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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Analysis of Legacy {sup 85}Kr Waste Form Samples - 16319

In the late 1970's, the Idaho Chemical Processing Plant conducted an R and D program on krypton encapsulation. Legacy samples composed of {sup 85}Kr incorporated in solid zeolite 5A material, and five small metal tubes containing a mixture of the zeolite combined with a glass matrix resulting from hot isostatic pressing, have been preserved. These samples were shipped to Oak Ridge National Laboratory (ORNL) in mid-FY 2014. After receipt and repackaging, the samples were examined by x-ray imaging and x-ray tomography. Two of the capsules appear to contain an amorphous mass within the capsules. An image of one capsule clearly shows the saw marks on the capsule and a quantity of loose pellet or bead-like material remaining in the capsule. An image of another capsule shows similar bead-like material within the intact capsule. One of the capsules was previously opened. The end of this capsule appears to have been cut off, and there are additional saw marks on the side of the capsule. X-ray tomography allowed the capsules to be viewed along the three axes. Of most interest was determining whether there was any residual material in the closed end of this capsule. The images confirmed the presence of residual material within this capsule. The material appears to be compacted but still retains some of the bead-like morphology. Based on the initial nondestructive assay (NDA) and the fact that there are two obviously breached samples, a plan was developed for exploratory tests that would be conducted with the breached specimens before opening the three intact capsules. Preparations are being made to determine the fraction of krypton/xenon remaining in the waste form matrix and the amount of rubidium remaining in the waste form matrix. The inner surface of the breached capsules will be examined for corrosion. (authors)

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

International Collaborations on Radioactive Waste Disposal in Salt

This report is a summary of the international collaboration and laboratory work funded by the US Department of Energy Office (DOE) of Nuclear Energy Spent Fuel and Waste Science & Technology (SFWST) as part of the Sandia National Laboratories Salt R&D work package. This report satisfies milestone level-four milestone M4SF-19SNO10303064. Several stand-alone sections make up this summary report, each completed by the participants. The first two sections discuss international collaborations on geomechanical benchmarking exercises (WEIMOS), granular salt reconsolidation (KOMPASS), engineered barriers (RANGERS), and documentation of Features, Events, and Processes (FEPs).

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

SRNL Phase 1 Assessment of the WAC/DQO and Unit Operations for the WTP Waste Qualification Program

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is currently transitioning its emphasis from a design and construction phase toward start-up and commissioning. With this transition, the WTP Project has initiated more detailed assessments of the requirements related to actual processing of the Hanford Site tank waste. One particular area of interest is the waste qualification program to be implemented to support the WTP. Given the successful implementation of similar waste qualification efforts at the Savannah River Site (SRS), based on critical technical support and guidance from the Savannah River National Laboratory (SRNL), WTP requested the utilization of subject matter experts from SRNL to support a technology exchange to perform a review of the WTP waste qualification program, discuss the general qualification approach at SRS, and to identify critical lessons learned through the support of DWPF's sludge batch qualification efforts. As part of Phase 1, SRNL subject matter experts in critical technical and/or process areas reviewed specific WTP waste qualification information. The Phase 1 review was a collaborative, interactive, and iterative process between the two organizations. WTP provided specific analytical procedures, descriptions of equipment, and general documentation as baseline review material. SRNL subject matter experts reviewed the information and, as appropriate, requested follow-up information or clarification to specific areas of interest. This process resulted in multiple teleconferences with key technical contacts from both organizations resolving technical issues that lead to the results presented in this report. This report provides the results of SRNL's Phase 1 review of the WAC-DQO waste acceptance criteria and processability parameters, and the specific unit operations which are required to support WTP waste qualification efforts. The review resulted in SRNL providing concurrence, alternative methods, or gap identification for the proposed WTP analytical methods or approaches. For the unit operations, the SRNL subject matter experts reviewed WTP concepts compared to what is used at SRS and provided thoughts on the outlined tasks with respect to waste qualification. Also documented in this report are recommendations and an outline on what would be required for the next phase to further mature the WTP waste qualification program.

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

In Situ Cold-Spray Repair of Environmental Mission Infrastructure for Component Life Extension - 19424

Cold spray offers a potential option for performing in situ repair of U.S. Department of Energy (DOE) complex infrastructure critical to the Office of Environmental Management cleanup mission. Application to the DOE complex has the potential to reduce cost, schedule, and safety risk impacts associated with component failures and would negate the need to procure and construct replacement infrastructure. A concept and technical basis are presented for developing remotely deployable cold-spray systems for performing in situ repairs in hazardous environments such as nuclear waste tanks, piping, and process vessels to extend component service life or return critical infrastructure to service. (authors)

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

Iron Phosphate Glass for Vitrifying Hanford AZ102 LAW in Joule Heated and Cold Crucible Induction Melters - 12240

An iron phosphate composition for vitrifying a high sulfate (∼17 wt%) and high alkali (∼80 wt%) Hanford low activity waste (LAW), known as AZ-102 LAW, has been developed for processing in a Joule Heated Melter (JHM) or a Cold Crucible Induction Melter (CCIM). This composition produced a glass waste form, designated as MS26AZ102F-2, with a waste loading of 26 wt% of the AZ-102 which corresponded to a total alkali and sulfate (represented as SO{sub 3}) content of 21 and 4.4 wt%, respectively. A slurry (7 M Na{sup +}) of MS26AZ102F-2 simulant was melted continuously at temperatures between 1030 and 1090 deg. C for 10 days in a small JHM at PNNL and for 70 hours in a CCIM at INL. The as-cast glasses produced in both melters and in trial laboratory experiments along with their canister centerline cooled (CCC) counterparts met the requirements for the Product Consistency Test (PCT) and the Vapor Hydration Test (VHT) responses in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract. These glass waste forms retained up to 77 % of the SO{sub 3} (3.3 wt%), 100% of the Cesium, and 33 to 44% of the rhenium (used as a surrogate for Tc) all of which either exceeded or were comparable to the retention limit for these species in borosilicate glass nuclear waste form. Analyses of commercial K-3 refractory lining and the Inconel 693 metal electrodes used in JHM indicated only minimum corrosion of these components by the iron phosphate glass. This is the first time that an iron phosphate composition was melted continuously in a slurry fed JHM and in the US, thereby, demonstrating that iron phosphate glasses can be used as alternative hosts for vitrifying nuclear waste. The following conclusions are drawn from the results of the present work. (1) An iron phosphate composition, designated as MS26AZ102F-2, containing 26 wt% of the simulated high sulfate (17 wt%), high alkali (80 wt%) Hanford AZ-102 LAW meets all the criteria for processing in a JHM and CCIM. This composition produces a homogeneous glass with a density of 2.80 ± 0.04 g/cm{sup 3} after melting between 1000 and 1050 deg. C for 3 to 5 h. (2) This is the first time that an iron phosphate glass was melted in the JHM continuously for 10 days achieving a specific melting rate of 1010 kg/m2/day and in the CCIM for 70 hrs with a melting rate of 664 kg/m{sup 2}/day. (3) The analyzed (ICP-AES) compositions for all the glasses prepared under several melting conditions are in excellent agreement with the target composition. The variation of melting conditions include: (i) use of small scale/short melting time to large scale/long melting time (300 g to 80 kg, 4 h to 10 days) operations including melting in the JHM and CCIM, (ii) use of dry or wet (slurry) melter feed, (iii) addition of reductant (sugar) in the batch, and (iv) bubbling the melt with air. (4) The chemical durability as measured by PCT and VHT for the quenched and CCC-treated waste forms prepared from laboratory, JHM or CCIM melting exceeds the DOE requirements for LAW. (5) Depending upon the melting time (4 h to 10 days), the average concentration of SO{sub 3} in the MS26AZ102F-2 iron phosphate glass waste form varies from about 1.78 (RSM) to 3.74 (laboratory melt) wt% which corresponds to an SO{sub 3} retention of 41 to 86% when melted between 1030 at 1050 deg. C in air. (6) The retention of SO{sub 3} in the glass was reduced when a reductant (sugar) was added to the slurry. (7) The retention of other problem components like Cesium and Rhenium/Tc-99 is also high; from 92 to 100% for Cs{sub 2}O under different melting conditions, and from 66 to 33% for Re{sub 2}O{sub 7} for melts processed from 3 to 5 hours and up to 10 days. (8) Corrosion tests on Inconel 693 and K-3 refractory at temperatures between 1000 and 1050 deg. C indicate that both materials should be suitable for melting iron phosphate glasses. (9) The viscosity and electrical conductivity of the MS26AZ102F-2 melt are within the acceptable limits for RSM and CCIM processing. (10)This work has shown that iron phosphate compositions are a potential alternative to borosilicate glasses for vitrifying problematic wastes such as Hanford AZ-102 LAW that contains high amounts sulfate and alkali oxides, and that iron phosphate melts such as MS26AZ102F-2 can be processed in the JHM or CCIM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Rainier Mesa/Shoshone Mountain Flow and Transport Model Report Nevada National Security Site, Nevada, Revision 1

The purpose of this report is to describe and present the latest CAI analyses for the Rainier Mesa/Shoshone Mountain (RM/SM) CAU and review their role in the broader, historical context of CAI activities at RM/SM. As described in Sections 1.4 and 1.5, the current work was initiated after specific concerns arose during a previous CAI phase, which eventually prompted a revision of the FFACO process and the adoption of an alternative modeling strategy (AMS) for CAI work at RM/SM. The remainder of this report will describe the CAI modeling activities conducted through 2017 under the revised FFACO process and AMS discussed above.

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DOE OSTI2013

Sulfur-Modified Zero-Valent Iron for Remediation Applications at DOE Sites - 13600

Many DOE remediation sites have chemicals of concern that are compounds in higher oxidation states, which make them both more mobile and more toxic. The chemical reduction of these compounds both prevents the migration of these chemicals and in some cases reduces the toxicity. It has also been shown that zero-valent iron is a very effective substance to use in reducing oxygenated compounds in various treatment processes. These have included the treatment of halogenated hydrocarbons in the form volatile organic compounds used as solvents and pesticides. Zero-valent iron has also been used to reduce various oxidized metals such as chromium, arsenic, and mercury in order to immobilize them, decrease their toxicity, and prevent further transport. In addition, it has been used to immobilize or break down other non-metallic species such as selenium compounds and nitrates. Of particular interest at several DOE remediation sites is the fact that zero-valent iron is very effective in immobilizing several radioactive metals which are mobile in their oxidized states. These include both technetium and uranium. The main difficulty in using zero-valent iron has been its tendency to become inactive after relatively short periods of time. While it is advantageous to have the zero-valent iron particles as porous as possible in order to provide maximum surface area for reactions to take place, these pores can become clogged when the iron is oxidized. This is due to the fact that ferric oxide has a greater volume for a given mass than metallic iron. When the surfaces of the iron particles oxidize to ferric oxide, the pores become narrower and will eventually shut. In order to minimize the degradation of the chemical activity of the iron due to this process, a modification of zero-valent iron has been developed which prevents or slows this process, which decreases its effectiveness. It is called sulfur-modified iron, and it has been produced in high purity for applications in municipal water treatment applications. Sulfur-modified iron has been found to not only be an extremely economical treatment technology for municipal water supplies, where very large quantities of water must be treated economically, but it has also been demonstrated to immobilize technetium. It has the added benefit of eliminating several other harmful chemicals in water supplies. These include arsenic and selenium. In one large-scale evaluation study an integrated system implemented chemical reduction of nitrate with sulfur-modified iron followed by filtration for arsenic removal. The sulfur-modified iron that was used was an iron-based granular medium that has been commercially developed for the removal of nitrate, co-contaminants including uranium, vanadium and chromium, and other compounds from water. The independent study concluded that 'It is foreseen that the greatest benefit of this technology (sulfur-modified iron) is that it does not produce a costly brine stream as do the currently accepted nitrate removal technologies of ion exchange and reverse osmosis. This investigation confirmed that nitrate reduction via sulfur-modified iron is independent of the hydraulic loading rate. Future sulfur-modified iron treatment systems can be designed without restriction of the reactor vessel dimensions. Future vessels can be adapted to existing site constraints without being limited to height-to-width ratios that would exist if nitrate reduction were to depend on hydraulic loading rate'. Sulfur-modified iron was studied by the Pacific Northwest National Laboratory (PNNL) for its effectiveness in the reduction and permanent sequestration of technetium. The testing was done using Hanford Site groundwater together with sediment. The report stated, 'Under reducing conditions, TcO{sub 4} is readily reduced to TcIV, which forms highly insoluble oxides such at TcO{sub 2}.nH{sub 2}O. However, (re)oxidation of TcIV oxides can lead to remobilization. Under sulfidogenic conditions, most TcIV will be reduced and immobilized as Tc{sub 2}S{sub 7}, which is less readily re-mobilized, even under oxic conditions. This process should be favored by stimulation of sulfidogenic conditions'. The sulfur-modified iron provides the sulfur, together with the iron, to maintain this stable sequestration of technetium. As a result of these and other studies demonstrating the cost-effectiveness of sulfur-modified iron in treating technetium and other hazardous compounds in Hanford Site groundwater and its cost-effectiveness in reducing nitrate, the Richland Operations Office of the Department of Energy issued a change order to the Central Plateau Contractor providing for the testing of sulfur-modified iron in a mobile pilot unit at the Hanford Site. Further testing is anticipated to produce refinements in operating conditions and further optimization of the existing process. (authors)

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DOE OSTI2015

AREVA TN Universal Centralized Storage System - 15581

AREVA TN has developed a new used nuclear fuel storage product concept: NUHOMS{sup R} Universal Storage Facility (USF) to meet the needs and capabilities of centralized storage. This NUHOMS{sup R} USF is a horizontal system retaining the superior seismic stability, shielding performance, and ease of canister retrieval of the existing NUHOMS{sup R} system, while creating a single, simplified storage module that can store canisters of any sizes, including horizontal NUHOMS{sup R} canisters as well canisters from vertical storage systems or canisters developed for future direct geologic storage. The NUHOMS{sup R} USF has a reduced footprint and uses less material compared to existing NUHOMS{sup R} system. The canister handling process has been renovated to simplify the canister loading process. (authors)

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DOE OSTI2020

Sources of Conservatism in Making Corrective Action Decisions at the NNSS - 20477

During the course of closing nuclear testing release sites at the Nevada National Security Site (NNSS) under the Federal Facility Agreement and Consent Order (FFACO) using the Risk-Based Corrective Action (RBCA) evaluation process, conservative assumptions and estimates were used in many phases of the work. These assumptions and estimates were agreed to during Data Quality Objective (DQO) meetings and published in the FFACO documents. However, the potential cumulative effect of these multiple sources of conservatism was not evaluated. The RBCA process ensures that corrective action decisions minimize the risk of making Type I errors (not applying corrective action to areas that exceed action levels) at the expense of making Type II errors (applying corrective action to areas that do not exceed action levels). Rather than fully characterizing the spatial distribution of contaminants throughout each site, the RBCA process uses a bounding methodology that conservatively estimates spatial boundaries beyond which contamination is not present that exceed corrective action levels. This process accepts that these boundaries may encompass areas that do not exceed corrective action levels and, therefore, may be somewhat larger than for fully characterized sites. This is justified at sites where the reduction in investigation costs are more than any potential increase in corrective action costs. To minimize the risk of making Type I errors, conservatism was used throughout the process. This study documents the many sources of conservatism used in the RBCA process. Conservatism was used in the RBCA process in determining potential receptors, determining action levels, collecting samples, converting sample results into receptor dose, and accounting for uncertainty. This provides decision-makers with a high level of confidence that unacceptable risk to receptors will not result from exposure to residual contamination outside the corrective action boundaries. However, it does not provide an estimate of the actual extent of areas exceeding action levels. This study provides an evaluation of the effects of compounding conservatism on RBCA process decisions. (authors)

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DOE OSTI2018

Sensitivity study of a full-scale industrial spray-injected fluidized bed reactor

The industrial fluidized bed reactor (FBR) described here is designed to convert an aqueous solid laden stream into a consistent solid product material. Additionally, the FBR is heated to 650 °C and the aqueous reactant is spray-atomized into the fluidized bed, converting it into a granular solid through a series of heterogeneous chemical reactions. The objective of this study is to use the MFiX two-fluid model as a screening tool to investigate the effects of off-nominal conditions for the purpose of establishing performance guidelines. This information will be helpful to assess whether the identified boundaries of the system are steep or gradual. Initially, 13 independent operating parameters were sampled in a one-at-a-time manner using high, low, and base values and ranked against a set of performance criteria. Five operating parameters were found to have the largest effect (bed particle size, bed particle density, coal particle size, spray feed flow rate, and fluidizing gas flow rate) on three quantities of interest—bed differential temperature, solids velocity, and bed voidage. Latin hypercube sampling was used to generate a minimal number of random values for various combinations of input parameters. The spray feed flow rate was the most significant parameter for the temperature differences, and the coal particle and bed particle size were significant contributors. The bed particle size had the largest effect on the low velocity of the bed particles, with the coal particle size as a contributing second effect. The high solid packing shows the coal particle size with the largest effect, and significant secondary contributions from the feed flow rate and fluidizing gas flow rate. Finally, the fits to the five-dimensional Gaussian Process models were 0.7797, 0.8664, and 0.9440 for the temperature, velocity, and solids packing respectively.

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

Testing of Optimized Bubbler Configuration for HLW Melter

The principal objective of this work was to determine the glass production rate increase and ancillary effects of adding more bubbler outlets to the current WTP HLW melter baseline. This was accomplished through testing on the HLW Pilot Melter (DM1200) at VSL. The DM1200 unit was selected for these tests since it was used previously with several HLW waste streams including the four tank wastes proposed for initial processing at Hanford. This melter system was also used for the development and optimization of the present baseline WTP HLW bubbler configuration for the WTP HLW melter, as well as for MACT testing for both HLW and LAW. Specific objectives of these tests were to: Conduct DM1200 melter testing with the baseline WTP bubbling configuration and as augmented with additional bubblers. Conduct DM1200 melter testing to differentiate the effects of total bubbler air flow and bubbler distribution on glass production rate and cold cap formation. Collect melter operating data including processing rate, temperatures at a variety of locations within the melter plenum space, melt pool temperature, glass melt density, and melter pressure with the baseline WTP bubbling configuration and as augmented with additional bubblers. Collect melter exhaust samples to compare particulate carryover for different bubbler configurations. Analyze all collected data to determine the effects of adding more bubblers to the WTP HLW melter to inform decisions regarding future lid re-designs. The work used a high aluminum HLW stream composition defined by ORP, for which an appropriate simulant and high waste loading glass formulation were developed and have been previously processed on the DM1200.

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

Modeling Solute Thermokinetics in LiCl-KCl Molten Salt for Nuclear Waste Separation (NEUP 10-418)

Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt at the eutectic composition (58 mol% LiCl, 42 mol% KCl), which is used for treating spent EBR-II fuel. The same process being used for EBRII fuel is currently being studied for widespread international implementation. The methods will focus on first-principles and first- principles derived interatomic potential based simulations, primarily using molecular dynamics. Results will be validated against existing literature and parallel ongoing experimental efforts. The simulation results will be of value for interpreting experimental results, validating analytical models, and for optimizing waste separation by potentially developing new salt configurations and operating conditions.

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DOE OSTI2023

Summary of SRNL Support to the DOE-ORP Enhanced Waste Glass Program for FY23

In fiscal year 2023 (FY23) Savannah River National Laboratory (SRNL) continued tasked work for the Office of River Protection (ORP) to expand glass compositional regions accessible for low-activity waste (LAW) and high-activity waste (HLW) vitrification processing. Experimental work continued in four primary technical areas focused on processing and performance of glasses relevant to the Hanford missions. The data and results from this work will be used to expand and validate the glass models being developed at Pacific Northwest National Laboratory (PNNL) for waste processing and acceptance. This report summarizes the activities and deliverables associated with work performed in FY23.

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DOE OSTI2015

Toward environmentally-benign utilization of nonmetallic fraction of waste printed circuit boards as modifier and precursor

Highlights: • Environmental impacts of electronic waste and specifically waste printed circuit boards. • Review of the recycling techniques of waste printed circuit boards. • Advantages of physico-mechanical recycling techniques over chemical methods. • Utilization of nonmetallic fraction of waste printed circuit boards as modifier/filler. • Recent advances in the use of nonmetallic fraction of waste printed circuit boards as precursor. - Abstract: Electronic waste, including printed circuit boards, is growing at an alarming rate due to the accelerated technological progress and the shorter lifespan of the electronic equipment. In the past decades, due to the lack of proper economic and environmentally-benign recycling technologies, a major fraction of e-waste generated was either destined to landfills or incinerated with the sole intention of its disposal disregarding the toxic nature of this waste. Recently, with the increasing public awareness over their environment and health issues and with the enaction of more stringent regulations, environmentally-benign recycling has been driven to be an alternative option partially replacing the traditional eco-unfriendly disposal methods. One of the most favorable green technologies has been the mechanical separation of the metallic and nonmetallic fraction of the waste printed circuit boards. Although metallic fraction, as the most profitable component, is used to generate the revenue of the separation process, the nonmetallic fraction (NMF) has been left isolated. Herein, the recent developments in the application of NMF have been comprehensively reviewed and an eco-friendly emerging usage of NMF as a value-added material for sustainable remediation has been introduced.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

MANAGING BERYLLIUM IN NUCLEAR FACILITY APPLICATIONS

Beryllium plays important roles in nuclear facilities. Its neutron multiplication capability and low atomic weight make it very useful as a reflector in fission reactors. Its low atomic number and high chemical affinity for oxygen have led to its consideration as a plasma-facing material in fusion reactors. In both applications, the beryllium and the impurities in it become activated by neutrons, transmuting them to radionuclides, some of which are long-lived and difficult to dispose of. Also, gas production, notably helium and tritium, results in swelling, embrittlement, and cracking, which means that the beryllium must be replaced periodically, especially in fission reactors where dimensional tolerances must be maintained. It has long been known that neutron activation of inherent iron and cobalt in the beryllium results in significant {sup 60}Co activity. In 2001, it was discovered that activation of naturally occurring contaminants in the beryllium creates sufficient {sup 14}C and {sup 94}Nb to render the irradiated beryllium 'Greater-Than-Class-C' for disposal in U.S. radioactive waste facilities. It was further found that there was sufficient uranium impurity in beryllium that had been used in fission reactors up to that time that the irradiated beryllium had become transuranic in character, making it even more difficult to dispose of. In this paper we review the extent of the disposal issue, processes that have been investigated or considered for improving the disposability of irradiated beryllium, and approaches for recycling.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Shipping Grossly Dewatered resins in the RT-100 Type B Transportation Cask - 16068

Ion-exchange resins are used in the purification of coolant water in boiling water reactor nuclear power plants. The spent resins are a form of low-level or intermediate-level radioactive waste. These resins may have a relatively high specific activity level and are normally transported to a disposal facility within a secondary container placed in a Type B transportation cask. During transportation, hydrogen gas may build up in the cask due to radiolysis of the hydrogenous material in the contents. In the licensing process, a Type B cask is carefully analyzed to ensure that the maximum amount of hydrogen gas generated does not exceed 5%. The calculation to determine the mole fraction of hydrogen gas generated is a function of many variables, including the properties of the contents and the volume ratios of the contents. Because water generates hydrogen gas when it undergoes radiolysis, the amount of water present in the waste is normally limited. Typically, Type B transportation casks are allowed to carry dewatered resins, meaning less than 1% free-standing water by volume. To ensure compliance with this limitation, the dewatering process requires specialized equipment to pump the water from the secondary container. In an effort to enhance the usability of the RT-100 Type B transportation cask, Robatel Technologies re-analyzed the hydrogen gas generation calculation for contents including 'grossly dewatered' resins. Grossly dewatered resins are defined as resins containing less than 20% free-standing water by volume. This higher limit is made possible by a more thorough hydrogen gas generation calculation, which ensures that the flammability limit is not reached during shipment. In July 2015, The USNRC issued Revision 1 of the Certificate of Compliance for the Robatel Technologies RT-100 Type B transportation cask. This revision allows the cask to be used for grossly dewatered resins. With the revised Certificate of Compliance, the RT-100 is the first of its kind whose authorized contents include greater than 1% free liquids by volume. As a result of this change, the cask user can meet the 20% maximum free-standing water volume with a much simpler gravity-driven dewatering process. It is no longer necessary to have a complex dewatering system to ensure compliance with regulatory requirements. (authors)

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

Brine flow in heated geologic salt.

This report is a summary of the physical processes, primary governing equations, solution approaches, and historic testing related to brine migration in geologic salt. Although most information presented in this report is not new, we synthesize a large amount of material scattered across dozens of laboratory reports, journal papers, conference proceedings, and textbooks. We present a mathematical description of the governing brine flow mechanisms in geologic salt. We outline the general coupled thermal, multi-phase hydrologic, and mechanical processes. We derive these processes governing equations, which can be used to predict brine flow. These equations are valid under a wide variety of conditions applicable to radioactive waste disposal in rooms and boreholes excavated into geologic salt.

Kuhlman, Kristopher L.↗
DOE OSTI2015

Coupling of Nuclear Waste Form Corrosion and Radionuclide Transports in Presence of Relevant Repository Sediments

Assessments of waste form and disposal options start with the degradation of the waste forms and consequent mobilization of radionuclides. Long-term static tests, single-pass flow-through tests, and the pressurized unsaturated flow test are often employed to study the durability of potential waste forms and to help create models that predict their durability throughout the lifespan of the disposal site. These tests involve the corrosion of the material in the presence of various leachants, with different experimental designs yielding desired information about the behavior of the material. Though these tests have proved instrumental in elucidating various mechanisms responsible for material corrosion, the chemical environment to which the material is subject is often not representative of a potential radioactive waste repository where factors such as pH and leachant composition will be controlled by the near-field environment. Near-field materials include, but are not limited to, the original engineered barriers, their resulting corrosion products, backfill materials, and the natural host rock. For an accurate performance assessment of a nuclear waste repository, realistic waste corrosion experimental data ought to be modeled to allow for a better understanding of waste form corrosion mechanisms and the effect of immediate geochemical environment on these mechanisms. Additionally, the migration of radionuclides in the resulting chemical environment during and after waste form corrosion must be quantified and mechanisms responsible for migrations understood. The goal of this research was to understand the mechanisms responsible for waste form corrosion in the presence of relevant repository sediments to allow for accurate radionuclide migration quantifications. The rationale for this work is that a better understanding of waste form corrosion in relevant systems will enable increased reliance on waste form performance in repository environments and potentially decrease the need for expensive engineered barriers.Our current work aims are 1) quantifying and understanding the processes associated with glass alteration in contact with Fe-bearing materials; 2) quantifying and understanding the processes associated with glass alteration in presence of MgO (example of engineered barrier used in WIPP); 3) identifying glass alteration suppressants and the processes involved to reach glass alteration suppression; 4) quantifying and understanding the processes associated with Saltstone and Cast Stone (SRS and Hanford cementitious waste forms) in various representative groundwaters; 5) investigating positron annihilation as a new tool for the study of glass alteration; and 6) quantifying and understanding the processes associated with glass alteration under gamma irradiation.

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