waste forms for High Dose Spent Adsorbents - 18129

Decommissioning and clean-up of nuclear facilities requires the development of new technologies to condition radioactive wastes, producing passively safe waste packages of minimal volume, to reduce storage and disposal costs. We have applied hot isostatic pressing to demonstrate conceptual waste forms for high dose spent adsorbents (i.e. ion exchange materials) present on the Sellafield site, UK, and Fukushima Daiichi site, Japan. These radioactive wastes pose several challenges which demand early conditioning to produce a passively safe wasteform, including: a wet and granular nature, and hence dispersible as a result of loss of containment; a very high dose rates as a result of their selectivity for short lived radionuclides (e.g. Sr-90, Cs-137, Co-60); radiogenic self-heating, as a result of the concentration of such lived radio nuclides; hydrogen production as a result of radiolysis of entrained water [1]. Here, we demonstrate the conversion of natural inorganic ion exchange materials into multiphase ceramic waste forms, achieving a waste loading of 100 wt.% and density in excess of 97% theoretical. High resolution thermogravimetric analysis coupled with mass spectroscopy was utilised to characterise the evolution of water and volatiles during the in-can bake out step, prior to the HIP cycle. This allowed optimisation of the bake out parameters (temperature, time and vacuum) to enable complete removal of water and volatiles, affording ceramic bodies with minimal residual porosity by hot isostatic pressing at 1250 deg. C for 4 h in stainless steel canisters. Characterisation of the ceramic waste forms by SEM /EDX and X-ray diffraction revealed the nuclides of concern to be incorporated within well-known natural mineral and synthetic phases, with Cs partitioning into the K-feldspar. (authors)

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Task Order 22 – Engineering and Technical Support, Deep Borehole Field Test. AREVA Summary Review Report

Under Task Order 22 of the industry Advisory and Assistance Services (A&AS) Contract to the Department of Energy (DOE) DE-NE0000291, AREVA has been tasked with providing assistance with engineering, analysis, cost estimating, and design support of a system for disposal of radioactive wastes in deep boreholes (without the use of radioactive waste). As part of this task order, AREVA was requested, through a letter of technical direction, to evaluate Sandia National Laboratory’s (SNL’s) waste package borehole emplacement system concept recommendation using input from DOE and SNL. This summary review report (SRR) documents this evaluation, with its focus on the primary input document titled: “Deep Borehole Field Test Specifications/M2FT-15SN0817091” Rev. 1 [1], hereafter referred to as the “M2 report.” The M2 report focuses on the conceptual design development for the Deep Borehole Field Test (DBFT), mainly the test waste packages (WPs) and the system for demonstrating emplacement and retrieval of those packages in the Field Test Borehole (FTB). This SRR follows the same outline as the M2 report, which allows for easy correlation between AREVA’s review comments, discussion, potential proposed alternatives, and path forward with information established in the M2 report. AREVA’s assessment focused on three primary elements of the M2 report: the conceptual design of the WPs proposed for deep borehole disposal (DBD), the mode of emplacement of the WP into DBD, and the conceptual design of the DBFT. AREVA concurs with the M2 report’s selection of the wireline emplacement mode specifically over the drill-string emplacement mode and generically over alternative emplacement modes. Table 5-1 of this SRR compares the pros and cons of each emplacement mode considered viable for DBD. The primary positive characteristics of the wireline emplacement mode include: (1) considered a mature technology; (2) operations are relatively simple; (3) probability of a radiological release due to off-normal events are relatively low; (4) costs are relatively low; and (5) maintenance activities are relatively simple. The primary drawback associated with the wireline emplacement mode for DBD is the number of emplacement trips-in to the borehole, which results in a relatively higher probability for a drop event. Fortunately, the WPs can be engineered with impact limiters that will minimize the likelihood of a breach of the WP due to a drop. The WP designs presented in the M2 report appear to be focused on compatibility with the drill-string emplacement mode (e.g., the threaded connections). With the recommendation that the wireline emplacement mode be utilized for the DBFT, some changes may be warranted to these WPs. For example, the development of a WP release connection that is more reliable than the currently credited connection, which is considered to have a high failure probability, and the integration of an impact limiter into its design. The M2 report states the engineering demonstration of the DBFT will occur in the FTB over a 4-year period. AREVA recommends development and testing of the WP emplacement handling equipment occur separately (but concurrently, if not earlier) from the FTB at a mock-up facility. The separation of this activity would prevent schedule interference between the science and engineering thrusts of the project. Performing tests in a mock-up facility would allow additional control and observation compared to the FTB. The mock-up facility could also be utilized as a training facility for future operations. Terminal velocity and impact limiter testing would require the FTB for testing, since these areas would be difficult to reproduce in a limited depth mock-up. Although only at the end of the conceptual stage of design development, DBD appears to be a viable solution for some waste forms produced by the nuclear industry. However, regulatory requirements have yet to be established for pre- and post-closure performance of DBD and should be established as soon as possible. Some of the main areas of focus from a regulatory perspective include: (1) establishing acceptable performance requirements for the long-term behavior of DBD; (2) determining acceptable borehole abandonment criteria; (3) establishing retrievability requirements; (4) developing a consensus on the factor of safety (FoS) for the emplacement mode and WP; and (5) establishing safety and safeguards performance requirements for DBD. Although conservative requirements have been utilized to provide the foundation for the conceptual design of DBD, regulatory requirements and feedback are necessary to confirm recommendations made herein and to ensure the long-term performance of DBD is acceptable. The combination of the M2 report and this SRR is intended to facilitate the completion of the conceptual design for DBD for the Cs and Sr capsules and calcined waste forms. Using the conceptual design, preliminary design activities (the second stage of a three-stage process described in the M2 report) can proceed and the DBFT utilized to support, demonstrate, and confirm engineering elements of this design.

42 ENGINEERING↗

Initial Site-Specific Studies for Removing SNF from Six Shutdown Sites - 18042

The US Department of Energy (DOE) is exploring options for developing a large-scale transportation system for the safe transport of spent nuclear fuel (SNF) and high-level radioactive waste to future storage or disposal facilities. This complex, large-scale transportation system will involve coordinated, integrated activities. Interfaces with various transportation and non-transportation activities must be defined, and systems must be developed to ensure successful integration when the system becomes operational. To provide an understanding of the tasks and interfaces required for the complete de-inventory of particular sites where SNF is stored, DOE solicited bids from integrating contractors to develop initial site-specific de-inventory reports. The reports were developed for several nuclear power plant sites selected to represent a range of geographic locations, existing on-site SNF storage systems, and expected transportation modal options. Each report is presented in a self-contained narrative describing a potential campaign for removing all SNF and greater-than-class-C low-level radioactive waste (GTCC waste) from each specified site. Each study narrative includes site information; an analysis of the near-site transportation infrastructure; a concept of operations for the de-inventory; an estimated budget and schedule for the duration of operations; and a safety and security plan with site-specific considerations, emergency response considerations, and recommendations for the next steps DOE could take to de-inventory the site. The reports provide cost estimates for analytical purposes only; they do not attempt to specify the party or parties responsible for the costs. These reports present a first look at how an integrating contractor might recommend removing SNF from particular sites. They are written from the perspective of the contractor and do not represent DOE's plans. At this writing, six sites have been studied: - Connecticut Yankee, - Big Rock Point, - Humboldt Bay, - Maine Yankee, - Trojan, - Kewaunee. DOE may develop de-inventory reports for all reactor sites, with an initial focus on shutdown reactor sites. The order in which sites have been selected for study should not be considered an indicator of the order used by DOE to remove SNF from the sites. Since the SNF's destination is not important for the purpose of this work, DOE tasked the contractor with planning transportation of the SNF from the site to a generic railroad location in the approximate geographical center of the United States. These reports represent an initial study of the specific sites. Per the terms of the contract, the contractor was not permitted to engage in dialogue with power plant site personnel; near-site railroad personnel; or any local, state, or regional stakeholders about the studies. DOE provided site-specific data. The contractor was tasked to rely on past corporate experience, data available from DOE, and publicly available data when making site-specific recommendations for removing all SNF from each site. (authors)

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Update on Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 19092

This paper provides an overview of the progress to date, and discussion of the path forward, related to designing, fabricating and testing prototype railcars that will comply with the safety standard S-2043. This standard was developed by the Association of American Railroads (AAR) specifically for railcars used to transport High-Level Radioactive Material (HLRM). AAR defines the term HLRM to include both spent nuclear fuel (SNF) and high-level radioactive waste (HLW). DOE is in the process of developing and testing prototype railcars that will satisfy Standard S-2043. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. (authors)

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Regulatory Consideration on Decommissioning Facilitation of Nuclear Power Plants In Korea

As the number of decommissioning NPPs worldwide increases, the necessity of decommissioning feasibility design at the initial design/construction stage of NPP is emphasized. Also as demand for NPP decommissioning increase, the generation of large amount of radioactive waste and environment contamination are expected. Based on the decommissioning experiences of nuclear facilities, it is recognized that the specific design and operational procedure are essentially required to minimize the radioactive waste generation and radiation exposure during decommissioning. Recently the legal framework of decommissioning in Korea was enhanced through the revision of Nuclear Safety Act (NSA) and its subordinate status in 2015. On the other hand, the Korean first NPP, Kori unit 1, will be scheduled permanently shut-down in mid-2017. The Korean regulatory body have experience for decommissioning of small-sized nuclear facilities like research reactors and nuclear fuel cycle facility. However we have no experience on decommissioning of NPP in Korea. Therefore the comprehensive and systematic technical standards and guidelines on safety review and inspection should be required for the Korean first case on decommissioning of NPP. According to the revised NSA, NPP operators in Korea should submit the decommissioning plan (DP) for the permanent shut-down NPP as well as the existing operating NPPs. The operator shall prepare and submit the design features and measures for decommissioning facilitation as one of DP items and regulatory body shall review the feasibility of decommissioning. Therefore the safety review guideline for those decommissioning facilitation should be required. In this paper, we studied the overseas safety requirements, design features and operational procedures for decommissioning facilitation. Through the overseas survey results, the regulatory position and major management plan were proposed for development of detailed implementation guideline on decommissioning facilitation of NPPs in Korea.

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Developing a Process for Collaboration-Based Siting of a Federal Consolidated Interim Storage Facility in the United States: Learning from Communities Living with Legacy Waste

In June of 2023, the U.S. Department of Energy (DOE), Office of Nuclear Energy (NE) announced the selection of its Collaboration-Based Siting (CBS) Consortia, a group of 12 awardees, including the Consortium for Risk Evaluation with Stakeholder Participation (CRESP) (led by Vanderbilt University) to assist DOE-NE with the development of its process for siting a federal consolidated interim storage facility (FCISF) for spent nuclear fuel (SNF) storage. At this time, DOE NE is not soliciting interested host communities, rather the CBS Consortia are tasked with in-depth engagement, mutual learning, and capacity building to provide DOE NE with feedback on the CBS process. CRESP’s objective is to engage communities in two regions (the Pacific Northwest and the Southeast) with sites currently storing defense- and research-related SNF to foster learning concerning the best and worst practices in community participation in risk-informed decision making. This includes learning from existing structures for public input in radioactive waste management decision making [e.g. citizen advisory boards (CABs)] and other local parties about how to build and sustain trust among the parties. CRESP has been working to engage stakeholders and Tribes surrounding two DOE sites historically differing in receptiveness to engaging with DOE and trusting in DOE to accomplish its missions. CRESP’s approach to date has consisted of (1) engaging voluntarily members and former members of the DOE Office of Environmental Management’s CABs to develop a mutual understanding of their perspectives, values, and experiences related to risk and participatory decision making; (2) engaging members of those communities that would likely be part of any radioactive waste management discussions and who may provide valuable feedback for the development of the CBS process; and in the longer term, (3) engaging communities to foster knowledge sharing on understanding and definitions of risk and how risk is factored into community decision making. Our emphasis is to identify opportunities for improving risk communication frameworks, strategies, and decision making. The selection of a future FCISF site is likely to result in the creation of a structure similar to a CAB composed of members of the local community. We anticipate that these insights can help DOE NE learn from existing participatory risk communication structures in place at sites storing defenseor research-related SNF to understand best practices for fostering enduring and participatory relationships with future CABs. Within this paper, we (1) describe CRESP’s overall approach to assisting DOE NE with maturing the CBS process; (2) present a summary of preliminary phase 1 project results, including the development of a body of knowledge (describing available resources to support community engagement, risk communica tion, and participatory decision making) and community ecosystem information (leveraging demographic, social, economic, and environmental attribute mapping and advanced sentiment analysis of social media data); and (3) based on these results, provide observations for future research opportunities to support the development of CBS processes for radioactive waste management facilities, generally.

collaboration-based siting↗

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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Effects of Gamma Radiation on Cementitious Materials in Repository Environment - 16300

Cementitious materials are widely used in the management of radioactive waste and have to perform as required for the period of service life of the installation. In order to assess the long term durability of cement paste in a deep borehole for radioactive wastes, specimens were cast in laboratory and exposed to a gamma radiation field. The effects of the radiation in the specimens were evaluated by X-Ray Diffraction, as changes observed in the mineralogy and microstructure of the material after irradiation doses varying between 0 (without radiation) and 10 MGy. It was observed that the specimens exposed to higher doses of gamma radiation (up to 3 MGy) presented differences between surface and core, when compared with the samples with no radiation or exposed to low doses. These results are important to model the behavior of cementitious materials under the conditions expected in a borehole repository and to provide data for the safety assessment of the installation. (authors)

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Hanford Site Composite Analysis Technical Approach Description: Radionuclide Inventory and Waste Site Selection Process.

The updated Hanford Site Composite Analysis will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at Hanford following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1-1, Radioactive Waste Management (DOE, 2001), related to managing low-level waste disposal facilities at the Hanford Site.

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Implementation of the Glycolate Destruction Process with Sodium Permanganate in the Defense Waste Processing Facility (DWPF) - 20141

The Savannah River Site (SRS) seeks to replace formic acid with glycolic acid as a chemical reductant in the Defense Waste Processing Facility (DWPF), where borosilicate glass is mixed with high-level radioactive waste, melted at high temperatures, then poured into stainless steel canisters for safe storage. Prior research and development have demonstrated the feasibility and advantages of the new nitric-glycolic acid flowsheet over the current nitric-formic acid flowsheet to chemically adjust the radioactive sludge slurry waste in the Sludge Receipt and Adjustment Tank (SRAT) prior to vitrification. Use of glycolic acid reduces hydrogen and ammonia generation during the reduction process, thus reducing flammability hazards and purge requirements in the DWPF processing vessels. In addition, glycolic acid will support the new Salt Waste Processing Facility (SWPF) by allowing for receipt of higher volumes of waste due to reduction of flammable gases, increased boil-up rates in the vessels, and easing the processing of future sludge batches containing high levels of mercury. However, during DWPF operations, trace amounts of glycolate are anticipated to be entrained in the recycle stream and sent to the Concentration, Storage, and Transfer Facilities (CSTF) via the Recycle Collection Tank (RCT). The RCT receives condensate from the Slurry Mix Evaporator Condensate Tank (SMECT) and Off-gas Condensate Tank (OGCT) that may contain unreacted glycolate. A literature review found the potential for hydrogen generation due to thermolysis of glycolate in caustic CSTF conditions. Savannah River National Laboratory (SRNL) conducted several tests on simulant and radioactive waste from various tanks at SRS and confirmed that glycolate added to waste at concentrations of approximately 1000 mg/L and higher generates hydrogen through thermolysis. There are potential operating scenarios in the future where glycolate concentration could exceed 1000 mg/L in the 2H Evaporator Drop Tank, the highest accumulation point for DWPF organics received via recycle. These findings presented a need to develop another process to mitigate the impact of glycolic acid to the CSTF. Several DWPF compatible options were evaluated for their ability to destroy glycolate in the DWPF recycle stream under various processing conditions, thereby mitigating its introduction to the CSTF. Downselection testing identified sodium permanganate as the most promising and simplest option to oxidize glycolate. It has been demonstrated to be effective in various RCT simulant compositions, operating temperatures, oxidant addition rates, solution pH, and initial glycolate concentrations. A feasibility study identified existing tanks within the facility that could be used for receiving and storing the commercially available sodium permanganate, as well as feeding it to the RCT for processing. The use of existing equipment as well as the ability to purchase the pre-mixed chemical reduces implementation and operational complexity. The downstream impacts of the sodium permanganate reaction were also evaluated and demonstrated to have minimal impact on other chemical species present in the RCT and CSTF, as well as on the materials of construction of the RCT and CSTF vessels and associated piping and instrumentation. Future work is in progress to validate the glycolate destruction process with testing in actual radioactive RCT waste before implementation in DWPF. (authors)

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Design, Fabrication and Start-up of Test Facilities for the LAWPS Full-Scale IX Column Test and Engineering Scale Test - 17418

The U.S. Department of Energy (DOE), Office of River Protection's (ORP) primary mission is to retrieve and treat Hanford's tank waste and close the Tank Farms to protect the Columbia River. Mixed radioactive waste is stored in 177 underground tanks at the Hanford Site as reported in DOE/ORP 2003-02, Environmental Impact Statement for Retrieval, Treatment, and Disposal of Tank Waste and Closure of the Single Shell Tanks at the Hanford Site, Richland WA, Inventory and Source Term Data Package. The 177 underground tanks are estimated to contain about 56 million gallons of waste. The Low Activity Waste Pretreatment System (LAWPS) Project provides for the early production of immobilized low activity waste (ILAW) by feeding LAW directly from Tank Farms to the Hanford Waste Treatment Plant's (WTP's) LAW Facility, bypassing the Pretreatment Facility. Prior to the transfer of feed to the WTP LAW Vitrification Facility, tank supernatant waste will be pretreated in the LAWPS to meet the WTP LAW waste acceptance criteria. The LAWPS will also facilitate the return of secondary liquid wastes from the WTP LAW Vitrification Facility to the Tank Farms. In November of 2015 DOE's Tank Operations Contractor, Washington River Protection Solutions (WRPS), awarded a contract to the AECOM Team to design, fabricate and operate two test platforms to generate performance data at two different scales and demonstrate technology maturation. The first platform models the ion exchange (IX) columns at full scale, and the second platform models the integrated flowsheet (cross flow filter (CFF) and IX) at 1/9. scale. Both test platforms underwent shakedown and start-up testing in August 2016 and moved into actual testing and NQA-1 data generation shortly thereafter. This paper covers the engineering design, procurement, construction, installation, and acceptance testing of the test platforms (which are now operational) under an aggressive fast track schedule, and initial test data as available. (authors)

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