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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Radioactive Waste Management in a Treatability and Optimisation Perspective - 17527

In order to meet recycling criteria or to fulfil Waste Acceptance Criteria (WAC) for disposal there are many different management concepts for radioactive waste that can be applied. Which concept or set of concepts to use to optimize waste management depends on several factors, among them, the regulatory framework for recycling, the availability of a disposal facility and the acceptance criteria for disposal. The required pre-treatment and other pre-disposal activities should be carefully analyzed for each of the disposal options, as they are significant contributors to the total cost for managing the waste. Waste Acceptance Criteria are living documents that should be updated when regulatory framework changes as well as when new treatment methods and new packaging and barrier systems develop. Each waste treatment facility, has its own specific limitations and therefore its own WAC/CFA (conditions for acceptance), which depends both on the plant design and national legislation. Many waste owners, countries or licensees, have their preferred, and in some cases well established, waste routes for certain waste streams. In some countries, the full chain from waste treatment through to disposal of radioactive waste is in place, whilst other countries still are in the process of establishing their repository program. The worst case scenario for any licensee or waste owner is waste which is not possible to qualify for disposal. This can be due to poor or degraded waste containers, unspecified content in waste packages, or wastes that are not suitable for disposal in their current form. Such deficiencies result in expensive actions related to waste retrieval, new characterisation and/or costly investigations. The EDF Group has a fleet of waste treatment facilities and has, for decades, provided waste treatment services to domestic and international customers. This paper will discuss important waste owner considerations and EDF experiences related to the optimization of the waste management process. This will be from both a treatment and an end-state perspective. (authors)

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DOE OSTI2016

Final Closure of the Maxey Flats Low-Level Radioactive Waste Disposal Site - 16088

This paper describes the final closure remedy for the former Maxey Flats Low-Level Radioactive Waste Disposal Site (MFDS), the challenges presented at the site due to the condition of the disposal facility and waste, and their resolution through comprehensive geo-synthetic design. The MFDS is an inactive landfill located in Kentucky on a ridge surrounded by a buffer zone. Historically, waste disposal occurred without reducing voids or without compacting the placed waste. Settlement and water management issues led to releases of radioactive contamination into surrounding waters, resulting in the closure of the facility in 1977. The final closure remedy consists of 1) development and implementation oversight of a sump abandonment design to ensure no direct pathways for infiltrating water to contact the waste and 2) development and implementation oversight of the Final Closure Period (FCP) cap design. Challenges successfully addressed by the FCP scope included: Abandon sumps to minimize storm water migrating into the waste; Address potential existing groundwater waste interaction; Maximum use of on-site cap construction materials; Design cap to fit very restricted site footprint; Assess and address potential isolated differential settlement; Maximize passive storm water management; Provide alternative site access during construction. The FCP remedy and resultant design satisfies these challenges through use of layers of geo-grid-reinforced soil and geo-synthetic materials, contours and slopes that minimize the quantities of construction materials while maximizing the use of available material, probabilistic subsidence modeling, an integrated storm water management system ensuring the slow release of storm water into receiving streams, and a dedicated construction haul road system. (authors)

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DOE OSTI2016

Applicability of Wireless Power Transfer for Monitoring Technology of Radioactive Waste Geological Disposal - 16156

To assure the engineered barrier of a high-level radioactive waste repository functions correctly, long-term monitoring inside the backfilled deposition holes and tunnels is planned. Cables are used to power monitoring equipment and send signals however complete isolation will not be possible. Previously developed monitoring systems send information wirelessly and used batteries. Due to the limited battery life, these systems cannot be used for long-term monitoring. In this paper a wireless power transfer method based on magnetic resonant coupling is presented. The necessary transmitter and receiver coils have been developed. The performance of the individual coils has been tested. The performance of the system of transmitter and receiver with both air and bentonite as medium was tested. The results showed that the system with bentonite performed similarly well as the system with air. Wireless power transfer based on magnetic resonant coupling is therefore a feasible method for high-level radioactive waste repository monitoring applications which could in principle work for a period of 100 year or longer. (authors)

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DOE OSTI2016

Developing a New National Programme and Implementation Plan for the Management and Disposal of Spent Fuel and Radioactive Wastes in Romania - 16391

The European 'Waste Directive' (2011/70/Euratom) required each Member State to report by August 2015 on their National Programme for management of spent fuel and radioactive wastes. In Romania, the Nuclear Agency (ANDR) was responsible for preparing the National Programme report, and this was done in a step-wise manner that involved the regulators and waste producers. The starting point for the process was an assessment of existing policy against the evolving situation in the nuclear sector in the country, most notably the decision to progress with the construction of two new CANDU reactors. The final agreed National Programme report submitted to the European Commission describes the activities currently undertaken to manage spent fuel and radioactive wastes in Romania. The report also sets out, at a high-level, the additional activities that will need to be taken in the future (and an indicative timeline), including those steps that will be required to implement a new surface repository for short-lived wastes, and a geological repository for spent fuel and long-lived wastes. Additional, on-going work is intended to lead to a revision of Government policy and the development of a more detailed implementation strategy, underpinned by a programme of R and D and robust financing plans. (authors)

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DOE OSTI2016

Disposal of Greater Than Class C Low-Level Radioactive Waste in Andrews County, Texas - 16408

The U.S. Nuclear Regulatory Commission (NRC) staff has recommended approval of an initiative to permit the Texas Commission on Environmental Quality (TCEQ) to license the disposal of commercial and co-mingled Greater Than Class C Low-Level Radioactive Waste (GTCC LLW) in Andrews County, Texas. Such a framework is clearly articulated in the Section 274 of the Atomic Energy Act of 1954, as amended. The NRC staff also recommended approval to proceed with a rulemaking to align the Waste Classification Tables in Title 10 of the Code of Federal Regulations (CFR), Part 61.55 with the paragraph 61.2 definition of 'waste' to provide a disposal pathway for waste containing certain alpha-emitting transuranic elements with concentrations exceeding 100 nanocuries per gram (nCi/g). A pathway for the disposal of commercial and co-mingled GTCC LLW, including waste with certain alpha emitting transuranic elements exceeding 100 nCi/g (TRU waste), might emerge if the initiative described in the Commissions' July 22, 2015 document, SECY-15-0094, Historical and Current Issues Related to Disposal of GTCC LLW, is approved. The NRC staff recommendations envisioned the NRC and TCEQ working together to review a site-specific performance assessment that would be submitted by Waste Control Specialists LLC (WCS) as part of an amendment to its existing Radioactive Materials License No. R04100. The TCEQ's recommendations were prompted by a petition for rulemaking filed by WCS to allow the disposal of GTCC waste in Texas. On September 10, 2014, the TCEQ Commissioners directed their staff to engage stakeholders and request input from the NRC and the U.S. Department of Energy (DOE) regarding actions needed to proceed with any future rulemaking. On January 30, 2015, the TCEQ requested clarification from the NRC regarding their legal authority and jurisdiction to regulate GTCC, GTCC-like, and TRU waste. The deliberations that have transpired over the past year have raised important legal, policy, and technical matters. This initiative could provide a safe disposal pathway for certain waste streams that have been orphaned over the past 30 years. One important realization that has emerged is the conservative assumptions used to establish the upper limits for Class C LLW when Part 61 was promulgated in 1981 would be much too restrictive when applied to a modern, highly engineered disposal facility located in arid west Texas. Waste that was not suitable for near surface disposal in 1981 could certainly be demonstrated to be safely disposed at the WCS Federal Waste Disposal Facility in Andrews County, Texas. This paper will address the legal, policy, and technical matters supporting this important topic currently under consideration both nationally and within the State of Texas. (authors)

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DOE OSTI2016

Development and Application of the Stakeholder Tool for Assessing Radioactive Transportation (START) - 16516

The U.S. Department of Energy, Office of Nuclear Energy is laying the groundwork for implementing interim storage of spent nuclear fuel (SNF) pursuant to the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste. This effort includes planning to address technical and logistical aspects of transporting SNF to an interim storage facility. A key project component is the development and utilization of a decision-support tool to evaluate transportation routing options and emergency preparedness. START (Stakeholder Tool for Assessing Radioactive Transportation) is a web-based application that utilizes geographic information systems (GIS) technology to represent transportation network operations as well as proximate features, such as tribal lands, emergency response capability, schools and environmentally-sensitive areas. Modes and routes between shipment origins and destinations designated by the user can be evaluated according to multiple routing criteria, and users can impose constraints that require the route to pass through or avoid specified locations. Considerable information is produced describing route efficiency, safety, security and environmental exposure that enable stakeholders to understand the implications of alternative route options. This paper describes START's capabilities and illustrates its use. (authors)

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DOE OSTI2018

Testing and Deployment of Remote Systems to Remove Highly Radioactive Soil at Hanford's 324 Building - 18043

U.S. Department of Energy (DOE) contractor CH2M HILL Plateau Remediation Company (CHPRC) is in the process of procuring, testing and deploying remotely operated tools to safely remove highly contaminated soil from beneath the 324 Building on the Hanford Site in southeastern Washington State. The 324 Building operated for more than 40 years, completing a variety of high-dose engineering development and demonstration activities inside six hot cells that shielded the workers from the highly radioactive materials. Activities to decommission, decontaminate and demolish the facility began in 2008. In 2010, soil contamination was found in the soil under one of the hot cells, and building deactivation was paused to evaluate the soil contamination. Sampling efforts, using borehole technology, revealed significant under-building contamination with dose rates measuring has high as 130 Sv/hr (13,000 R/hr). The highly-contaminated soil has to be removed before traditional open air deactivation activities can continue. CHPRC completed proof of concept testing and built a full-scale test facility to demonstrate the effectiveness of three primary tools that will accomplish the floor and soil removal. The tools include a commercially available concrete saw, and a commercially available excavator arm that have been modified to operate remotely in the high-radiation environment within the hot-cell. In order to reduce the wear on the hot-cell overhead cranes, a remote operated rail and cart system has been designed to transport soil and debris out of the hot-cell to an airlock area where the waste will be separated, according to its level of radioactivity. A camera and light system has been designed to provide visual and audio coverage of the hot-cell area that will allow the operations to proceed from remote control stations. This paper will describe the primary remote operated equipment designed for the project and the challenges and processes for developing, testing and deploying the equipment necessary to excavate the soil from below the 324 Building. (authors)

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DOE OSTI2018

The Evolution of Hot-Isostatic Pressing for the Treatment of Radioactive Wastes - 18276

Hot-isostatic Pressing (HIPing) was first developed by Battelle in the 1950's for applications in the nuclear program. The use of HIPing as a means of treating radioactive wastes was first proposed and patented in Sweden in the 1970's and subsequently gained traction in the US, where throughout the 1980's and 90's work was undertaken at Rockwell, Lawrence Livermore National Laboratory (LLNL), and Idaho National Laboratory (INL) on waste-forms for US Defense wastes. This development was coupled with front-end processing options leading to conceptual plant designs which included hot-cell layouts and costings. Industrial maturation of hot-isostatic press (HIP) technology over the last 20+ years has seen it increasingly become a candidate for treating a variety of waste streams. In this paper, we discuss HIP concepts for treating radioactive wastes, present some recent work undertaken by the authors to demonstrate the feasibility of HIPing for sludge wastes and ion exchangers, and discuss the options and technology challenges for implementation. (authors)

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DOE OSTI2018

Buffer Material Degradation in High Level Radioactive Waste Disposal - 18454

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

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DOE OSTI2018

Immobilization of Radioactive Materials Using Porous Iron Composite Media - 18582

High levels of nitrate (NO{sub 3}) derived from agriculture and naturally occurring radioactive materials (NORM), primarily uranium (U) and natural decay chain products that have been mobilized by the influx of oxygenated, high-alkalinity water, are a threat to shallow groundwater resources in California. Porous Iron Composite (PIC) materials were evaluated as a water treatment option in a series of batch experiments. While the use of Zero Valent Iron (ZVI) to address groundwater contaminated with NO{sub 3} or U has been demonstrated, the ability of PICs to chemically reduce and immobilize U and other associated contaminants in oxygenated alkaline groundwater where soluble uranyl-carbonate species dominate and high NO{sub 3} levels may compete for reductive capacity remain a concern. Additionally, ZVI materials have been used to immobilize other redox sensitive radionuclides (e.g., technetium-99 ({sup 99}Tc)) with limited success. Batch tests were conducted using an artificial groundwater surrogate (i.e., CAGW for California Groundwater) containing ∼ 100 mg L{sup -1} NO{sub 3}{sup -}, 200 mg L{sup -1} alkalinity, 100 μg L{sup -1} U, 100 μg L{sup -1} rhenium (Re) as a non-radioactive surrogate for the nuclear fission product technetium-99 ({sup 99}Tc), 100 μg L{sup -1} stable strontium (Sr) as a surrogate for radium (Ra) and the fission product strontium-90 ({sup 90}Sr), and 50 μg L{sup -1} arsenic (As), a common groundwater contaminant. Five gram samples of the PIC material were reacted with 200 mL of CAGW under two different atmospheric conditions to evaluate the impact of O{sub 2}. Batch tests were also conducted in the presence and absence of another oxidized species, NO{sub 3}. At specific time intervals, a small aliquot was collected from each batch treatment for chemical analysis (i.e., U, Re, As, Fe, NO{sub 3}, etc.) and the pH and ORP of the remaining suspension was measured. The PIC was effective at fully reducing NO{sub 3}{sup -} levels over the extended reaction time, even though the reactor pH quickly increased above 8.0 during equilibration. Soluble levels of U and Re decreased in a similar fashion to NO{sub 3}{sup -} in the presence and absence of O{sub 2}. Soluble As levels decreased more rapidly than other contaminants, but remained at ∼ 1 μg L{sup -1}. Very little nitrite (NO{sub 2}{sup -}) was detected in the batch tests, with NH{sub 4}{sup +} levels consistent with the amount of NO{sub 3}{sup -} reduced. Similar rates of immobilization for U, Re, and As were observed in the presence and absence of NO{sub 3}{sup -}. (authors)

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DOE OSTI2019

Development of the B-25 Low Level Radioactive Waste Container in Taiwan - 1900x

Chinshan Nuclear Power Plant will be permanently stopped its operation in the middle of 2019 and will be the first nuclear power plant stepping into the decommissioning phase in Taiwan. According to Nuclear Reactor Facilities Regulation Act, a nuclear facility must be dismantled in 25 years after being permanently ceased operation. In its decommissioning plan, the B -25 low level radioactive waste container was one of the proposed containers used for this decommissioning task. However, the B-25 waste container was not used or developed in Taiwan before. To meet this need, Institute of Nuclear Energy Research developed a new type of B-25 waste container and will apply to the authority for its usage license. This article provided the information of the low level radioactive waste containers have been approved in Taiwan and described the design specifications of the developed B-25 waste container. Besides, the numerical evaluation and test results were also included in this article. Based on the evaluation and test results, the developed B-25 waste container satisfied the requirements of the storage container and industrial package type 2. It is believed that the usage license would be granted soon after applying it to the authority. (authors)

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DOE OSTI2019

Panel Session 84: Challenges and Lessons Learned in Radioactive Waste Packaging, Transportation, and Disposal

This panel focused on challenges and lessons learned in radioactive waste packaging, transportation, compliance and disposal. Participants provided their perspectives and suggestions on how to increase efficiencies through eliminating inconsistencies between specifications and other requirements to provide products at a reasonable cost and enhance communication to enable better service. Presentations proposed include: 1 - Low Level Waste Type (B) Cask Shipment Challenges (Aleksandr Gelfond, EnergySolutions (USA)); 2 - Radioactive Packaging Challenges (Jeff England, NAC International (USA)); 3 - Challenges and Lessons Learned in Radwaste Packaging, Transportation, Compliance and Receipt (Mary Magleby, URS - CH2M Oak Ridge LLC (USA))

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DOE OSTI2020

Update on the Status of Deep Borehole Disposal of High-Level Radioactive Waste in Germany - 20285

The phase-out of nuclear energy in Germany will take place in 2022. A site for final disposal of high level radioactive waste (HLRW) has not yet been found, but a site selection process was restarted by Act on the Search for and Selection of a Site for a Disposal Facility for High-Level Radioactive Waste (Site Selection Act - StandAG 2017). This act was based on a recommendation by a commission which also advised to follow up the development of deep borehole disposal (DBD) as an alternative option for final disposal of HLRW. This paper summarizes briefly the status of DBD in Germany and if this option should be pursued in Germany. Although there are some merits of DBD, it can only be a real option if research and development is supported. The technical equipment for boreholes of the required size will only be developed if there is funding and a feasibility test. Furthermore, any concept of DBD and technology must be detailed further, and some requirements of the act must be reconsidered. The paper concludes that despite the possible merits of DBD, the political and financial support for R and D will only be provided for the time being if DBD is pushed by interested parties. Alternatively, if the site selection procedure is not progressing well, then the alternative option of DBD may find greater interest in the future. However, its availability will be severely limited due to the missing development of detailed concepts and its demonstration of feasibility. (authors)

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DOE OSTI2016

Hydraulic Conductivity of Geosynthetic Clay Liners to Low-Level Radioactive Waste Leachate

Hydraulic conductivity was evaluated for eight commercially available geosynthetic clay liners (GCLs) permeated with leachate characteristic of low-level radioactive waste (LLW) disposal facilities operated by the U.S. Department of Energy (DOE). Two of the GCLs (CS and GS) contained conventional sodium bentonite (Na-B). The others contained a bentonite–polymer mixture (CPL, CPH, GPL1, GPL2, and GPH) or bentonite–polymer composite (BPC). All GCLs (except GPL2 and GPH) were permeated directly with two synthetic LLW leachates that are essentially identical, except one has no radionuclides (nonradioactive synthetic leachate, or NSL) and the other has radionuclides (radioactive synthetic leachate, or RSL). Hydraulic conductivities to RSL and NSL were identical. For the CS and GS GCLs, the hydraulic conductivity gradually increased by a factor of 5–25 because divalent cations in the leachate replaced native sodium cations bound to the bentonite. The CPL, GPL1, and GPL2 GCLs with low polymer loading (1.2–3.3%) had hydraulic conductivities similar to the conventional GCLs. In contrast, hydraulic conductivity of the CPH, GPH, and BPC GCLs with high polymer loading (≥5%) to RSL or NSL was comparable to, or lower than, the hydraulic conductivity to deionized water. Permeation with leachate reduced the swell index of the bentonite in all of the GCLs. A conceptual model featuring pore blocking by polymer hydrogel is proposed to explain why the hydraulic conductivity of bentonite–polymer GCLs to LLW leachates remains low even though the leachate inhibits bentonite swelling.

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DOE OSTI2017

Antioxidant Depletion and Service Life Prediction for HDPE Geomembranes Exposed to Low-Level Radioactive Waste Leachate

Antioxidant depletion was evaluated in high-density polyethylene (HDPE) geomembrane (GM) coupons immersed in radioactive synthetic leachate (RSL) representative of leachate in low-level radioactive waste (LLW) disposal facilities operated by the U.S. Department of Energy’s environmental restoration programs. Depletion was assessed at four temperatures (25, 50, 70, and 90°C). Comparative tests were conducted with nonradioactive synthetic leachate (NSL) having the same chemistry as RSL, except radionuclides were excluded. Control tests were conducted with deionized (DI) water. One-sided exposure tests were also conducted with the HDPE GM in RSL at 70 and 90°C. Specimens were removed periodically and tested to determine antioxidant depletion, melt flow index (MFI), crystallinity, and stress crack resistance. Standard and high-pressure oxidative induction time (OIT) tests were conducted to measure antioxidant depletion. Antioxidant depletion rates in RSL were only slightly greater than those in NSL, indicating that radionuclides in the LLW leachate had only minor effect on antioxidant depletion. Antioxidant depletion in RSL and NSL occurred 1.1–1.9 times faster than in DI water. Antioxidant depletion in HDPE GMs exposed to RSL and NSL at 90°C was complete after 12 months. MFI decreased systematically after 15 months of immersion, which is indicative of crosslinking due to oxidation. Crystallinity also decreased slightly after 12 months, whereas stress crack resistance (SCR) showed no trend over the immersion period. Arrhenius modeling was used to extrapolate the laboratory data at elevated temperatures to a typical LLWliner at 15°C. Predicted time for antioxidant depletion for a 2-mm-thick HDPE GM in a composite liner in contact with LLW leachate is estimated to be 730 years. The total service life of HDPE GMs in contact with LLWleachate is estimated to be at least 1975 years.

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DOE OSTI2017

Unreviewed Disposal Question Evaluation: Disposal of the Lawrence Livermore National Laboratory EnergyX Macroencapsulation Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Lawrence Livermore National Laboratory (LLNL) EnergyX Macroencapsulation mixed low-level waste (MLLW) (BCLAMACRONCAP, Revision 2 [LLNL 2017]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the LLNL EnergyX Macroencapsulation MLLW meets all U.S. Department of Energy (DOE) Manual DOE M 435.1-1, Radioactive Waste Management Manual, Chapter IV, Section P performance objectives (DOE 1999). The LLNL EnergyX Macroencapsulation waste stream is recommended for acceptance without conditions.

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DOE OSTI2018

FY 2017 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy Standard Disposal Authorization Statement and Tank Closure Documentation (DOE-STD-5002-2017) requires preparation of an Annual Summary Report comparing low-level radioactive waste disposal facilities operations with the Disposal Authorization Statement (DAS) and its supporting technical documents including the performance assessment (PA) and the composite analysis (CA). The fiscal year (FY) 2017 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) was prepared by reviewing planned and discovered changes, waste receipts, monitoring results, research and development (R&D) results, and the status of DAS conditions and issues to assess the validity of the DASs, PAs, and CAs.

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DOE OSTI2019

Unreviewed Disposal Question Evaluation: Disposal of the Idaho Nuclear Technology and Engineering Center Mixed Low-Level Waste Requiring Macroencapsulation at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation (UDQE) assesses whether the Idaho National Laboratory (INL), Idaho Nuclear Technology and Engineering Center (INTEC) Mixed Low-Level Waste (MLLW) requiring macroencapsulation (INEL166322NR1, Revision 0 [INL 2019]) is suitable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). Disposal of the INTEC MLLW requiring macroencapsulation waste stream meets all the performance objectives contained in U.S. Department of Energy (DOE) Manual DOE M 435.1-1, “Radioactive Waste Management Manual,” Chapter IV, Section P (DOE 2011). The INTEC MLLW requiring macroencapsulation waste stream is recommended for acceptance without conditions.

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