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

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At least 577 records · Page 32
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DOE OSTI2014

Technetium Inventory, Distribution, and Speciation in Hanford Tanks

The purpose of this report is three fold: 1) assemble the available information regarding technetium (Tc) inventory, distribution between phases, and speciation in Hanford’s 177 storage tanks into a single, detailed, comprehensive assessment; 2) discuss the fate (distribution/speciation) of Tc once retrieved from the storage tanks and processed into a final waste form; and 3) discuss/document in less detail the available data on the inventory of Tc in other "pools" such as the vadose zone below inactive cribs and trenches, below single-shell tanks (SSTs) that have leaked, and in the groundwater below the Hanford Site. A thorough understanding of the inventory for mobile contaminants is key to any performance or risk assessment for Hanford Site facilities because potential groundwater and river contamination levels are proportional to the amount of contaminants disposed at the Hanford Site. Because the majority of the total 99 Tc produced at Hanford (~32,600 Ci) is currently stored in Hanford’s 177 tanks (~26,500 Ci), there is a critical need for knowledge of the fate of this 99 Tc as it is removed from the tanks and processed into a final solid waste form. Current flow sheets for the Hanford Waste Treatment and Immobilization Plant process show most of the 99 Tc will be immobilized as low-activity waste glass that will remain on the Hanford Site and disposed at the Integrated Disposal Facility (IDF); only a small fraction will be shipped to a geologic repository with the immobilized high-level waste. Past performance assessment studies, which focused on groundwater protection, have shown that 99 Tc would be the primary dose contributor to the IDF performance.

storage tanks↗
DOE OSTI2013

The aqueous corrosion behavior of technetium - Alloy and composite materials

Metal waste forms are under study as possible disposal forms for technetium and other fission products. The alloying of Tc is desirable to reduce the melting point of the Tc-containing metal waste form and potentially improve its corrosion resistance. Technetium-nickel composites were made by mixing the two metal powders and pressing the mixture to make a pellet. The as-pressed composite materials were compared to sintered composites and alloys of identical composition in electrochemical corrosion tests. As-pressed samples were not robust enough for fine polishing and only a limited number of corrosion tests were performed. Alloys and composites with 10 wt% Tc appear to be more corrosion resistant at open circuit than the individual components based on linear polarization resistance and polarization data. The addition of 10 wt% Tc to Ni appears beneficial at open circuit, but detrimental upon anodic polarization. Qualitatively, the polarizations of 10 wt% Tc alloys and composites appear like crude addition of Tc plus Ni. The 1 wt% Tc alloys behave like pure Ni, but some effect of Tc is seen upon polarization. Cathodic polarization of Tc by Ni appears feasible based on open circuit potential measurements, however, zero resistance ammetry and solution measurements are necessary to confirm cathodic protection.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Joint EM-NE-International Study of Glass Behavior over Geologic Time Scales - 12303

Vitrification has been chosen as the best demonstrated available technology for waste immobilization worldwide. To date, the contributions of physical and chemical processes controlling the long-term glass dissolution rate in geologic disposal remain uncertain, leading to a lack of international consensus on a rate law for glass corrosion. Existing rate laws have overcome uncertainty through conservatism, but a thorough mechanistic understanding of waste form durability in geologic environments would improve public and regulator confidence. If it is possible to take credit for the true durability of the waste form in repository system evaluations, then it is possible to design the repository with less conservatism with concomitant cost savings. To gain a fundamental understanding of the dissolution rate law, six nations have joined together to formulate a joint plan for collaborative research into the mechanisms controlling the long-term corrosion of glass. This report highlights the technical program plan behind the US portion of this effort, with an emphasis on the current understanding (and limitations) of several mechanistic theories for glass corrosion. Some recent results are presented to provide an example of the ongoing research. Atom probe tomography has been used to provide a high-resolution analysis of elemental concentration gradients present at the hydrated glass / pristine glass interface in SON68 after 25.75 years of corrosion in a simulated granitic groundwater at 90 deg. C. The most valuable result of these initial studies is the success of the technique. Characterization by APT had never been previously demonstrated for glass corrosion layers. The resolution of APT is a powerful addition to the tools with which we can investigate the mechanisms dominating glass corrosion. Some other key results of this study include the observation that the elemental interfacial width between the hydrated glass and pristine glass appears to be much sharper (∼2 nm for B, Na and Al) than had been previously measured using nanoSIMS (∼240 nm). It is not clear whether the APT analysis and nanoSIMS characterizations were possibly performed on topographically unique regions, or whether nanoSIMS overestimated the elemental width. However, the APT data seems very convincing that the elemental width can be much sharper than was previously thought. This result calls into question some of the assumptions made for the diffusion-control models of glass dissolution, since such a sharp profile would not match the diffusion coefficients used to date. Other results, such as the observation of apparently layered concentration profiles, show that gel evolution is not as simple as is currently assumed in nearly every model. This task is a good example of the collaborative nature of the I-TEAM effort. Based on experimental needs and differences in expertise, scientists from DOE and CEA worked together to change the level of understanding in the field. These types of interactions are nearly ubiquitous among the tasks in the technical program plan. With the excellence of the team in place and the willingness of the participants to work together for a common understanding, the stated goal of consensus on the mechanistic basis for radionuclide release from glass is well within reach. (authors)

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DOE OSTI2016

On the Robustness of a Cement-Based Matrix for the Conditioning of Evaporator Concentrates - 16417

The presented work is research in progress on the robustness of a cement-based matrix designed for the conditioning of evaporator concentrates with highly variable chemical composition. A change in the waste composition can induce significant alteration of the cement-waste form. However, all criteria imposed by the authorities on the immobilized waste need to be met and the processability of all these possible waste compositions has to be guaranteed. Defining a domain of acceptable waste compositions usually requires to investigating many parameters, which results, with classical approaches, in a high number of trials. Design of experiments (DOE) is shown to be a fruitful alternative which enables to keep the total number of runs within reason without compromising the quality of the achieved results. An example is given, showing how a Plackett-Burman design can be used to determine, among all the waste components, those which significantly influence the properties of the cement-waste forms when they vary within the experimental domain of interest. (authors)

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DOE OSTI2024

Evaluating Iodine Immobilization Technologies: Cermets, Polycermets, and Polyhalmets

The work in this report documents the efforts conducted to assess the feasibility of some of the ideas documented in Pacific Northwest National Laboratory invention disclosure reports (IDRs) including: 1) Iodine capture in polyacrylonitrile (PAN)-containing composite sorbents (32451-E). In this work, the composites evaluated included Ag0, Bi0, Cu0, Bi2S3, and Cu2S embedded in PAN. 2) Metal iodide removal from these sorbents through dissolution in dimethyl sulfoxide (DMSO) (32729-E). In this work, PAN dissolution was evaluated for multiple types of sorbents including Ag-Pan, Bi-PAN, Cu-PAN, Bi2S3-PAN, and Cu2S-PAN. 3) Using metal-sulfide sorbents for iodine capture (32647-E). In this work, the composites evaluated under this IDR included Ag2S, Bi2S3, and Cu2S embedded in PAN. 4) Using low-melting metals to immobilize (encapsulate) iodine-loaded and polymer-containing sorbents into polymer-ceramic-metal (called polycermet) or polymer-halide-metal (called polyhalmet) composite waste forms (32625-E). In this work, the iodine-loaded PAN composites included AgI-PAN, BiI-PAN, and CuI-PAN. 5) Ceramic-metal composite waste form synthesis of polymer-containing materials using low-melting metals like bismuth, tin, or bismuth-tin alloys (32537-E). In this work, the metals evaluated included Bi, 58Bi-42Sn eutectic. 6) Cermets for immobilizing commercial sorbents loaded with radioiodine (32806-E). In this work, AgIX (iodine-loaded silver faujasite zeolite) was evaluated in cermet form.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

New Formulation of Cast Stone for {sup 99}Tc Immobilization in Sulfate Rich Secondary Waste Stream - 16098

A low-temperature, cementitious waste form called Cast Stone (or grout) has been selected for solidification of the Effluent Treatment Facility (ETF)-treated wastes at the Hanford Site. Because the waste stream to be solidified may be relatively high in sulfate, the current Cast Stone dry blend formulation with ordinary Portland cement (OPC), fly ash (FA), and blast furnace slag (BFS) (47% BFS, 45% FA, and 8% OPC) may not be adequate. Thus a new formulation, called secondary waste grout, was proposed and tested. The secondary waste grout uses hydrated lime [Ca(OH){sub 2}] to tie up the sulfate initially such that ettringite [Ca{sub 6}Al{sub 2}(SO{sub 4}){sub 3}(OH){sub 12}.26(H{sub 2}O)] forms rapidly instead of forming as the cementitious waste form cures. Three different liquid waste simulants (242-A evaporator, Environmental Restoration Disposal Facility [ERDF] leachates, and Hanford Tank Waste Treatment and Immobilization Plant [WTP] off-gas condensates) were prepared and solidified using this new dry blend formulation. Baseline grout dry mixes were 20 wt% hydrated lime (HL), 35 wt% OPC, and 45 wt% BFS, and two different water-to-dry mix (w/dm) ratios (0.5 and 0.6) of free water mass to mass of dry blend were tested. After curing for a minimum of 28 days, the grout solidified waste forms were characterized and used for Tc leachability testing using the EPA 1315 Method. X-ray diffraction analysis confirmed that ∼12-15% of ettringite was formed in early stage of 28-day curing for the grouts prepared with the high sulfate WTP simulant and HL addition. In addition, the new formulation of secondary waste grout prepared with HL (20%), OPC (35%), and BFS (45%) at either 0.5 or 0.6 water to dry mix ratio yields higher average Tc interval leachability index (LI) values (12.5 to 13.9) after 28 to 90 days of leaching using deionized water, which is higher than the desired {sup 99}Tc LI value of 12 for grout destined for burial in the Integrated Disposal Facility at the Hanford Site. (authors)

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DOE OSTI2017

TN{sup R} MW- Multi Waste Package 'All In One Solution' For Conditioning - Transport - Storage - Up To Final Disposal - 17196

For the management of radioactive waste, either during nuclear facility operation or decommissioning, one must consider a diverse array of nuclear waste forms of various types, volumes and activities. Examples are: activated metallic waste; legacy waste; sludge; resins; and orphan waste. Customers are frequently faced with the obligation of undertaking multiple, costly waste management operations. These operations activities include handling, reconditioning or material transfers between packages, as in the case when moving material from on-site storage to transportation, and then from transportation to final disposal. To address this issue, a new - highly flexible up to reversible - cask system, the TN{sup R} MW, is being developed as an 'All in One Solution'. TN{sup R} MW has a total weight of 10 metric tons and is compliant with the 2012 International Atomic Energy Agency (IAEA) regulations. It was developed to provide flexibility and designed with the capability to be adaptable to the various needs of the nuclear industry. From IP-2 and Type A containers to Type B(U) and B(U)F casks, this cask offers different options to sufficiently address the requirements of each step of the waste stream from on-site and international transport to long-term interim storage and final disposal. Licensing and manufacturing of a number of packagings in this TN{sup R} MW family are underway. (authors)

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DOE OSTI2015

Deep Borehole Field Test: Characterization Borehole Science Objectives.

Deep Borehole Disposal (DBD) of high-level radioactive wastes has been considered an option for geological isolation for many years (Hess et al. 1957). Recent advances in drilling technology have decreased costs and increased reliability for large-diameter (i.e., ≥50 cm [19.7”]) boreholes to depths of several kilometers (Beswick 2008). These advances have therefore also increased the feasibility of the DBD concept (Brady et al. 2009), and the current field test, introduced herein, is a demonstration of the DBD concept and these advances. The US Department of Energy (DOE) Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013) specifically recommended developing a research and development plan for DBD as a key strategy objective. DOE’s Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel (DOE 2014a) concludes “effective implementation of a strategy for management and disposal of all High-Level Waste and Spent Nuclear Fuel” would include focused research on deep boreholes, especially to retain flexible options for disposal of physically smaller DOEmanaged solid radioactive waste forms. More information regarding the characteristics, quantities, and sizes of these physically smaller waste forms is in the Evaluation of Options for Permanent Geologic Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste (SNL 2014).

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DOE OSTI2016

CRADA FINAL REPORT FOR CRADA NO. CR-08-001 CEMENTITIOUS BARRIERS PARTNERSHIPS

The Cementitious Barriers Partnership (CBP), led by the Savannah River National Laboratory, has achieved its ultimate purpose to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex. The CBP achieved its goal by developing experimental programs with results that supported the development of a reasonable and credible set of software tools to improve understanding and prediction of the structural, hydraulic and chemical performance of cementitious barriers and waste forms used in nuclear applications. These experimental data and software tools are and will continue to reduce the uncertainties in the current methodologies for assessing cementitious barrier performance and increase the consistency and transparency of the assessment process. The CBP has made a significant impact in reducing risk and uncertainties in the performance assessments that directly impact U.S. Department of Energy (DOE) environmental cleanup and closure programs. The CBP has developed an experimentally-based, integrated modeling tool that serves as a screening tool known as the CBP Software Toolbox, Version 3.0. The CBP products reduce the uncertainties of current methodologies for assessing cementitious barrier performance and increasing the consistency and transparency of the DOE assessment process. This CBP Software Toolbox predicts the hydraulic properties and chemical stability of cement matrix phases and release fluxes of constituents, in response to variable boundary conditions. Therefore, the CBP tools are reducing the uncertainty by coupling multiscale and multi-physics processes, including physical-chemical evolution and transport phenomena applied to heterogeneous materials with changing boundary conditions. In addition, the CBP Experimental and Characterization Program has provided test data to support chemical and physical degradation mechanisms for cementitious barriers. These tools have application to low activity waste forms, high level waste tank closure, D&D and entombment of major nuclear facilities, landfill waste acceptance criteria, and in-situ grouting and immobilization of vadose zone contamination. The CBP also provided: 1) a forum for DOE-NRC technical exchange, 2) material characterization to support performance assessment (PA) predictions, 3) uncertainty reduction in PA predictions, 4) base case performance to improve PA predictions, and 5) improvement in understanding and quantification of moisture and contaminant transport used in PAs. Additional CBP accomplishments include: sponsorship of a national test bed workshop to obtain collaboration in establishing the path forward to obtain actual data supporting future predictions of cementitious barrier performance, and participation in an International Atomic Energy Agency (IAEA) Cooperative Research Project on the use of cementitious barriers for low-level radioactive waste treatment and disposal.

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DOE OSTI2023

Technology Development Roadmap for Volatile Radionuclide Capture and Immobilization

An “Off-Gas and Waste Forms Strategy Workshop” was held in Washington, D.C. on January 31, and February 1, 2023, to review the current state of the art (baseline) technologies for capturing and immobilizing these volatile radionuclides in waste forms (WFs). The discussions in that workshop were used to develop a roadmap for future research and development to mature technologies that are presently not ready for use in a commercial reprocessing facility and develop and demonstrate technologies that provide more safety, more simplicity, or lower costs compared to the current baseline.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Peridynamic modeling of cementitious materials for nuclear waste management

Radioactive and hazardous waste generated from fuel processing plants, nuclear reactors, and hospitals, requires effective management strategies. Cementitious materials are widely applied for these needs, serving as structural materials, reactive barriers, or waste forms. In these applications cracking poses a significant risk to performance, driven by inconsistent shrinkage of the components and the varying strength and permeability of their interfaces. Traditional modeling approaches face challenges in representing the complex fracture behavior of cementitious materials due to the reliance on spatial derivatives and difficulties with mesh generation. Here, to overcome these limitations, we employ peridynamics, a novel continuum mechanics formulation that uses integrals to describe mechanical equilibrium, avoiding discontinuities associated with traditional methods. Through incorporation of a bi-linear softening model and quasistatics, an experimentally validated model for Portland cement concrete samples was created. Mechanical parameters, including compressive strength and elastic modulus were validated and variation due to aggregate packing was evaluated. Additionally, sensitivity analysis of the peridynamic parameters to the bulk material properties was established. The results lay the groundwork for evaluating the impact of unique conditions of cementitious waste forms that can be assessed to improve the reliability of waste management strategies.

Aggregates↗
DOE OSTI2013

Building flexibility into the design of a pilot plant for the immobilisation of Pu containing residues and wastes

NNL and ANSTO on behalf of Sellafield Ltd have developed a process for the immobilisation of a range of Pu containing wastes and residues. Following the inactive demonstration of the technology the project is now focusing on the design of an active pilot plant capable of validating the technology and ultimately immobilising a waste inventory containing around 100 kg plutonium. The diverse wastes from which it is uneconomic to recover Pu, require a flexible process with a wide product envelope capable of producing a wasteform suitable for disposal in a UK repository. Ceramics, glass ceramics and metal encapsulated waste-forms can be delivered by the process line which incorporates size reduction and heat treatment techniques with the aim of feeding a hot isostatic pressing process designed to deliver the highly durable waste-forms. Following a demonstration of feasibility, flowsheet development is progressing to support the design which has the aim of a fully flexible facility based in NNL's Central Laboratory on the Sellafield site. Optimisation of the size reduction, mixing and blending operations is being carried out using UO{sub 2} as a surrogate for PuO{sub 2}. This work is supporting the potential of using an enhanced glass ceramic formulation in place of the full ceramic with the aim of simplifying glove box operations. Heat treatment and subsequent HIPing strategies are being explored in order to eliminate any carbon from the feeds without increasing the valence state of the uranium present in some of the inventory which can result in an unwanted increase in wasteform volumes. The HIP and ancillary systems are being specifically designed to meet the requirements of the Sellafield site and within the constraints of the NNL Central Laboratory. The HIP is being configured to produce consolidated product cans consistent with the requirements of ongoing storage and disposal. With the aim of one cycle per day, the facility will deliver its mission of immobilising the identified waste and residues inventory within 3 years. During that period it will also be used to demonstrate the potential of this technology to deliver the immobilisation of a proportion of the UK plutonium stockpile that may not be suitable for use as MOx fuel should that decision be taken. (authors)

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DOE OSTI2015

ITEAM - An international mechanistic approach to modeling glass behavior over geologic time scales - 15527

Vitrification is a common worldwide solution for radioactive waste immobilization. Although the excellent chemical durability of glass has been demonstrated in ancient man-made and natural analogues, the lack of an international consensus on a glass corrosion rate law limits the application of glass in high-level waste repository performance models. The contributions of physical and chemical processes controlling the long-term glass dissolution rate in geologic disposal remain uncertain resulting in this non-consensus. A thorough mechanistic understanding of waste form durability in geologic environments would improve public and regulator confidence. In addition, this understanding could lead to cost savings if it becomes possible to take credit for the true durability of the waste form itself in system evaluations rather than requiring expensive engineered barrier systems. To this end, five nations (US, UK, France, Japan, and Belgium) have joined together to formulate a joint plan for collaborative research into the mechanisms controlling the long-term corrosion of glass. Over the last five years, the coordinated research performed by these countries has produced a new level of understanding about the glass corrosion mechanisms, specifically as they apply to glasses similar in composition to the French R7T7 glass. Some recent results are presented as they relate to several mechanisms that are thought to control glass alteration at long time periods. It is important to understand the effects of each particular mechanism on the overall radionuclide release and the relative importance of each process model to the performance assessment calculation. In addition, a new series of experiments calls out potential sources of error due to the use of irregular shapes (such as crushed glass powder) in dissolution experiments. (authors)

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DOE OSTI2019

Multi-glass investigation of Stage III glass dissolution behavior from 22 to 90 °C triggered by the addition of zeolite phases

Evaluating the corrosion of glass waste forms for nuclear waste immobilization is a prime metric of their performance. Stage III behavior, the delayed acceleration of glass corrosion, remains the most potentially impactful aspect of glass corrosion behavior. Using the addition of various zeolite phases to trigger this effect, the properties of Stage III behavior were evaluated for three high-level waste glass compositions: SRL-202A, SON68, and AFCI. We demonstrate that Stage III behavior can occur at temperatures as low as 22°C, with rate acceleration observed at 90 and 70 °C for all studied glasses, and at 40 °C and 22 °C for two of the compositions (SRL-202A and SON68). Using an Arrhenius fit, the activation energy of the process was found to be highly uncertain, with values between 39 and 68 kJ·mol -1 . Stage III behavior was triggered for each of the glasses studied by the addition of each of the four zeolites studied (Na-P1, Na-P2, Analcime, and Chabazite). The rates of alteration during Stage III varied very little, from 0.0058 to 0.023 g·m -2 ·d -1 for the accelerated region immediately following zeolite addition. The acceleration was discovered to be transient, with the alteration rate decreasing and then accelerating again as the experiments proceeded. The solution composition was compared to the strong base/weak acid discriminator developed by Jantzen et al. and it was found that the addition of zeolites can overcome slightly non-optimal solution conditions for zeolite formation to trigger Stage III behavior.

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DOE OSTI2015

Technetium Getters to Improve Cast Stone Performance

The cementitious material known as Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. Two radionuclides of particular concern in these waste streams are technetium-99 (99Tc) and iodine-129 (129I). These radioactive tank waste components contribute the most to the environmental impacts associated with the cleanup of the Hanford site. A recent environmental assessment of Cast Stone performance, which assumes a diffusion controlled release of contaminants from the waste form, calculates groundwater in excess of the allowable maximum permissible concentrations for both contaminants. There is, therefore, a need and an opportunity to improve the retention of both 99Tc and 129I in Cast Stone. One method to improve the performance of Cast Stone is through the addition of “getters” that selectively sequester Tc and I, therefore reducing their diffusion out of Cast Stone. In this paper, we present results of Tc and I removal from solution with various getters with batch sorption experiments conducted in deionized water (DIW) and a highly caustic 7.8 M Na Ave LAW simulant. In general, the data show that the selected getters are effective in DIW but their performance is comprised when experiments are performed with the 7.8 M Na Ave LAW simulant. Reasons for the mitigated performance in the LAW simulant may be due to competition with Cr present in the 7.8 M Na Ave LAW simulant and to a pH effect.

Cast Stone↗
DOE OSTI2015

Cementitious Barriers Partnerships (FY2015 Mid-Year Status Report)

The DOE-EM Office of Tank Waste Management Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. Therefore, the CBP ultimate purpose is to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex. This status report highlights the CBP 2015 Software and Experimental Program efforts and accomplishments that support DOE needs in environmental cleanup and waste disposal. DOE needs in this area include: Long-term performance predictions to provide credibility (i.e., a defensible technical basis) for regulator and DOE review and approvals, Facility flow sheet development/enhancements, and Conceptual designs for new disposal facilities. In 2015, the CBP plans to release its newest version, the CBP Software Toolbox – “Version 3.0”, which will include new STADIUM carbonation and damage models, a new SRNL module for transport properties and flow in fractured and in-tact cementitious materials, and a new LeachXS/ORCHESTRA (LXO) oxidation module. In addition, improved STADIUM sulfate attack and chloride models will be included. This is in addition to the LXO modules for sulfate attack, carbonation, constituent leaching, and percolation with radial diffusion (for leaching and transport in cracked cementitious materials), of which will also be improved for Version 3.0. These STADIUM and LXO models are applicable and can be used by both DOE and the Nuclear Regulatory Commission (NRC) for service life and long-term performance evaluations and predictions of nuclear and radioactive waste containment structures across the DOE complex. In 2015, the Cementitious Barriers Partnership is providing tangible progress toward fulfilling the objective of developing a set of software tools and experimental programs to improve understanding and prediction of the long-term structural, hydraulic and chemical performance of cementitious barriers used in nuclear applications. To reflect this progress, CBP partners authored and are currently drafting many reports including six papers that were presented at WM2015.

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DOE OSTI2015

In-service Inspection of Radioactive Waste Tanks at the Savannah River Site – 15410

Liquid radioactive wastes from the Savannah River Site (SRS) separation process are stored in large underground carbon steel tanks. The high level wastes are processed in several of the tanks and then transferred by piping to other site facilities for further processing before they are stabilized in a vitrified or grout waste form. Based on waste removal and processing schedules, many of the tanks will be required to be in service for times exceeding the initial intended life. Until the waste is removed from storage, transferred, and processed, the materials and structures of the tanks must maintain a confinement function by providing a barrier to the environment and by maintaining acceptable structural stability during design basis events, which include loadings from both normal service and abnormal (e.g., earthquake) conditions. A structural integrity program is in place to maintain the structural and leak integrity functions of these waste tanks throughout their intended service life. In-service inspection (ISI) is an essential element of a comprehensive structural integrity program for the waste tanks at the Savannah River Site (SRS). The ISI program was developed to determine the degree of degradation the waste tanks have experienced due to service conditions. As a result of the inspections, an assessment can be made of the effectiveness of corrosion controls for the waste chemistry, which precludes accelerated localized and general corrosion of the waste tanks. Ultrasonic inspections (UT) are performed to detect and quantify the degree of general wall thinning, pitting and cracking as a measure of tank degradation. The results from these inspections through 2013, for the 27 Type III/IIIA tanks, indicate no reportable in-service corrosion degradation in the primary tank (i.e., general, pitting, or cracking). The average wall thickness for all tanks remains above the manufactured nominal thickness minus 0.25 millimeter and the largest pit identified is approximately 1.70 millimeter deep (i.e., less than 10% through-wall). Improvements to the inspection program were recently instituted to provide additional confidence in the degradation rates. Thickness measurements from a single vertical strip along the accessible height of the primary tank have been used as a baseline to compare historical measurements. Changes in wall thickness and pit depths along this vertical strip are utilized to estimate the rate of corrosion degradation. An independent review of the ISI program methodology, results, and path forward was held in August 2009. The review recommended statistical sampling of the tanks to improve the confidence of the single strip inspection program. The statistical sampling plan required that SRS increase the amount of area scanned per tank. Therefore, in addition to the baseline vertical strip that is obtained for historical comparisons, four additional randomly selected vertical strips are inspected. To date, a total of 104 independent vertical strips along the height of the primary tank have been completed. A statistical analysis of the data indicates that at this coverage level there is a 99.5% confidence level that one of the worst 5% of all the vertical strips was inspected. That is, there is a relatively high likelihood that the SRS inspection program has covered one of the most corroded areas of any of the Type III/IIIA waste tanks. These data further support the conclusion that there are no significant indications of wall thinning or pitting. Random sampling will continue to increase the confidence that one of the worst 5% has been inspected. In order to obtain the additional vertical strips, and minimize budget and schedule impacts, data collection speed for the UT system was optimized. Prior to 2009, the system collected data at a rate of 32 square centimeters per minute. The scan rate was increased to 129 - 160 square centimeters per minute by increasing the scanner step and pixel sizes in the data acquisition set-up. Laboratory testing was utilized to optimize the scan index/pixel size such that the requirements for wall thinning and pit detection were still maintained. SRS continues to evaluate improvements to ultrasonic equipment.

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