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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Effect of decades of corrosion on the microstructure of altered glasses and their radiation stability

Understanding the microstructural evolution of glasses during their interaction with water and radiation is of fundamental importance in addressing the corrosion of nuclear waste forms under geological disposal conditions. Here we report the results of more than 21 years of corrosion of two borosilicate glasses showing the formation of mesoporous C–S–H gels in Ca-bearing glasses and a mainly microporous microstructure in Al-bearing glasses. These porous corroded glasses were then irradiated with heavy ions to simulate the effects of recoil nucleus damage and monitored in real time using transmission electron microscopy with in situ ion irradiation. The ballistic collisions remarkably healed the porous corroded glasses to a pore-free homogeneous microstructure. Besides providing new insights and predictions about how doped glasses and actual waste forms may evolve under corrosion and irradiation, the results highlight the non-universal nature of the existing corrosion models and the important role that the glass composition and radiation damage play in the evolution of the microstructure during corrosion.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Development of pyro-processing technology at CRIEPI for carving out the future of nuclear fuel cycle

Pyro-processing has been attracting increasing attention as a promising candidate as an advanced nuclear fuel cycle technology. It provides economic advantage as well as reduction in proliferation risk and burden of long live radioactive waste, especially when it is combined with advanced fuels such as metallic or nitride fuel which gives excellent burning efficiency of minor actinides (MA). CRIEPI has been developing pyro-processing technology since late eighties with both domestic and international collaborations. In the early stage, electrochemical and thermodynamic properties in LiCl-KCl eutectic melt, and fundamental feasibility of core technology like electrorefining were chiefly investigated. Currently, stress in the process chemistry development is also placed on supporting technologies, such as treatment of anode residue and high temperature distillation for cathode product from electrorefining, and so on. Waste treatment process development, such as studies on adsorption behavior of various FP elements into zeolite and conditions for the fabrication of glass-bonded sodalite waste form, are steadily improved as well. In parallel, dedicated pyro-processing equipment such as zeolite column for treatment of spent electro-refiner salt is currently in progress. Recently, an integrated engineering-scale fuel cycle tests were performed funded by Japanese government (MEXT) as an important step before proceeding to large scale hot demonstration of pyro-processing. Oxide fuels can be readily introduced into the pyro-processing by reducing them to metals by adoption of electrochemical reduction technique. Making use of this advantage, the pyro-processing is currently under preliminary evaluation for its applicability to the treatment of the corium, mainly consisting of (U,Zr)O{sub 2}, formed in different composition during the accident of the Fukushima Daiichi nuclear power plant. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Bubbling Behavior in a Waste Glass Melter

Legacy tank waste from five decades of plutonium production at the Hanford Site during the Cold War will be vitrified into a stable waste form for disposal. A Waste Treatment Plant, currently under construction, houses Joule heated ceramic-lined waste glass melters wherein radioactive waste will be combined with glass formers to produce borosilicate glass. Through testing conducted in pilot-scale melters with surrogate waste, as well as operational experience at other sites, it has been demonstrated that the use of bubblers can dramatically improve the glass production rate. However, knowledge of the bubble dynamics in the molten glass is necessary for prediction and design of improved melting processes. A computational fluid dynamics model of the melter has been developed to investigate the processes affecting feed melt rate. The model shows that the convection currents within the melt pool due to Joule heating supplied by the electrodes are enhanced by forced air bubbling. In this paper, parameters influencing the formation and rise of a bubble stream in a viscous liquid are investigated. Experimental data and observations from laboratory and pilot-scale testing are integrated with first principles to develop a set of equations that form a mechanistic bubbling model to be incorporated into the melter model.

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

Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed.

Kruger, Albert A.↗
DOE OSTI2012

Fukushima Nuclear Crisis Recovery: A Modular Water Treatment System Deployed in Seven Weeks - 12489

On March 11, 2011, the magnitude 9.0 Great East Japan earthquake, Tohoku, hit off the Fukushima coast of Japan. This was one of the most powerful earthquakes in recorded history and the most powerful one known to have hit Japan. The ensuing tsunami devastated a huge area resulting in some 25,000 persons confirmed dead or missing. The perfect storm was complete when the tsunami then found the four reactor, Fukushima-Daiichi Nuclear Station directly in its destructive path. While recovery systems admirably survived the powerful earthquake, the seawater from the tsunami knocked the emergency cooling systems out and did extensive damage to the plant and site. Subsequent hydrogen generation caused explosions which extended this damage to a new level and further flooded the buildings with highly contaminated water. Some 2 million people were evacuated from a fifty mile radius of the area and evaluation and cleanup began. Teams were assembled in Tokyo the first week of April to lay out potential plans for the immediate treatment of some 63 million gallons (a number which later exceeded 110 million gallons) of highly contaminated water to avoid overflow from the buildings as well as supply the desperately needed clean cooling water for the reactors. A system had to be deployed with a very brief cold shake down and hot startup before the rainy season started in early June. Joined by team members Toshiba (oil removal system), AREVA (chemical precipitation system) and Hitachi-GE (RO system), Kurion (cesium removal system following the oil separator) proposed, designed, fabricated, delivered and started up a one of a kind treatment skid and over 100 metric tons of specially engineered and modified Ion Specific Media (ISM) customized for this very challenging seawater/oil application, all in seven weeks. After a very short cold shake down, the system went into operation on June 17, 2011 on actual waste waters far exceeding 1 million Bq/mL in cesium and many other isotopes. One must remember that, in addition to attempting to do isotope removal in the competition of seawater (as high as 18,000 ppm sodium due to concentration), some 350,000 gallons of turbine oil was dispersed into the flooded buildings as well. The proposed system consisted of a 4 guard vessel skid for the oil and debris, 4 skids containing 16 cesium towers in a lead-lag layout with removable vessels (sent to an interim storage facility), and a 4 polishing vessel skid for iodine removal and trace cesium levels. At a flow rate of at least 220 gallons per minute, the system has routinely removed over 99% of the cesium, the main component of the activity, since going on line. To date, some 50% of the original activity has been removed and stabilized and cold shutdown of the plant was announced on December 10, 2011. In March and April alone, 10 cubic feet of Engineered Herschelite was shipped to Seabrook Nuclear Power Plant, NPP, to support the April 1, 2011 outage cleanup; 400 cubic feet was shipped to Oak Ridge National Laboratory (ORNL) for strontium (Sr-90) ground water remediation; and 6000 cubic feet (100 metric tons, MT, or 220,400 pounds) was readied for the Fukushima Nuclear Power Station with an additional 100 MT on standby for replacement vessels. This experience and accelerated media production in the U.S. bore direct application to what was to soon be used in Fukushima. How such a sophisticated and totally unique system and huge amount of media could be deployable in such a challenging and changing matrix, and in only seven weeks, is outlined in this paper as well as the system and operation itself. As demonstrated herein, all ten major steps leading up to the readiness and acceptance of a modular emergency technology recovery system were met and in a very short period of time, thus utilizing three decades of experience to produce and deliver such a system literally in seven weeks: - EPRI - U.S. Testing and Experience Leading to Introduction to EPRI - Japan and Subsequently TEPCO Emergency Meetings - Three Mile Island (TMI) Media and Vitrification Experience by PNNL - Commercial Nuclear Power Plant Media Experience (including long term Cs removal) - DOE Low Active Waste (LAW) and High Level Waste (HLW) in High Salt and pH Conditions Media and Vitrification Experience - National Laboratory (e.g. Oak Ridge National Laboratory, ORNL) Ground Water Media Experience - Gulf Oil Spill Media Experience in Seawater - All Media Had to be Fully Tested at High Rad Levels in Seawater and Oil Before Arriving in Japan - Final Waste Form and Disposal Experience (e.g., vitrification) - 100 Metric Tons (6000 cubic feet or 220,400 pounds) of Media had to be Immediately Available with the same amount in production as replacement media. [To date, for 2011, 400 MT of media have been prepared for Japan alone.] - Remote Operation, Modular Water Treatment Equipment Design and Fabrication in both Commercial NPP and DOE Canyon Operations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Disposal Systems Evaluation Framework DSEF User Manual (V.3.0)

The Disposal Systems Evaluation Framework (DSEF) has been developed at Lawrence Livermore National Laboratory (LLNL) to formalize and facilitate the development and documentation of repository conceptual design options for a range of waste forms, geologic environments, repository design concepts, and repository operating modes. DSEF is one component of the Engineered Barrier System (EBS) family of work packages at several national laboratories led by Sandia National Laboratories (SNL), in support of the DOE Office of Nuclear Energy, specifically of the Used Fuel Disposition (UFD) Campaign within the Office of Used Nuclear Fuel Disposition Research and Development (NE-53). DSEF is a knowledge management system that allows the user to intelligently access and draw data from a case library of hundreds of completed thermal analyses (currently 721 cases in the library), and draw input from databases of material properties and repository development cost data (currently the data in these databases are drawn from more than 100 references). The core functionality of DSEF is provided by a Microsoft Office Excel 2010 workbook with macros and form controls that create a structured environment and walk the user through the steps of creating the input data required to perform disposal system evaluations internally and by interfacing with external programs. The user can choose to work with built-in data from the case library of previous analysis cases, from the material property databases, or can define their own input descriptions and data. DSEF uses the Excel worksheet protection (implemented on a sheet by sheet basis) to prevent inadvertent over-writing of formulas with data. When worksheet protection is enabled, the user can see all underlying formulas, and enter data in unprotected cells (the unprotected cells generally correspond to the yellow highlighted cells). Worksheet protection also prevents restructuring the files (adding or deleting rows or columns) and changing the formatting, as well as preventing access to the Visual Basic Editor to prevent modification of the macros. If the user needs to modify the worksheet structure, formulas, or macros, they can contact LLNL to get the password to unprotect the worksheets. However, care should be taken in making any modifications, since the macros and formulas are dependent in some cases on the current structure, range names, and macro subroutines DSEF is set up to retrieve the results of external thermal analysis program runs, summarize them graphically, and provide a template for adding the new results to the Case Library. The Case Library worksheet in DSEF is set up to create a Case Library data record for each new analysis case, at the top of the list of prior cases. Therefore, a separate DSEF file (with unique name) should be saved for each run that is to be archived. Users should submit a copy of their DSEF files to LLNL so that their results can be added to the overall Case Library and made available to other laboratories and stakeholders. LLNL will maintain the DSEF Case Library and make the most current files available on the SNL SharePoint site. The Revision History sheet documents the file evolution. One of the key objectives of DSEF is to allow the user to compare the results of thermal analyses to the thermal constraints of the design components. These comparisons can then be used to consider design, layout, and operating mode modifications to find combinations that can meet the thermal constraints for a given combination of waste form, geologic medium, surface storage time, and system concept of operation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Incorporation of Technetium into Spinel Ferrites

Technetium ( 99 Tc) is a problematic fission product for the long-term disposal of nuclear waste due to its long half-life, high fission yield, and to the environmental mobility of pertechnetate, the stable species in aerobic environments. One approach to preventing 99 Tc contamination is using sufficiently durable waste forms. We report the incorporation of technetium into a family of synthetic spinel ferrites that have environmentally durable natural analogs. A combination of X-ray diffraction, X-ray absorption fine structure spectroscopy, and chemical analysis reveals that Tc(IV) replaces Fe(III) in octahedral sites and illustrates how the resulting charge mismatch is balanced. When a large excess of divalent metal ions is present, the charge is predominantly balanced by substitution of Fe(III) by M(II). When a large excess of divalent metal ions is absent, the charge is largely balanced by creation of vacancies among the Fe(III) sites (maghemitization). In most samples, Tc is present in Tc-rich regions rather than being homogeneously distributed.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Temperature-dependent void formation and growth at ion-irradiated nanocrystalline CeO 2 –Si interfaces

Ceria is a thermally stable ceramic that has numerous applications in the nuclear industry, including use in nuclear fuels and waste forms. Recently, interest has surged in nanostructured ceria due to its increased mechanical properties and electronic conductivity in comparison with bulk ceria and its ability to self-heal in response to energetic ion bombardment. Here, nanocrystalline ceria thin films grown over a silicon substrate are irradiated to fluences of up to ~4 × 10 16 ions/cm 2 under different irradiation conditions: with differing ion species (Si + and Ni + ), different ion energies (1.0–1.5 MeV), and at varying temperatures (160–600 K). Finally, while the nanocrystalline ceria is found to exhibit exceptional radiation resistance under all tested conditions, severe ion irradiation-induced mixing, void formation, and void growth are observed at the ceria/silicon interface, with the degree of damage proving to be temperature dependent.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2020

Iodosodalite synthesis with hot isostatic pressing of precursors produced from aqueous and hydrothermal processes

Iodosodalite powders were synthesized using aqueous and hydrothermal methods, with or without sodium borosilicate glass binders, and then hot isostatically pressed (HIP) at 900 ? and 175 MPa for 3 h to convert into waste forms. After HIPing, the structures, compositions, morphologies, porosities, and leach rates of the aqueous and hydrothermally produced iodosodalite samples were compared. X-ray diffraction patterns of HIPed samples showed that iodosodalite remained as the dominant phase. However, the fraction of the iodosodalite phase decreased after HIPing, indicating decomposition of the iodosodalite structure during the HIP process. Scanning electron microscopy and chemical imaging on the cross sections of HIPed samples showed homogeneous elemental distribution for aqueous-grown iodosodalite, whereas hydrothermally grown iodosodalite samples had more heterogeneity and porosity. The densities of HIPed samples with more glass binder were generally higher than sampled HIPed without binder. The leach tests on samples containing 20 mass% glass binder showed that the iodine dissolution rate of HIPed hydrothermally grown iodosodalite was about 4 times higher than iodosodalite by aqueous method. This study provides alternative techniques for immobilizing iodine-streams in iodosodalite waste forms.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

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

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

42 ENGINEERING↗
DOE OSTI2018

Innovative thermal distillation method using solar heat localization

The thermal distillation uses energy to evaporate water and subsequently condense the vapor to liquid form. Waste energy from industrial processes, electricity, or solar energy are the examples of an input energy for thermal distillation systems. In this paper, we have designed, fabricated, and tested an eco-friendly and low- cost distillation system which uses solar heat localization for evaporation of water and later condense the vapor back to liquid as liquid water. The unique feature of this method is the high rate of evaporation when compared to other thermal solar distillation/desalination systems. In this method, the incoming heat is confined to the porous media where bottom layer acts a porous insulator. The capillary force in the porous structure pumps up the liquid water to the hot surface, within the porous medium, where liquid heats up and changes to vapor phase. Since there is no moving part in this method, and due to its independency to electricity, it a great candidate for deployment in remote areas that access to grid power is limited and maintenance is not feasible.

Jaladi, Divya↗
DOE OSTI2016

Novel Metal Sulfides to Achieve Effective Capture and Durable Consolidation of Radionuclides

This report documents the work done under NEUP grant to examine the capability of novel chalcogels and some binary metal chalcogenides as a host matrix for the capture of gaseous iodine and the feasibility of their iodine-laden materials to be converted into a permanent waste form. The presented work was conducted over last two years. A number of novel chalcogels Zn 2 Sn 2 S 6 , Sb 4 Sn 4 S 12 , NiMoS 4 , CoMoS 4 , antimony sulfide (SbS x ) chalcogels, silver functionalized chalcogels and binary metal sulfides (Sb 2 S 3 ) were developed and studies for their iodine absorption efficacies. A new and simple route was devised for the large scale preparation of antimony sulfide chalcogel. The chalcogel was obtained by treating Sb 2 S 3 with Na 2 S in the presence of water followed by addition of formamide. The obtained gels have a low-density sponge like network of meso porous nature having BET surface area of 125 m 2 /g. The chalcogels, silver functionalized chalcogel and the binary metal sulfides were exposed to iodine vapors in a closed container. Silver-functionalized chalcogels and Sb 2 S 3 powders showed iodine uptake up to 100 wt%, the highest iodine uptake of 200 wt% was observed for the SbS-III chalcogel. The PXRD patterns of iodine-laden specimens revealed that iodine shows spontaneous chemisorption to the matrix used. The iodine loaded chalcogels and the binary chalcogenides were sealed under vacuum in fused silica ampoules and heated in a temperature controlled furnace. The consolidated products were analyzed by PXRD, energy dispersive spectroscopy (EDS), UV-Vis and Raman spectroscopy. The final products were found to be amorphous in most of the cases with high amount (~4-35 wt%) of iodine and aapproximately ~60- 90 % of the absorbed iodine could be consolidated into the final waste form. Alginate reinforced composite scaffolds with SbS/SnS chalcogels and Sb 2 S 3 bulk powder were also fabricated aiming to study their efficacy as host matrices in capturing the gaseous molecular iodine in dynamic mode from nuclear spent fuel. The obtained composites looks robust in comparison to their respective pristine chalcogels and Sb 2 S 3 bulk powder.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Preliminary Evaluation of Removing Spent Nuclear Fuel From Shutdown Sites - Fort Calhoun Site Visit - 18048

This is a technical paper that does not take into account contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. An updated evaluation of removing spent nuclear fuel (SNF) from 14 shutdown nuclear power plant sites was conducted during 2017. The major update since 2016 was the addition of the recently shutdown Fort Calhoun site to the evaluation. The Fort Calhoun site is located on the western shore of the Missouri River in Washington County in eastern Nebraska, about 31 km north of Omaha, Nebraska. Fort Calhoun has been shut down since October 24, 2016 and final removal of SNF from the reactor vessel was completed on November 13, 2016. The site visit to Fort Calhoun occurred in May 2017 and involved meetings with the Fort Calhoun staff and the Union Pacific Railroad. Participants in the site visit included the Federal Railroad Administration, Pahrump Paiute Tribe, Prairie Island Indian Community, Nebraska State Patrol, Nebraska Department of Health and Human Services, Iowa Department of Transportation Office of Rail Transportation, Iowa Department of Transportation Office of Motor Vehicle Enforcement, and the Council of State Governments-Midwest. The Fort Calhoun site was found to have two transportation mode options for the removal of SNF, rail and barge. In the case of rail, an onsite rail spur could be reinstalled from the existing Cargill industrial spur that runs through the Fort Calhoun site, or an onsite trans-load onto the Cargill industrial spur could be performed. In 2006, a barge was used to ship steam generators, a pressurizer, and a reactor vessel head to the Fort Calhoun site; consequently, barge transport of SNF offsite is potentially an option. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

SRS Sludge Batch Qualification and Processing: Historical Perspective and Lessons Learned

This report provides a historical overview and lessons learned associated with the SRS sludge batch (SB) qualification and processing programs. The report covers the framework of the requirements for waste form acceptance, the DWPF Glass Product Control Program (GPCP), waste feed acceptance, examples of how the program complies with the specifications, an overview of the Startup Program, and a summary of continuous improvements and lessons learned. The report includes a bibliography of previous reports and briefings on the topic.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

A Robust Power Remote Manipulator for Use in Waste Sorting, Processing, and Packaging - 12158

Disposition of radioactive waste is one of the Department of Energy's (DOE's) highest priorities. A critical component of the waste disposition strategy is shipment of Transuranic (TRU) waste from DOE's Oak Ridge Reservation to the Waste Isolation Plant Project (WIPP) in Carlsbad, New Mexico. This is the mission of the DOE TRU Waste Processing Center (TWPC). The remote-handled TRU waste at the Oak Ridge Reservation is currently in a mixed waste form that must be repackaged in to meet WIPP Waste Acceptance Criteria (WAC). Because this remote-handled legacy waste is very diverse, sorting, size reducing, and packaging will require equipment flexibility and strength that is not possible with standard master-slave manipulators. To perform the wide range of tasks necessary with such diverse, highly contaminated material, TWPC worked with S.A. Technology (SAT) to modify SAT's Power Remote Manipulator (PRM) technology to provide the processing center with an added degree of dexterity and high load handling capability inside its shielded cells. TWPC and SAT incorporated innovative technologies into the PRM design to better suit the operations required at TWPC, and to increase the overall capability of the PRM system. Improving on an already proven PRM system will ensure that TWPC gains the capabilities necessary to efficiently complete its TRU waste disposition mission. The collaborative effort between TWPC and S.A. Technology has yielded an extremely capable and robust solution to perform the wide range of tasks necessary to repackage TRU waste containers at TWPC. Incorporating innovative technologies into a proven manipulator system, these PRMs are expected to be an important addition to the capabilities available to shielded cell operators. The PRMs provide operators with the ability to reach anywhere in the cell, lift heavy objects, perform size reduction associated with the disposition of noncompliant waste. Factory acceptance testing of the TWPC Powered Remote Manipulators has completed at SAT's Colorado facility, and on-site training at TWPC is scheduled to start in early 2012. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} for Fukushima Daiichi Water Treatment Secondary Wastes - 20212

The Japanese government is supporting development work implemented jointly by Veolia subsidiaries Kurion Japan, K.K., Veolia Nuclear Solutions (VNS), Inc., and Veolia Nuclear Solutions Federal Services, LLC for treating radioactive waste generated from Fukushima Daiichi Nuclear Power Station (NPS) water treatment using the GeoMelt{sup TM} In-Container Vitrification (ICV){sup TM} technology. The initial work consisted of glass formulation and engineering-scale testing which was completed in 2018, in the frame of an IRID (International Research Institute for Nuclear Decommissioning) program as part of a project subsidized by Japan's Ministry of Economy, Trade and Industry (METI). The Fukushima Daiichi NPS Mid- and Long-Term Road-map requires investigation of methods to stabilize solid wastes (and to immobilize radioisotopes in the wastes) generated as a result of emergency response and decommissioning activities. Cooling water treatment has resulted in a significant amount of solid and slurry secondary wastes (mostly adsorbents and ion-exchange materials) which will require processing at some point. GeoMelt{sup R} ICV{sup TM} is a joule-heated melter technology which uses a refractory-lined single-use container combining the melter and disposal container. There is no pouring required nor concerns with refractory corrosion which allows the process to accommodate a wide range of waste chemistries and high waste loadings. The testing described here consisted of three engineering-scale melts, each processing between 212 kg and 240 kg of waste simulants, glass formers, and non-radioactive cesium (Cs) and strontium (Sr) tracers. Continuous isokinetic stack sampling of off-gas emissions was performed for each test in order to calculate Cs and Sr retention in the glass wasteform. Single-pass retention of Cs in the final glass wasteform ranged from 91.46 to 99.30%, and single-pass retention of Sr ranged from 99.76 to 100%. Planned particulate recycle will increase these retention levels. Melt 1 processed a mixture of KUR-EH (a zeolite-based ion-exchange material), simulated Advanced Liquid Processing System (ALPS) Carbonate and Iron Slurries, and glass additives. Melt 2 processed a mixture of KUR-EH, KUR-TSG (a titanate-based adsorbent), and glass additives. Melt 3 processed a mixture of KUR-EH, simulated barium sulfate/iron ferrocyanide sludge (AREVA sludge), and glass additives. Waste loadings for these melts ranged from 70 weight percent (wt%) to 82 wt%. Vitrification produces a waste form much denser than the stored water treatment secondary waste wastes, resulting in significant volume reduction. Volume reductions for the three tests ranged from 74 to 79 vol%. Vitrification produces a chemically durable wasteform. Pacific Northwest National Laboratory (PNNL) tested three glass samples from each engineering-scale melt) by the Materials Characterization Center 1 (MCC-1) test, an international standard leach test of the chemical durability of nuclear waste glasses. PNNL also obtained one U.S. reference glass (EA Glass) and two Japan reference glasses P0798) and tested these under the same MCC-1 conditions (90 deg. C, 10 m-1, DIW, and 7, 14, 28-day) as the GeoMelt{sup R} ICV{sup TM} glasses. The GeoMelt{sup R} ICV{sup TM} glasses exhibited lower total normalized releases and 14- to 28-day normalized release rates than the three reference glasses. These results suggest that the GeoMelt{sup R} ICV{sup TM} glasses have durabilities on par with high-level waste glasses under standard test conditions. Post-melt process sampling and analysis indicated no Cs migration into the melter refractory materials and very little deposition of Cs or Sr onto the melter hood or off-gas piping internals. The results of the testing indicated good Cs retention in the glass, high volume reduction and waste loadings, and excellent chemical durability. These factors are important to minimize treatment costs and to protect workers and the environment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Radioactive waste immobilization of Hanford sludge in magnesium potassium phosphate ceramic forms

Here, this paper evaluates immobilization of Hanford K-Basin tank sludge in magnesium potassium phosphate ceramic forms. The waste forms were produced using two simulated non-radioactive sludge streams, each with distinct characteristics and composition. Ceramicrete with wollastonite as filler was used as the matrix for this purpose. The resulting waste forms were tested for their mechanical properties, radiation stability, and leaching resistance. In another series of tests, the samples were vitrified in borosilicate glass and glass waste forms were produced. The Product Consistency Test, the American Nuclear Society's ANS 16.1 test, and the Toxicity Characteristic Leaching Procedure, which are used to develop waste acceptance criteria in the United States for permanent storage of treated waste, were used for evaluation of the leaching resistance of all waste forms.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗