Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIOACTIVE WASTE PROCESSING”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Office of River Protection Advanced Low-Activity Waste Glass Research and Development Plan

The U.S. Department of Energy Office of River Protection (ORP) has initiated and leads an integrated Advanced Waste Glass (AWG) program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product performance requirements. The integrated ORP program is focused on providing a technical, science-based foundation for making key decisions regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities in the context of an optimized River Protection Project (RPP) flowsheet. The fundamental data stemming from this program will support development of advanced glass formulations, key product performance and process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste vitrification facilities. These activities will be conducted with the objective of improving the overall RPP mission by enhancing flexibility and reducing cost and schedule.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

A Comparison of Social and Economic Engagement Approaches: Lessons from Recent International Radioactive Waste Projects - 17478

The central challenge of any proposed radioactive waste management project is securing not only regulatory permissions, but also a social license to operate. 'Social license' generally refers to a community's acceptance or approval of a specific project. Oftentimes development of social license occurs outside of, or in parallel to, formal regulatory or permitting processes, and requires sustained commitment and proposed economic benefits to communities by proponents to acquire, build, and sustain social capital in the context of trust-based relationships. Historically, radioactive waste management projects have defaulted to demagoguery and 'siege warfare' modes of thinking and action by both proponents and opponents. While radioactive waste management projects are always inherently controversial, it is only recently that project owners and developers have experimented with alternative forms of social engagement outside of, or in addition to, formal regulatory processes. This paper compares and contrasts recent projects in Canada, Australia, the United Kingdom, and the United States, by examining distinct approaches to social engagement, with a focus on new methods that have been used to promote mutual understanding, foster rational debate, build confidence, promote economic development, and secure social license. The projects examined span the life-cycle of radioactive waste management development, from ones that are purely aspirational, to ones in early planning stages, to ones seeking a host community, to one that is focused on sustainment of existing social license. Several notable lessons and innovative practices are identified that will be of interest to radioactive waste management project owners and developers, non-governmental organizations, potential host communities, and political decision-makers for such projects. (authors)

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

Radioactive Waste Minimization - Clearance for Reuse, Recycling and Disposal - 19598

Virtually all activities related to radioactive materials and to the utilization of nuclear energy generate radioactive waste. Due to limited storage facilities and waning public acceptance for long-term disposal of radioactive waste as well as the associated cost, implementation of strategies and methods for its effective reduction are mandatory. Clearance of materials is one possible concept to reach this goal. The International Atomic Energy Agency (IAEA) defines the process of clearance as 'the removal of radioactive materials or radioactive objects within authorized practices from any further regulatory control by the regulatory body.' Thus, these materials are no further treated as radioactive waste and may be recycled or disposed of conventionally. In quantitative terms, clearance is generally related to the '10 μSv-Concept' This means that the effective dose expected to be incurred by any member of the public due to cleared materials is of the order of 10 μSv or less in a year. A combination of appropriate clearance concepts, methods and measurement techniques, following an integrated clearance strategy for a project, allows the return of up to 97 % of the mass of a decommissioned NPP to the conventional waste cycle. This includes the concept of clearance for a specific purpose, often referred to as 'conditional clearance', i.e. the case where a certain condition has to be fulfilled before the act of clearance has been completed. As a technical expert organization, TUV NORD EnSys accompanied the development, implementation and execution of clearance procedures in more than 13 nuclear power-plants and installations, either operational or in various states of decommissioning. Thus, TUV Nord EnSys has abundant experience in and a comprehensive overview on the implementation of different clearance strategies and procedures. Considering the increased waste-volumes and additional waste-streams to be handled during decommissioning, the transition to a decommission-focused clearance system is of vital importance for successful waste-minimization. This paper provides a view on the perspective of a technical support organisation (TSO) towards implementation of a holistic clearance strategy into the decommissioning concept for radioactive waste minimization. (authors)

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

Applying lean management to aging infrastructure - 15416

Currently the Savannah River Site liquid waste system consists of 51 waste tanks, two operating evaporators, a salt processing pilot plant, the DWPF vitrification facility, the salt-stone Processing Facility and an Effluent Treatment Plant. Six of the waste tanks have been grouted and closed and the full scale Salt Waste Processing Facility (SWPF) is under construction. Some of the waste tanks and associated infrastructure were placed into operation as early as 1954 and DWPF went into operation in 1996. Due to this long history of operation, the SRS liquid waste system is being challenged by aging infrastructure issues and the stored radioactive liquid waste has been identified by regulators as the highest environmental risk in the state. Much of the DOE complex faces similar infrastructure issues. Remedies that can mitigate the impact of those issues are needed all across the complex. At SRS, Lean Management has been adopted as a business process to help address some of the impacts. Lean Management got its start decades ago in the manufacturing sector and has its roots in the work of W. Edwards Deming and his contemporaries. In more recent years the process has matured in the Toyota Production System and migrated into fields as diverse as services industries, health care, and now the DOE complex. Lean Management has a 'secret sauce' that must be correct to avoid becoming the 'flavor of the month'. The easy part is the talking. The hard part is the doing. The payoff is in the doing. Obviously as facilities and equipment age, more and more funds are needed to maintain, refurbish, and replace them. Funding of course is tight for all our sites. Lean management generates work efficiency so a given set of resources can produce more. It must be noted that this isn't by working harder but rather by focusing scarce resources on the few activities that truly add value. A positive aspect for companies that have been on this journey is employees feel less stress and higher motivation. The work they do matters and they have fewer outside impediments to distract and sap their energy. SRR has used the Lean process to streamline both work planning and control and the supply chain to improve maintenance efficiency and reduce the Corrective Maintenance backlog. This paper explains how this was done, the challenges encountered, and how organizations can use this methodology for their own improvements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Zavitsas’ hydration model for electrolytes only stable in the presence of another: NaAl(OH) 4 in aqueous NaOH solution

Many aqueous electrolytes are only stable in the presence of another electrolyte, such as electrolytes that are only soluble in strong acid or base. An example is sodium aluminate [NaAl(OH) 4 ], which is only stable in aqueous NaOH. This complicates developing thermodynamic parameters because it is difficult to separate the contributions of individual electrolytes in multicomponent solutions to measured bulk thermodynamic properties. The present study develops a method to determine the liquid phase parameters from solubility data for Zavitsas’ Hydration model, a model that incorporates hydration parameters into the activities of dissolved species. Furthermore, this study uses gibbsite [Al(OH) 3 ] solubility data in aqueous NaOH solution to develop Zavitsas' model parameters for NaAl(OH) 4 simultaneously with the equilibrium constants. A good fit of the solubility data was found, showing that Zavitsas’ model is effective for this system.

Bayer process↗
DOE OSTI2022

Effects of time-dependent deformation of shale on the integrity of a geological nuclear waste repository

Safety assessment of geological nuclear waste repositories is essential for the permanent disposal of spent nuclear fuels and high-level radioactive waste. The long-term integrity of the host rock as well as the engineered barrier system (e.g., bentonite buffer) surrounding nuclear waste canisters is of particular importance as decay heat from nuclear waste canisters, which will last over thousands of years, significantly disturbs the thermo-hydromechanical (THM) state of the repository. In this study, THM coupled simulations were carried out to investigate the effect of time-dependent deformation (i.e., creep) of shale on the long-term integrity of a generic subsurface nuclear waste repository. The Norton-Bailey creep model, which is also known as the Lemaitre-Menzel-Schreiner model, was employed to simulate the power-law type creep that is observed in shales. The TOUGH-FLAC simulator was employed for the THM coupled modeling of the repository. The objective of this study is to assess the effect of creep in different shales (i.e., high creep shale vs. low creep shale) on long-term stress and permeability changes in the repository. Results show potential advantages of constructing repositories in the high creep shale, as deviatoric stress levels in the formation decreased to zero in 100 years since the emplacement of nuclear waste canisters and the permeability also decreased to the undamaged, intrinsic levels in 10,000 years. Also, mean effective stress levels in the bentonite buffer increased by 100% in the high creep shale case relative to the low creep shale case at 10,000 years due to creep-induced contraction of the nuclear waste disposal tunnel. However, in earlier periods (e.g., 1000 years), the stress levels in the bentonite buffer were twice smaller in the high creep shale case than in the low creep shale case, which shows a tradeoff between the intermediate- (~1000 years) and long-term (>10,000 years) compaction levels in the bentonite buffer depending on the creep characteristics of the host shale.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Modeling nuclear waste disposal in crystalline rocks at the Forsmark and Olkiluoto repository sites – Evaluation of potential thermal–mechanical damage to repository excavations

We conduct coupled thermo-hydro-mechanical modeling of a KBS-3V repository design in crystalline rocks, using data and conditions from the Forsmark in Olkiluoto repository sites in Sweden and Finland. The study focuses on repository performance related to the impact of thermal and hydraulic evolution on the potential for thermal–mechanical damage to underground repository excavations. For the designs and conditions considered at the Forsmark and Olkiluoto repository sites, the simulations show a peak temperature well under the adopted performance target of a 100°C maximum temperature, whereas there is still a high potential for thermal–mechanical damage to the KBS-3V waste deposition holes. The thermal–mechanical damage is much more likely if rock permeability is so low that it delays saturation and swelling of bentonite-clay-based backfill beyond the time for the thermal–mechanical peak, which occurs 50 to 100 years after nuclear waste deposition. We also found that sidewalls of the KBS-3V emplacement tunnels are vulnerable to tensile fracturing due to the combined effect of thermal stressing and backfill swelling. The study highlights a strong interaction between bentonite-based backfill and host rock through capillary suction along with induced rock desaturation. A careful design and selection of the bentonite-clay-based backfill materials for KBS-3V tunnels and deposition holes can facilitate a timely saturation and backfill swelling that in turn can minimize thermal–mechanical damage.

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

Efficient Capture of Perrhenate and Pertechnetate by a Mesoporous Zr Metal–Organic Framework and Examination of Anion Binding Motifs

At the Hanford Site in southeastern Washington state, the U.S. Department of Energy intends to treat 56 million gallons of legacy nuclear waste by encasing it in borosilicate glass via vitrification. This process ineffectively captures radioactive pertechnetate (TcO 4 –) because of the ion’s volatility, thereby requiring a different remediation method for this long-lived (t 1/2 = 2.1 × 10 5 years), environmentally mobile species. Currently available sorbents lack the desired combination of high uptake capacity, fast kinetics, and selectivity. Here, we evaluate the ability of the chemically and thermally robust Zr 6 -based metal–organic framework (MOF), NU-1000, to capture perrhenate (ReO 4 –), a pertechnetate simulant, and pertechnetate. Our material exhibits an excellent perrhenate uptake capacity of 210 mg/g, reaches saturation within 5 min, and maintains perrhenate uptake in the presence of competing anions. Additionally, experiments with pertechnetate confirm perrhenate is a suitable surrogate. Single-crystal X-ray diffraction indicates both chelating and nonchelating perrhenate binding motifs are present in both the small pore and the mesopore of NU-1000. Postadsorption diffuse reflectance infrared Fourier transform spectroscopy (DRIFTS) further elucidates the uptake mechanism and powder X-ray diffraction (PXRD) and Brunauer–Emmett–Teller (BET) surface area analysis confirm the retention of crystallinity and porosity of NU-1000 throughout adsorption.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Laboratory Optimization Tests of Technetium Decontamination of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Melter Off-Gas Condensate Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The baseline plan for disposition of this stream is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrification mission duration and quantity of glass waste.

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

Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation, and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator, in the Effluent Management Facility (EMF), and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator, so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would reduce the need for closely integrated operation of the LAW melter and the Pretreatment Facilities. Long-term implementation of this option after WTP start-up would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other operational complexities such a recycle stream presents. In order to accurately plan for the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to accurately account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, and determine the distribution of key regulatory-impacting constituents. The LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures, have limited solubility in the glass waste form, and represent a materials corrosion concern, such as halides and sulfate. Because this stream will recycle within WTP, these components will accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfates in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to demonstrate evaporation of a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, in order to predict the composition of the effluents from the EMF evaporator to aid in planning for their disposition. This document describes the results of that test using the core simulant. This simulant formulation is designated as the “core simulant”; other additives will be included for specific testing, such as volatiles for evaporation or hazardous metals for measuring leaching properties of waste forms. The results indicate that the simulant can easily be concentrated via evaporation. During that the pH adjustment step in simulant preparation, ammonium is quickly converted to ammonia, and most of the ammonia was stripped from the simulated waste and partitioned to the condensate. Additionally, it was found that after concentrating (>12x) and cooling that a small amount of LiF and Na 3 (SO 4 )F precipitate out of solution. With the exception of ammonia, analysis of the condensate indicated very low to below detectable levels of many of the constituents in the simulant, yielding very high decontamination factors (DF).

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

Crystalline Silicotitanate ion exchange media long-term storage evaluation

Crystalline Silicotitanate a (CST) ion exchange media is produced in an engineered form, known as IONSIV™ IE-911. This material has been tested extensively for use in removing 137 Cs from Savannah River Site (SRS) and Hanford site tank waste. This inorganic material is highly selective for removing cesium ions, even in the presence of high concentrations of sodium and potassium ions and is tolerant of the harsh conditions of the varying tank chemistries. A batch of caustic washed IE-911 that had been produced by the supplier and received at SRS in 2001 was tested to verify that it was still in good condition after the extended storage. This testing involved standard batch contacts to measure the removal of Cs from simulated tank waste. This report describes the results of testing performed to verify that the IE-911 matches the performance of previous batches of this media. To confirm that the performance of IE-911 is similar to previous batches and that no degradation had occurred during storage, batch contact testing was performed using SRNL’s standard methodology using a non-radioactive SRS Average Waste Simulant composition utilized in previous studies. Batch contact kinetics testing was also conducted to determine cesium loading on this batch versus time. The cesium equilibrium loading isotherm in SRS simulant at 23°C was compared to ZAM isotherm model predictions and previous test results with IE-911 media. The measured and predicted distribution coefficients and cesium loading (“Q”) are shown in Table ES-1. The physical properties of the material were also determined, including density, moisture content, and particle size.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Realistic Geochemical Parameter Uncertainty for Performance Assessment Modeling - 17066

Performance assessments (PAs) include uncertain geochemical parameters that affect the fate and transport of contaminants in environmental media, such as partitioning coefficients for water/rock (K{sub d}) and water/air (K{sub H}), diffusion coefficients, and aqueous solubilities. A statistical approach to realistic, as opposed to 'conservative,' stochastic distribution development for geochemical parameters is presented here. For contaminants not identified as focus elements during screening, broad generic distributions are developed. If any parameter is flagged in the subsequent model sensitivity analysis (SA), it becomes a focus element for further research during the next round of distribution development. This iterative approach is efficient and technically defensible at each stage. Data are collected in databases and may be appropriately weighted or filtered. In the relatively few cases where data exist from local or analogous rocks, those data may be prioritized. Once established, the databases can be used for future PAs or other environmental models and can be expanded as needed based on model SA results, additional screening, and as additional relevant data become available. The iterative method is demonstrated for K{sub d} and K{sub H} in a PA model for a radioactive waste disposal facility. An SA was performed based on preliminary geochemical parameter distributions that showed certain model endpoints are sensitive to K{sub d} and K{sub H} values for some contaminants. This underscores the importance of including realistic uncertainty in these geochemical parameters. (authors)

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

Waste Information Management System with 2018-19 Waste Streams - 19106

The Waste Information Management System (WIMS) 2018-19 was updated to support the Department of Energy (DOE) accelerated cleanup program. The schedule compression required close coordination and a comprehensive review and prioritization of the barriers that impeded treatment and disposition of the waste streams at each site. Many issues related to waste treatment and disposal were potential critical path issues under the accelerated schedule. In order to facilitate accelerated cleanup initiatives, waste managers at DOE field sites and at DOE Headquarters in Washington, D.C., needed timely waste forecast and transportation information regarding the volumes and types of radioactive waste that would be generated by DOE sites over the next 35 years. Each local DOE site historically collected, organized, and displayed waste forecast information in separate and unique systems. In order for interested parties to understand and view the complete DOE complex-wide picture, the radioactive waste and shipment information of each DOE site needed to be entered into a common application. The WIMS application was therefore created to serve as a common application to improve stakeholder comprehension and improve DOE radioactive waste treatment and disposal planning and scheduling. WIMS allows identification of total forecasted waste volumes, material classes, disposition sites, choke points, technological or regulatory barriers to treatment and disposal, along with forecasted waste transportation information by rail, truck and intermodal shipments. The Applied Research Center (ARC) at Florida International University (FIU) in Miami, Florida, developed and deployed the web-based forecast and transportation system and is responsible for updating the radioactive waste forecast and transportation data on a regular basis to ensure the long-term viability and value of this system. (authors)

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

OE-WIPP Event Presentation

Information is given on waste generation at TA-55 and remediation needed to meet WIPP acceptance criteria, including the role of nitrate salts. Breaching of a particular waste-filled drum is reviewed, along with an accident analysis and steps for corrective actions and improved process management.

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

Results for the First, Second, and Third Quarter Calendar Year 2015 Tank 50H WAC slurry samples chemical and radionuclide contaminants

This report details the chemical and radionuclide contaminant results for the characterization of the Calendar Year (CY) 2015 First, Second, and Third Quarter sampling of Tank 50H for the Saltstone Waste Acceptance Criteria (WAC) in effect at that time. Information from this characterization will be used by Defense Waste Processing Facility (DWPF) & Saltstone Facility Engineering (D&S-FE) to support the transfer of low-level aqueous waste from Tank 50H to the Salt Feed Tank in the Saltstone Facility in Z-Area, where the waste will be immobilized. This information is also used to update the Tank 50H Waste Characterization System. Previous memoranda documenting the WAC analyses results have been issued for these three samples.

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

Integration of performance assessments and 3116 basis documents within the Savannah River Site liquid waste facilities - 15375

The end of the Cold War has left a legacy of over 136 million liters (36 million gallons) of radioactive waste in the aging waste tanks at the Department of Energy's Savannah River Site (SRS). A robust program is in place to remove waste from these tanks, treat the waste to separate it into a relatively small volume of high-level waste and a large volume of low-level waste, and to actively dispose of the low-level waste on-site. This process allows the US Department of Energy (DOE) to move forward with closure of the cleaned waste tanks and associated ancillary structures. To support performance-based, risk decision making for closure and disposal activities, three individual performance assessments (PAs) have been developed - salt-stone Disposal Facility, F-Tank Farm and H-Tank Farm. In addition, three Waste Determinations (WDs) have been made by the Secretary of Energy, in consultation with the Nuclear Regulatory Commission (NRC), under the authority of Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA Section 3116). To support these waste determination decisions, individual '3116 Basis Documents' were developed that describe the assumptions and conditions under which the criteria outlined in NDAA Section 3116 are met. The development of each of the PAs represented a long and arduous effort that involved the development of both deterministic and probabilistic conceptual models of the facilities, the development of assumptions and associated ranges, uncertainty and sensitivity analyses, model data collection and analyses, writing of the actual performance assessment, internal and external peer reviews, and ultimately approval by DOE. Similarly, the development of the 3116 Basis Documents represents similar time and financial commitments. At the end of these efforts, after years of preparation and millions of dollars in investment, we have thousands of pages of highly technical text, tables of data and hundreds of graphs and charts all contained in a series of thick binders that fit neatly on the shelf of a bookcase. We were done - right? Actually, we were just getting started. Performance Assessments and 3116 Basis Documents are living documents that shape day-to-day and long-term decision making in the operation of a low-level waste disposal facility (e.g., salt-stone Disposal Facility) or in area closure activities (e.g., F-Tank Farm or H-Tank Farm). Although these documents do not contain specifically described requirements per se, they do utilize specific assumptions that, if changed significantly, could result in no longer being able to demonstrate that there is reasonable assurance that the required performance objectives or the criteria from NDAA Section 3116 can be met. Savannah River Remediation LLC (SRR), the contractor responsible for Liquid Waste operations at SRS, has not only developed the PAs and 3116 Basis Documents to support this work and shepherded these documents through the DOE approval process, but has also developed the processes and procedures to ensure the successful implementation and integration of these documents into day-to-day decision making within the associated facilities. This paper describes these processes, including the Un-reviewed Waste Management Question process and the PA Maintenance process, and provides specific examples of their successful implementation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Release Fractions from Airborne Fluid and Slurry Spills

Release fractions are commonly used to make conservative estimates of emissions from processes and are applied to characterize industrial operations to safeguard workers and public. Release fractions cannot be derived theoretically; they must be measured. Appropriate measurements are limited. This paper describes tests that were conducted to provide large-scale release fraction data for estimating releases from three operations proposed for waste remediation at the Hanford Site in Richland, Washington. Experiments were conducted to measure release fractions of three processes: 1) pipe discharge of liquid and slurry in free fall into a liquid or slurry, 2) jet transfer of contents to the other side of the tank either impacting the tank wall or falling into the fluid pool, and 3) high pressure fan jet striking a steel plate or simulated waste from a small stand-off distance. The tests were conducted without forced ventilation in an enclosed prototypic ¼-scale model of a 3785 m^3 (1 million gallon) tank in use at the Hanford Site for storage of radioactive waste. Aerosol releases were characterized based on filter samples of aerosol collected near the process stream, tank wall, and tank top to determine release fraction at engineering scale. Particle size distribution measurements were analyzed to determine whether the particles were in the respirable range. The release fractions for the top of the unventilated tank ranged from 9x10-7 to 6x10-4 depending on the process simulated. The particle size distribution was determined to be log-normally distributed with geometric mean diameters ranging from 3 to 8 µm. Thus, all particles are conservatively considered to be “respirable.”

aerosol generation, release fraction, water spill,↗
DOE OSTI2015

Process and equipment development for hot isostatic pressing treatability study

Battelle Energy Alliance (BEA), LLC, has developed processes and equipment for a pilot-scale hot isostatic pressing (HIP) treatability study to stabilize and volume reduce radioactive calcine stored at Idaho National Laboratory (INL). In 2009, the U. S. Department of Energy signed a Record of Decision with the state of Idaho selecting HIP technology as the method to treat 5,800 yd^3 (4,400 m^3) of granular zirconia and alumina calcine produced between 1953 and 1992 as a waste byproduct of spent nuclear fuel reprocessing. Since the 1990s, a variety of radioactive and hazardous waste forms have been remotely treated using HIP within INL hot cells. To execute the remote process at INL, waste is loaded into a stainless-steel or aluminum can, which is evacuated, sealed, and placed into a HIP furnace. The HIP simultaneously heats and pressurizes the waste, reducing its volume and increasing its durability. Two 1 gal cans of calcine waste currently stored in a shielded cask were identified as candidate materials for a treatability study involving the HIP process. Equipment and materials for cask-handling and calcine transfer into INL hot cells, as well as remotely operated equipment for waste can opening, particle sizing, material blending, and HIP can loading have been designed and successfully tested. These results demonstrate BEA’s readiness for treatment of INL calcine.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗