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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

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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DOE OSTI2012

Retention of Anionic Species on Granite: Influence of Granite Composition - 12129

Technetium (Tc-99, T{sub 1/2} = 2.1.10{sup 5} yrs) and selenium (Se-79, T{sub 1/2} = 6.5.10{sup 4} yrs) belong among fission products, being produced by fission of nuclear fuel. Both elements can significantly contribute to risk due to their complicated chemistry, long life times, high mobility and prevailing anionic character. Therefore, knowledge of migration behaviour under different conditions can significantly improve input into performance and safety assessment models. Granite is considered as a potential host rock for deep geological disposal of radioactive waste in many countries. Granitic rocks consist usually of quartz, feldspar, plagioclase (main components), mica, chlorite, kaolinite (minor components). The main feature of the rock is advection governed transport in fractures, complemented with diffusion process from fracture towards undisturbed rock matrix. The presented work is focused on interaction of anionic species (TcO{sub 4}{sup -}, SeO{sub 4}{sup 2-}, SeO{sub 3}{sup 2-}) with granitic rock. Furthermore, the importance of mineral composition on sorption of anionic species was also studied. The batch sorption experiments were conducted on the crushed granite from Bohemian Massive. Five fractions with defined grain size were used for static batch method. Mineral composition of each granitic fraction was evaluated using X-ray diffraction. The results showed differences in composition of granitic fractions, even though originating from one homogenized material. Sorption experiments showed influence of granite composition on adsorption of both TcO4{sup -} and SeO3{sup 2-} on granitic rock. Generally, Se(IV) showed higher retention than Tc(VII). Se(VI) was not almost sorbed at all. Fe containing minerals are pronounced as a selective Se and Tc sorbent, being reduced on their surface. As micas in granite are usually enriched in Fe, increased sorption of anionic species onto mica enriched fractions can be explained by this reason. On the other hand, fractions enriched in feldspar did not show increased sorption affinity to Tc and Se. (authors)

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DOE OSTI2015

Saltstone 4QCY14 TCLP Toxicity and UTS Results

A Saltstone Disposal Facility (SDF) waste form was prepared in the Savannah River National Laboratory (SRNL) from a Tank 50H sample and Z-Area premix material for the fourth quarter of calendar year 2014 (4QCY14). After a 47 day cure, a sample of the SDF waste form was collected, and shipped to a certified laboratory for Toxic Characteristic and Universal Treatment Standards (UTS) analysis. The metals analysis is performed using the Toxic Characteristic Leaching Procedure (TCLP) 1 . The 4QCY14 saltstone sample results show that the saltstone is Resource Conservation Recovery Act (RCRA) nonhazardous, but is greater than the universal treatment standard for land disposal. The Saltstone Production Facility (SPF) and SDF were in a maintenance outage during the 4QCY14. Thus no processing or disposal of saltstone, as characterized by this 4QCY14 sample, occurred.

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

Cesium Ion Exchange Testing Using Crystalline Silicotitanate with Hanford Tank Waste 241-AP-107 - 19104

Pacific Northwest National Laboratory has been tasked with demonstrating the unit operations for the Low-Activity Waste Pretreatment System (LAWPS) with Hanford tank waste. Washington River Protection Solutions requested that Hanford tank waste collected from tank 241-AP-107 be processed using conceived pretreatment steps (suspended solids removal by filtration, Cs removal by ion exchange), then vitrified. Bench-scale ion exchange testing with approximately 9 L of filtered AP-107 supernate was conducted using crystalline silicotitanate (CST) ion exchange media. Batch contact testing was performed to evaluate CST performance on tank waste prior to processing it in the LAWPS. Subsequently, column testing was performed using prototypic LAWPS operations in a lead-lag column arrangement using 10-mL CST beds. Loading continued until the lag column reached the Hanford Tank Waste Treatment and Immobilization Plant waste acceptance criteria (WAC) for receiving supernatant waste for vitrification (a function of the Na and {sup 137}Cs concentrations). The Cs effluent from the lag column reached the WAC after processing half the feed volume. To keep the subsequent product effluent below the WAC, a new lag column was needed to continue processing and the original lag column was moved to the lead position. Greater than 99% of the Cs processed through the ion exchange system was collected on the CST in the columns. The AP-107 feed and column processing effluent underwent extensive characterization to better define waste characteristics and assess fractionation to the CST. Selected batch contact and lead column samples were analyzed to assess selected metal analyte load characteristics. Results of these analyses will be discussed. (authors)

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DOE OSTI2016

Characterization of Reconsolidated Crushed Salt from the BAMBUS Site.

Observational petrofabrics, thermal, mechanical, and hydrological measurements were made on reconsolidated salt samples extracted from the field site in which a study called Backfilling and Sealing of Underground Repositories for Radioactive Waste in Salt was conducted. Similar characterization was completed more than a decade ago, so this work furthers previous measurements after sustained consolidation in situ . Porosity determined by traditional point-counting on polished sections and helium porosimeter methods ranged from 20-25% with consolidation governed by brittle processes, as evidence of fluid-aided, grain-boundary processes was rarely observed. Thermal conductivity in the range of 2.3 W /( m * K ) is consistent for granular halite in this porosity range. Gas flow measurements yielded permeability of the order of 5e -13 m 2 . Pressure-sensitive compressive strengths at 0.5, 1.0, and 2.0 MPa confining pressure were 8, 9, and 14 MPa, respectively, with apparent elastic moduli increase with deformation.

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DOE OSTI2016

Characterization of Reconsolidated Crushed Salt from the BAMBUS Site

Observational petrofabrics, thermal, mechanical, and hydrological measurements were made on reconsolidated salt samples extracted from the field site in which a study called Backfilling and Sealing of Underground Repositories for Radioactive Waste in Salt was conducted. Similar characterization was completed more than a decade ago, so this work furthers previous measurements after sustained consolidation in situ . Porosity determined by traditional point-counting on polished sections and helium porosimeter methods ranged from 20-25% with consolidation governed by brittle processes, as evidence of fluid-aided, grain-boundary processes was rarely observed. Thermal conductivity in the range of 2.3 W/(m∙K) is consistent for granular halite in this porosity range. Gas flow measurements yielded permeability of the order of 5e- 13 m 2 . Pressure-sensitive compressive strengths at 0.5, 1.0, and 2.0 MPa confining pressure were 8, 9, and 14 MPa, respectively, with apparent elastic moduli increase with deformation.

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DOE OSTI2015

Longevity of geo-synthetic cover components - 15450

Geo-synthetics form many of the key components in modern landfill cell and final cover designs. Models for the long-term performance assessment of systems incorporating these components have traditionally relied on a small number of standard assumptions originally developed to address the needs of the Resource Conservation and Recovery Act (RCRA) Subtitle D solid waste industry. These past estimates demonstrated that, even under a series of highly conservative factors, the longevity of these components far exceeds the requirements for typical solid waste facilities. Traditionally, the interpretation of the geo-synthetic degradation research has been deterministic as opposed to probabilistic, often due to the relative scarcity of degradation data for particular formulations and environmental conditions. Long-term studies of the mechanical and chemical properties of geo-synthetics subjected to accelerated aging processes have been conducted in an attempt to better quantify their longevity to address applications where performance may be required for 1,000 years or more. Of the many cover system geo-synthetic components, geo-membranes appear to have the most significant impact on the long-term performance of cover systems. Based on available long-term aging studies, a series of property degradation curves and their corresponding model parameters are available for geo-membranes. A strong area of need in the performance assessment of shallow disposal facilities is how to connect these research findings with inputs to probabilistic models. Significantly, factors affecting geo-membrane performance, such as geo-membrane thickness, anti-oxidant package, molecular weight, oxygen exposure, and drainage, can be adjusted and re-engineered using a proposed approach to meet specific performance targets. (authors)

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DOE OSTI2018

Lessons Learned From Public Outreach: Effective Incorporation of the Stakeholder Voice - 18447

The licensing process for any facility associated with nuclear power or aspects of the nuclear fuel cycle can be expected to encounter intense stakeholder interest, and often significant challenges. Over the last decade, staff of the Center for Nuclear Waste Regulatory Analyses (CNWRA{sup R}) at Southwest Research Institute{sup R} (SwRI{sup R}) provided comprehensive support to the U.S. Nuclear Regulatory Commission (NRC) for public outreach efforts across a wide variety of nuclear activities. Many of these projects included public involvement and consultation activities for compliance with the National Environmental Policy Act (NEPA), including scoping, public meetings, responding to public comments on draft documents, and holding small meetings with affected local, state, and tribal governments. In addition, we have supported NRC public outreach related to nuclear facility licensing activities (including the proposed repository at Yucca Mountain), rulemaking, and regulatory guidance development. This experience has offered us a broad view of the challenges faced by industry, regulators, and stakeholders. In the process, we have observed some best practices that lead to effective stakeholder communication. Communication about complex technical issues is challenging, particularly for societally controversial topics, such as civilian use of nuclear power and management of radioactive waste. Most commonly, we have observed a lack of mutual trust and respect among industry, regulators, and stakeholders that creates an atmosphere ripe for suspicion and misunderstanding. Stakeholders have, in public comments, criticized the lack of appropriate venues to raise concerns in which they feel they can get meaningful responses from the nuclear industry and regulators. Conversely, industry faces difficulty in reaching compromise with stakeholders, who sometimes misinterpret or misrepresent factual information. Within this contentious environment, regulators are constrained by their statutory framework, even as they are challenged to be more responsive, transparent, and fair to both stakeholders and industry. In spite of these obstacles, there are examples where feedback from stakeholders was heard and utilized. Because of successful collection of appropriate feedback, regulators became more adept at understanding the full range of perspectives from both industry and stakeholders, and subsequently did a better job at effectively incorporating feedback and communicating their decisions and findings in a clear and understandable way. Ultimately, the success of these examples was based on a measure of mutual respect, which was reached through effective communication and implementation of some best practices. These best practices include (i) outreach planning that is abundantly mindful of the various stakeholder points of view, site history, regional characteristics, and probable future and related actions, combined with advance identification of key stakeholders; (ii) development and distribution of information that facilitates public participation; and (iii) use of avenues by which stakeholders can provide meaningful and effective comments. Illustrative examples include (i) how concerns raised by stakeholders were heard by an applicant and resulted in a revised project proposal; (ii) use of a variety of interaction formats that incorporated input across a broad spectrum of stakeholders; (iii) utilizing stakeholder feedback received during scoping; and (iv) how a facility underestimated the impact of 'outsider' stakeholder voices that were in contrast to the known local community. By moving past the viewpoint that public outreach is a 'necessary evil' and engaging only to check a procedural box, each participant can benefit from mutually satisfactory and improved outcomes. (author)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Current status of solid waste management in small island developing states: A review

Highlights: • Waste management is a matter of great concern for small island developing states. • On average, waste generation rate in these islands amounts to 1.29 kg/capita/day. • Illegal dumping and landfilling prevail in most small island developing states. • Sustainable waste management practices, previously absent, are now emerging. • However, many challenges still hinder the implementation of these practices. - Abstract: This article reviews the current status of waste management in Small Island Developing States (SIDS) and the challenges that are faced in solid waste management. The waste generation rates of SIDS were compared within the three geographic regions namely Caribbean SIDS, Pacific SIDS and Atlantic, Indian Ocean, Mediterranean and South China (AIMS) SIDS and with countries of the Organisation for Economic Co-Operation and Development (OECD). Only Pacific SIDS had a waste generation rate less than 1 kg/capita/day. The waste generation rates for the three SIDS regions averaged 1.29 kg/capita/day while that for OECD countries was at a mean value of 1.35 kg/capita/day. The waste compositions in the different SIDS regions were almost similar owing to comparable consumption patterns while these differed to a large extent with wastes generated in OECD countries. In SIDS, the major fraction of MSW comprised of organics (44%) followed by recyclables namely paper, plastics, glass and metals (total: 43%). In contrast, MSW in OECD countries consisted mainly of recyclables (43%) followed by organics (37%). This article also reviewed the other functional elements of the waste management systems in SIDS. Several shortcomings were noted in the process of waste collection, transfer and transport namely the fact of having outdated collection vehicles and narrow roads which are inaccessible. Among the waste management practices in SIDS, waste disposal via landfilling, illegal dumping and backyard burning were favoured most of the time at the expense of sustainable waste treatment technologies such as composting, anaerobic digestion and recycling.

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DOE OSTI2020

Panel Session 45: Savannah River National Laboratory Advancing the EM Mission Through Technology

This panel focused on Savannah River National Laboratory (SRNL) activities and addressed: 1) Initial development work that SRNL performed to develop the technologies and processes, as well as design the equipment for the DOE EM HLW tank waste treatment and soil and groundwater mission, 2) Implemented technology and process improvements to the EM systems, and 3) Current technologies being evaluated for future implementation to improve throughput and reduce cost/schedule. Panelists with presentations: Advancing the EM Mission: SRNL Technology Approach (Connie Herman); Nuclear Materials Management (William Bates); Adjust, Adapt, Attest, and Adjudicate (Bruce Wiersma); SRNL Research and Development for Technological Advancements in the Liquid Waste Mission at SRS (Frank Pennebaker); EM TD Applies Solutions at SRS (Andrew Fellinger); Remediation of Complex Groundwater Plumes: Three Decades of Cleanup Success (Carol Eddy-Dilek)

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DOE OSTI2012

Treatment of Radioactive Metallic Waste from Operation of Nuclear Power Plants by Melting - The German Way for a Consistent Recycling to Minimize the Quantity of Radioactive Waste from Operation and Dismantling for Disposal - 12016

During maintenance of nuclear power plants, and during their decommissioning period, a large quantity of radioactive metallic waste will accrue. On the other hand the capacity for final disposal of radioactive waste in Germany is limited as well as that in the US. That is why all procedures related to this topic should be handled with a maximum of efficiency. The German model of consistent recycling of the radioactive metal scrap within the nuclear industry therefore also offers high capabilities for facilities in the US. The paper gives a compact overview of the impressive results of melting treatment, the current potential and further developments. Thousands of cubic metres of final disposal capacity have been saved. The highest level of efficiency and safety by combining general surface decontamination by blasting and nuclide specific decontamination by melting associated with the typical effects of homogenization. An established process - nationally and internationally recognized. Excellent connection between economy and ecology. (authors)

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DOE OSTI2023

Thermal Characterization of Acid Treated Anion Exchange Resins

Anion exchange is a chemical separation and purification technique in which a solid phase ion exchanging material (i.e., anion exchange resin beads) interchanges its anions with the desired anions from a solution phase. Typical anion exchange resins (e.g., Bio-Rad AG 1-X8 strongly basic anion exchange resin) consist of a polymer resin bead of cross-linked polystyrene with quaternary ammonium functional groups (Figure 1). Anion exchange occurs at the resin functional groups by exchange of the counter ion of the quaternary ammonium (typically chloride or nitrate) for the anionic species of interest. Other resin polymers have been developed—such as cross-linked vinylpyridine/divinylbenzene utilized in Reillex HPQ anion exchange resin—to improve the resins’ resistance against degradation by oxidizing agents, strong acids, and radiation. Anion exchange is performed for the separation of transuranic (TRU) elements throughout the Department of Energy complex. At the Los Alamos National Laboratory (LANL), production scale quantities of these resins are handled at the Chemistry and Metallurgy Research Facility, the Plutonium Facility, and the Transuranic Waste Facility. Spent anion exchange resin will eventually be disposed of as TRU waste. This has prompted concerns regarding its safe disposal under potential hazard scenarios, in particular a thermal excursion of a TRU waste drum. There is a concern that a potential thermal excursion of a TRU waste drum containing anion exchange resin previously contacted with nitric acid may result in energetic side reactions and pressure buildup due to resin degradation by nitric acid and heat. Therefore, the objective of the experiments described in this report was to gather qualitative and quantitative data to support decisions regarding the thermal stability and safe disposal strategy of nitric acid treated anion exchange resins utilized in TRU processing operations.

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DOE OSTI2019

Canadian Nuclear Laboratories Waste Management: Challenges and the Integrated Waste Strategy - 19160

Canadian Nuclear Laboratories (CNL) currently faces various waste management challenges and CNL's Integrated Waste Strategy is adapting to manage them. As Canada's premier nuclear science and technology organization, CNL is a global leader in developing peaceful and innovative applications from nuclear technology through its expertise in the scientific field, including physics, metallurgy, chemistry, biology and engineering. As demands change, CNL is shifting its focus to alpha therapies for cancer treatments and sustainable energy solutions. The next decade will see CNL decommission the sites it manages across Canada, whilst transforming Chalk River Laboratories (CRL) into a new, world-class facility which will have the capability to demonstrate the commercial viability of advanced reactor designs including Small Modular Reactors (SMRs). This is being enabled through an accelerated decommissioning program for all CNL operated sites, while renovating CRL, which includes revitalizing essential site infrastructure, and investment in new, world-class science facilities in a modern campus environment. With decommissioning and revitalization occurring simultaneously, the long-term strategy and vision for CNL presents some unique challenges to waste management. Therefore, integrated, collaborative and coordinated waste management is a vital component to ensuring a successful future. The Integrated Waste Strategy coordinates, and provides alignment of, waste management across CNL and: - Includes all CNL operated (Atomic Energy of Canada Limited (AECL) owned) sites; - Considers all CNL managed waste (including radioactive and non-radioactive operational, decommissioning, legacy and historic waste) for the planned lifetime of the sites; - Defines waste routes using a life cycle approach (from the generation of waste to final disposal) A baseline Integrated Waste Strategy and associated action plan to mitigate identified gaps and issues have been produced. A program of work is underway to close those gaps identified in a consistent, transparent, prioritized and justified manner. One of the key challenges is to ensure that waste management planning is completed effectively and in a timely manner. This will enable the provision of suitable and sufficient waste capability to meet CNL's long-term strategy. The Integrated Waste Strategy has been developed and maintained through engagement at all levels across CNL, ensuring buy-in of proposed solutions. Key process stakeholders include generators of CNL managed waste to ensure integrated consideration of legacy, current and future waste generation; waste management personnel to identify requirements, including processing, storage, transport and disposal; AECL as the client and waste liability owner; and the public to ensure they understand the 'big picture' regarding waste strategy. (author)

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DOE OSTI2017

Evaluation of quartz melt rate furnace with the nitric-glycolic flowsheet

The Savannah River National Laboratory (SRNL) was tasked to support validation of the Defense Waste Processing Facility (DWPF) melter offgas flammability model for the Nitric-Glycolic (NG) flowsheet. The work is supplemental to the Cold Cap Evaluation Furnace (CEF) testing conducted in 20141 and the Slurry-fed Melt Rate Furnace (SMRF) testing conducted in 20162 that supported Deliverable 4 of the DWPF & Saltstone Facility Engineering Technical Task Request (TTR).3 The Quartz Melt Rate Furnace (QMRF) was evaluated as a bench-scale scoping tool to potentially be used in lieu of or simply prior to the use of the larger-scale SMRF or CEF. The QMRF platform has been used previously to evaluate melt rate behavior and offgas compositions of DWPF glasses prepared from the Nitric-Formic (NF) flowsheet but not for the NG flowsheet and not with continuous feeding.4 The overall objective of the 2016-2017 testing was to evaluate the efficacy of the QMRF as a lab-scale platform for steady state, continuously fed melter testing with the NG flowsheet as an alternative to more expensive and complex testing with the SMRF or CEF platforms.

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DOE OSTI2015

Laser size reduction of contaminated Magnox Pond skips - 15166

The Magnox Pond Program Team at Hinkley Point A Power Station have successfully demonstrated the use of an automated Computer Numeric Controlled (CNC) dry milling process to efficiently decontaminate radioactive pond skips that were once used to store and transport Magnox spent fuel assemblies between reactor sites and Sellafield for reprocessing. This decontamination process has ensured ILW skips meet metal melt recycling criteria, and even Out of Scope levels, while at the same time producing dry swarf that is easily disposed of as ILW. However, the dry milling method requires the skips to be size reduced to a very simple flat geometry to simplify the milling operation. Several size reduction techniques (mechanical and thermal) have been evaluated. Based on the advantages of remote handling, high cutting speed, flexibility with complex geometries, reduced secondary waste production and significantly reduced worker exposure, robotically operated fibre delivered laser cutting was chosen. This paper describes the development, installation and operation of an automated laser cutting system commissioned at Hinkley Point A to size reduce contaminated skips in to five flat sections. It was demonstrated that a complete skip can be remotely size reduced in less than 60 minutes. This successful demonstration, coupled with CNC milling, has the potential to save an estimated British Pounds 30 million across the Magnox sites alone, with much higher saving potentials for skips stored at Sellafield, enabling this new size reduction process to meet cost reduction objectives of the UK's Nuclear Decommissioning Authority. (authors)

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

Results for the October 2023 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the October 2023 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the SWPF DSS sample was obtained. The October 2023 Semiannual sample of the SWPF DSS is a composite from the six months of SWPF processing during the Third Quarter Fiscal Year 2023 (3QFY2023) and the Fourth Quarter Fiscal Year 2023 (4QFY2023).

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

Decontamination of organic wastes containing radionuclides - 15122

The Arvia Process is being adapted for use in the nuclear industry for the decontamination of oils containing radionuclides. This waste treatment process uses adsorption followed by electrochemical oxidation as an alternative technique to conventional incineration. A study has been carried out on components of the Arvia Process with the aim of understanding the changes in chemistry due to ionizing radiation. Samples of adsorbent in water have been exposed to gamma-rays from a Co-60 irradiator at absorbed doses of 3, 19, and 304 kGy. Preliminary results show that the concentration of inorganic ions and carbon in the water increased as a function of dose. This increase suggests that degradation of the adsorbent may take place at high doses. Further characterization of the microstructure and adsorption studies post-irradiation are planned. (authors)

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