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At least 433 records · Page 24
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DOE OSTI2017

Ethical Considerations in Determining the Societal Investment in Warning Future Generations about Latent Long-term Risks from Radioactive Waste Disposal Facilities - 17058

Writers and philosophers have long romanticized and pondered the great challenge of communicating with future generations. The only way that man can travel through time. Never has this notion been more studied and debated than in the nuclear waste disposal arena. Regulatory requirements in the United States and internationally, require that there be some form of notification of future generations in place once a deep geologic nuclear waste repository has been closed. It is considered a generational responsibility. Since the difficulty of this task is so daunting, it has been assumed that the monetary cost to accomplish this millennia-spanning message-in-a-bottle must also be extreme. This paper will discuss both the ethical and equitable implications across generations, as they pertain to the reality of future inadvertent intrusion scenarios and the associated cost of communicating through time to prevent them. (authors)

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DOE OSTI2018

Understanding Radionuclide Interactions with Cementitious Materials for Radioactive Waste Management - 18298

In the UK, ILW is currently managed using cementitious encapsulation in Portland cement (PC)-based grouts. The long-term behaviour of radionuclides in ILW cementitious waste-forms under varying local conditions of extended interim storage and geological disposal is not well understood and the behaviour of cement minerals in the retention of various radionuclides is poorly constrained. This work focuses on the Ca-Si-[Al,Mg,S] mineral system and how this may evolve with the introduction of radionuclides into the system. Here, we discuss the experimental program aimed at underpinning the uptake and retention mechanisms of a range of key radionuclides (U, Tc, Ce/Pu), with selected cement phases: calcium-silicate-hydrate (C-S-H), hydrotalcite (Mg{sub 6}Al{sub 2}CO{sub 3}(OH)16.4H{sub 2}O) and ettringite (Ca{sub 6}(Al{sub 2}O{sub 6})(SO{sub 4}){sub 3}.32H{sub 2}O). Described here is the syntheses of C-S-H and radionuclide incorporation experiments performed on this phase. Further to this, thermodynamic geochemical modelling has been adopted to explore the Ca-Si-U mineral system, aimed at identifying the mineral phases likely to form on contact of C-S-H phases with aqueous uranium, as a function of pH. (authors)

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DOE OSTI2013

Concrete in engineered barriers for radioactive waste disposal facilities - phenomenological study and assessment of long term performance

The paper aims to highlight recent developments at the Belgian Nuclear Research Centre SCK-CEN in experimental and numerical study of the coupled physical-chemical behaviour of concrete subject to chemical degradation. The discussion mainly focusses on three inter-linked research projects covering novel experimental methods to study the alteration of hydraulic and transport properties during carbonation and calcium leaching, a pore scale numerical model to capture microstructural changes due to the above degradation processes and a generic multiscale model aimed at determining evolution of the properties of a macrostructure over the long term. The paper also describes supplementary continuum scale numerical studies concerning concrete-clay interactions and geochemical impact on the physical structure of concrete. Preliminary findings from these studies show encouraging results such as the development of novel leaching, water permeability and diffusion apparatus, a robust pore scale model based on Lattice-Boltzmann method and a mesoscale study focused on the importance of interfacial transition zones on the effective diffusivity for linear and nonlinear diffusion problems. (authors)

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DOE OSTI2020

Panel Session 24B: Records, Knowledge, and Memory for Radioactive Waste Repositories: Generational Equity Focus

This panel focused on the latest thoughts, ideas, and methodologies being explored throughout the world on how to communicate with future generations regarding nuclear waste disposal. Scientists determined many years ago that geological disposal in a repository was the preferred solution for nuclear waste disposal given the longevity concerns of the waste(s). Future generations must be informed through records, memory keeping and permanent markers to ensure they are aware of, and knowledgeable of, the dangers associated with nuclear waste isolated from the biosphere. Panelists with presentations: You Want to Drill Where? Human Intrusion Messaging Considerations (Thomas Peake, Jonathan Major); Ethical Reflections on the Basic Reasons for RK and M Measures (Carl-Reinhold Brakenhielm); NEA Activities on Information, Data and Knowledge Management (Rebecca Tadesse); Records, Knowledge and Memory (RK and M) Across Generations: Recent Activities and Progress in Sweden (Claudio Pescatore)

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DOE OSTI2019

Annual Radioactive Waste Tank Inspection Program - 2018

The Department of Energy (DOE) requires that Savannah River Remediation LLC (SRR), the Liquid Waste contractor, maintain a robust and timely Structural Integrity Program for waste tanks that are leak-tight and for waste tanks that have leaked or are suspect. These requirements are promulgated in DOE Manual 435.1-1, Section II.Q.(2). A key aspect of SRR's implementation of these requirements is its annual tank inspection program. In 2018, the SRR Visual and In-Service Inspection Program used two visual imagery techniques to ascertain the integrity of the Savannah River Site (SRS) waste tanks: high resolution digital photography and closed-circuit television. The primary inspection method made use of direct photograph techniques, e.g., making a series of photographs providing detailed views of the tank and wide-angle photography for obtaining overviews of large areas. Closed-circuit television was then typically used to further investigate conditions found during the initial inspection. During 2018, a total of 4,853 photographs were taken and 47 video inspections were performed as part of this inspection program. Of the 51 waste tank systems at SRS that were designed, constructed, and placed into active service storing radioactive liquid waste, 43 remain in active service. Eight tanks have been cleaned, stabilized by filling with engineered grout, and removed from service (also known as operationally closed) -Tanks 5, 6, and 12 (Type I tanks); Tank 16 (Type II tank); and Tanks 17, 18, 19, and 20 (Type IV tanks). For the 43 waste tanks, inspections were successfully conducted in 2018 through all accessible annulus risers and inspection ports for nine Type I tanks, three Type II tanks, and 27 Type III/IIIA tanks; at least one inspection was made in the interior of four Type IV tanks (i.e., the tanks without annular spaces) above the liquid level in the tank. Rainwater continues periodically to be observed migrating into the annuli of some tanks. Evidence of this infiltration was primarily noted by observation of surface stains, and occasionally by calciferous deposits, changed configuration of salt deposits in the associated annulus, and mild surface corrosion. The conditions of the 43 tanks remained essentially unchanged from the conditions reported in 2017. The one key exception was the identification of eight new cracks in the Tank 15 primary tank wall. These newly identified cracks are located at 65, 67, 70, 76, 150, 152, 176 inches and 200 inches above the tank bottom. The identified cracks in Tank 15 do not constitute an impact on the structural integrity of the tank, as evaluated and confirmed in the referenced report, SRNS-STI-2008-00028, A Structural Impact Assessment of Flaws Detected During Ultrasonic Examination of Tank 15, and calculation, T-CLC-H-00890, Calculation of Critical Crack Lengths in HLW Tank 15. The 2018 inspection program confirmed the structural integrity and waste confinement capability of all 43 waste tanks.

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DOE OSTI2016

Hydrogeologic Lessons Learned from Deep Drilling Projects Applied to Deep Borehole Disposal of Radioactive Wastes - 16133

Hydrogeological results from the Kola SG-3, Gravberg-1, Cajon Pass, and KTB-VB and KTB-HB boreholes are applied to the US deep borehole disposal concept to assess whether old, saline, and reducing deep groundwater, in low bulk permeability crystalline basement rocks, could be found in a DOE proposed deep borehole field test. Each of the drilling projects faced characterization challenges, including obtaining sufficient information to characterize key parameters and their uncertainty, equipment limitations and failures, and obtaining samples that are representative of in situ conditions while maintaining sample integrity. Despite these characterization challenges, each project obtained sufficient data to directly measure or infer, for redox state, conditions at the 3 to 5 km (1.9 - 3.1 miles) target disposal depth of the US deep borehole disposal concept. The hydrogeologic data suggest that old, saline, and reducing deep groundwater could be found at a test site located in geologically stable crystalline basement. Although the DOE deep borehole testing program has not defined the magnitude of an acceptably low bulk permeability, bulk permeability results indicate that values used in current US deep borehole modeling efforts may be too low by an order of magnitude or more, as compared with in situ conditions. (authors)

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DOE OSTI2016

Migration Rates of Some Cations in Unsaturated Layer around the Near-Surface Disposal Facilities of Radioactive Wastes - 16116

In the porous layer around the near-surface underground, the intermittent recharge process of rainwater forms a layer perfectly not saturated with liquid phase. Here, such a layer is simply called 'unsaturated layer'. In this study, the retardation effects of radionuclide migration were examined under an unsaturated condition by using a column packed with silica sand particles. As tracer ions, sodium (Na), europium (Eu) and cesium (Cs) were selected for estimating the retardation effects under both the saturated and unsaturated condition. Particularly, Cs is one of the key elements included in the contaminated soil due to Fukushima Daiichi Nuclear Power Plant Accident. These tracer ions were injected into column with the condition of constant flow rate or constant pressure flow. To quantify the influence of the unsaturated condition on the migration of radionuclides, retardation coefficient (R{sub d}) and Peclet number (P{sub e}) were estimated by applying advection diffusion equation to the experimental results. As a result, the conflicting effects on the retardation effect in the unsaturated layer were suggested; one is the decrease in the retardation effect by interrupting the sorption of ions with gas phase, the other is the increase in the retardation effect by the clogging with gas phase in flow paths. For the assessment of the radionuclide migration in the unsaturated layer, it is important to clarify such complicated behaviors with considering key parameters related to the migration of ions. (authors)

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DOE OSTI2021

Modeling and Analysis of the Transport and Disposal of Beryllium Moderator Blocks and Greater than Class C (GTCC) Waste

Radioactive waste in the United states is categorized based on its source, radioactivity, and security risk. The categorization method employed by the Nuclear Regulatory Committee (NRC) for low-level radioactive waste is not currently applied to any waste generated by the Department of Energy (DOE) or disposed of in DOE facilities. As a result, there is a large volume of DOE-generated waste with characteristics similar to the NRC’s “Greater than Class C” (GTCC) waste category which presently do not have a path for long-term disposal. Much of this waste cannot undergo the standard commercial disposition process due to it being categorized as Transuranic (TRU) waste under DOE guidelines, due to elevated concentrations of key fission products. Possible solutions to this dilemma are explored in the Environmental Impact Statement for the “Disposal of Greater-Than-Class-C Low Level Radioactive Wave” (EIS-0375). This paper analyzes several EIS possible paths for disposal for this “orphaned” waste. One example that highlights this categorization issue is the spent beryllium cladding that has been extracted from the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) throughout its operational lifetime. These beryllium blocks surround the reactor’s main chambers and components and serve as neutron moderators. This paper will use this particular waste form to analyze the economic and technological viability of transporting and storing this material under the storage and transportation criteria of the Waste isolation Pilot Plant (WIPP), which was deemed the most feasible disposition path according to the EIS. The example of this waste form used in this paper that highlights this categorization issue is the spent beryllium cladding extracted from Idaho National Laboratory’s (INL) Advanced Test Reactor (ATR). These beryllium blocks surround the reactor’s components and serve as neutron moderators, and thus have accumulated high concentration of high-activity transuranic isotopes. This paper analyzed the viability of transporting and storing this material under the storage and transportation criteria of the five primary disposition paths covered in EIS-0375. An array of calculations was carried out to assess the viability of the various disposition paths for the beryllium shipments as well as other waste shipments that fall within the GTCC category. Modeling with MCNP 6.2 was conducted to determine whether the beryllium blocks could be safely stored and transported within a 72-B cask; the standard shipping container for remote-handled, transuranic waste, while also meeting regulatory limits at various disposition paths. A cost analysis of the transportation and long-term disposal paths mentioned in the EIS was also carried out using available data and information from similar waste shipments. Lastly, a geochemical analysis of the various geological repository discussed in the EIS was also carried out using the Geochemists Workbench Release 14. Long-term disposal of the beryllium blocks at the Waste Isolation Pilot Plant (WIPP) proved to be the most cost-effective long-term disposition option of the ones considered in the EIS. A dose rate calculation at both contact and remote-handling distances indicate that the analyzed beryllium shipments should not exceed the exposure limits at any of the considered locations. Greater-than-class-C waste has been left in a regulatory state of limbo for years, resulting in backlogs of inventory across several research sites in the United States. Due to its high activity and presence of transuranic isotopes, it is imperative to ensure that it remains inaccessible and sequestered both in its short-term interim as well as a long-term geologic time scale. The information and assessments done in the paper could potentially serve as a reference for any future shipments of this waste form at the WIPP facility as well as other disposition paths that may be considered in the future.

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DOE OSTI2013

Performance Impact of Fast Flow Paths Through Grout Monoliths Used for Radioactive Waste Disposal - 13224

Empty HLW handling and storage tanks at SRS and INL contain residual radioactivity; these tanks are being stabilized with cementitious grout during closure operations. The US NRC directed the Center for Nuclear Waste Regulatory Analyses (CNWRA{sup R}) to develop physical analogs of cementitious grout monoliths to investigate their potential to form fast flow pathways such as macro-cracks, separations between grout lifts, and annuli around pipes, supports, and along tank walls. CNWRA developed and tested 15 55-gal-drum-scale specimens and 2 larger specimens of tank-filling cementitious grout, and 9 specimens of pipe-filling grout. These experiments demonstrated that the size of fast flow pathways that develop and the peak temperatures attained during hydration are proportional to the scale of the specimen, and that annular apertures and bulk grout permeability generally increased with time post-placement. Cracks developed overnight following placement of each grout lift in the largest specimen, but developed more slowly in smaller specimens, perhaps due to a ∼20 deg. C difference in peak temperatures, which influence the thermal gradients that can induce cracking. Plastic and drying shrinkage commonly led to poor grout-to-metal and grout-to-grout bonding. Cracks, annular gaps, and grout flow lobe seams transmitted fluids during injection testing. Macro-scale flow pathways such as these are not readily observed in bench-scale specimens of cementitious tank grout. (authors)

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DOE OSTI2020

Life cycle Management of 324 Project High-Activity, Mixed/Low-Level Radioactive Waste - 20506

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) manages the 324 Building and is preparing to remotely excavate and disposition the radioactive soil beneath the building to allow further deactivation of the building. The 324 Building is a non-reactor Category 2 Nuclear Facility located in the 300 Area of the Hanford Site. Records indicate that in October 1986, approximately 516 liters of a concentrated liquid waste stream containing cesium-137 ({sup 137}Cs) and strontium-90 ({sup 90}Sr) spilled onto the floor of a hot cell, B-Cell, in the 324 Building. The spill contained an estimated 1.3 million curies (Ci) of radioactivity. A breach in the sump of B-Cell was discovered in November 2009 during characterization of the soil under the building. Exposure rates in excess of 10,000 R/hr were detected through a system of access pipes installed under the hot cell, and an estimated 1.557 E+05 Ci of {sup 137}Cs and 6.842 E+04 Ci of {sup 90}Sr are in the soil beneath B-Cell. The magnitude of the soil contamination below B-Cell represents one of the most challenging remediation activities in the DOE complex. The objective of CHPRC's 324 Project is to remotely excavate and disposition the highly contaminated soil under the hot cell so building deactivation and stabilization activities can resume. Depending on the volume and radioactivity of contaminated soil under B-Cell, the material will be dispositioned either at the Hanford Site's Environmental Restoration Disposal Facility (ERDF) or in grout monoliths that will be created in the hot cells (A, C and D) next to B-Cell during future building demolition activities. The current project scope includes installing structural supports under the walls of B-Cell and using remotely operated equipment to remove debris from the hot cell, remove debris mixed with grout on the floor, saw through and remove the cell floor and liner, and excavate contaminated soil under B-Cell to a depth of up to 3.66 meters (12 feet). Addition or removal of debris in A-Cell may be required to support B-Cell cleanout and excavation activities. The base approach is to disposition debris and soil bins in the Radiochemical Engineering Cell (REC) monoliths, and subsequently disposition the monoliths during building demolition. (authors)

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DOE OSTI2015

Hanford Double-Shell Tank AY-102 Radioactive Waste Leak Investigation Update

The presentation outline is: Briefly review leak integrity status of tank AY-102 and current leak behavior; Summarize recent initiatives to understand leak mechanism and to verify integrity of remaining waste confinement structures; describe planned waste recovery activities; and, introduce other papers on tank AY-102 topics.

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DOE OSTI2013

Decommissioning of the TRIGA mark II and III and radioactive waste management

KAERI has carried out decommissioning projects for two research reactors (KRR-1 and 2). The decommissioning project of KRR-1 (TRIGA Mark II) and 2 (TRIGA Mark III) was launched in 1997 with a total budget of 23.25 million US dollars. KRR-2 and all auxiliary facilities were already decommissioned, and KRR-1 is being decommissioned now. Much more dismantled waste is generated than in any other operations of nuclear facilities. Thus, the waste needs to be reduced and stabilized through decontamination or treatment before disposal. This paper introduces the current status of the decommissioning projects and describes the volume reduction and conditioning of decommissioning waste for final disposal. (authors)

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DOE OSTI2019

Leachate Treatment for Radioactive Waste Containment Mound: Pilot Test and Design - 19098

Canadian Nuclear Laboratories plans to construct an engineered containment mound (ECM) at its Chalk River Laboratories site in Ontario for safe management of radioactive and mixed waste at the facility. The ECM is designed for containment of 1,000,000 cubic meters of waste, and will be operated over a period of 50 years. Contact storm water, leachate, and other wastewater will be generated during operation of the ECM, and must be treated prior to discharge to the environment. The wastewater is projected to contain low concentrations of radionuclides and metals, as well as other cations and anions, and low concentrations of organic chemicals. Through comparison with surface water discharge criteria, radionuclides cesium-137 and strontium-90, and metals including aluminum, barium, copper, iron, lead, and manganese were projected to be present in the wastewater at concentrations that may exceed surface water discharge criteria, and were therefore targeted for treatment. Several technologies, including chemical precipitation, membrane filtration, ion exchange, and reverse osmosis were reviewed and tested on a laboratory scale using simulated wastewater prepared to replicate the concentrations of both target and non-target constituents. Based upon the results of the laboratory-scale tests, a six-week pilot test was conducted to simulate the performance of the technologies under continuous-flow conditions and determine full-scale design criteria. This paper describes the methodology and results of the six-week pilot test, and how the test results were used to inform the full-scale wastewater treatment system design. (authors)

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DOE OSTI2019

Plutonium Migration during the Leaching of Cemented Radioactive Waste Sludges

One of the most challenging components of the UK nuclear legacy is Magnox sludge, arising from the corrosion of Mg alloy-clad irradiated metallic U fuel that has been stored in high pH ponds. The sludges mainly comprise Mg hydroxide and carbonate phases, contaminated with fission products and actinides, including Pu. Cementation and deep geological disposal is one option for the long-term management of this material, but there is a need to understand how Pu may be leached from the waste, if it is exposed to groundwater. Here, we show that cemented Mg(OH) 2 powder prepared with Pu(IV) aq is altered on contact with water to produce a visibly altered ‘leached zone’, which penetrates several hundred microns into the sample. In turn, this zone shows slow leaching of Pu, with long-term leaching rates between 1.8–4.4 × 10 -5 % of total Pu per day. Synchrotron micro-focus X-ray fluorescence mapping identified decreased Pu concentration within the ‘leached zone’. A comparison of micro-focus X-ray absorption spectroscopy (µ-XAS) spectra collected across both leached and unleached samples showed little variation, and indicated that Pu was present in a similar oxidation state and coordination environment. Fitting of the XANES spectra between single oxidation state standards and EXAFS modeling showed that Pu was present as a mixture of Pu(IV) and Pu(V). The change in Pu oxidation from the stock solution suggests that partial Pu oxidation occurred during sample ageing. Similarity in the XAS spectra from all samples, with different local chemistries, indicated that the Pu oxidation state was not perturbed by macro-scale variations in cement chemistry, surface oxidation, sample aging, or the leaching treatment. These experiments have demonstrated the potential for leaching of Pu from cementitious waste forms, and its underlying significance requires further investigation.

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