Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank Farm Blend) By Fluidized Bed Steam Reformation (FBSR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jantzen, C. M.; Crawford, C. L.; Bannochie, C. J.

The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order,more » also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Supplemental Treatment is likely to be required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP’s LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750°C) continuous method by which LAW can be processed irrespective of whether the waste contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be comparable to LAW glass, i.e. leaches Tc-99, Re and Na at <2g/m 2 during ASTM C1285 (Product Consistency) durability testing. Monolithing of the granular FBSR product was investigated to prevent dispersion during transport or burial/storage. Monolithing in an inorganic geopolymer binder, which is

Effluent Management Facility Evaporator Bottom-Waste Streams Formulation and Waste Form Qualification Testing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Saslow, Sarah A.; Um, Wooyong; Russell, Renee L.

This report describes the results from grout formulation and cementitious waste form qualification testing performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates three grout formulations proposed for wide-range treatment of different waste stream compositions expected for the Hanford Effluent Management Facility (EMF) evaporator bottom waste. This work supports the technical development need for alternative disposition paths for the EMF evaporator bottom wastes and future direct feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious wastemore » form qualification testing. The work contained within this report relates to waste form development and testing, and does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY 2017 and future waste form development efforts. The provided results and data should be used by (1) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (2) decision makers interested in off-site waste form disposal, and (3) the U.S. Department of Energy, their Hanford Site contractors and stakeholders as they assess the IDF PA program at the Hanford Site. The results reported help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottom waste, and improve the technical defensibility of long-term waste form risk estimates.« less

Data Packages for the Hanford Immobilized Low Activity Tank Waste Performance Assessment 2001 Version [SEC 1 THRU 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

MANN, F.M.

Data package supporting the 2001 Immobilized Low-Activity Waste Performance Analysis. Geology, hydrology, geochemistry, facility, waste form, and dosimetry data based on recent investigation are provided. Verification and benchmarking packages for selected software codes are provided.

Integration of National Laboratory and Low-Activity Waste Pre-Treatment System Technology Service Providers - 16435

DOE Office of Scientific and Technical Information (OSTI.GOV)

Subramanian, Karthik H.; Thien, Michael G.; Wellman, Dawn M.

The National Laboratories are a critical partner and provide expertise in numerous aspects of the successful execution of the Direct-Feed Low Activity Waste Program. The National Laboratories are maturing the technologies of the Low-Activity Waste Pre-Treatment System (LAWPS) consistent with DOE Order 413.3B “Program and Project Management for the Acquisition of Capital Assets” expectations. The National Laboratories continue to mature waste forms, i.e. glass and secondary waste grout, for formulations and predictions of long-term performance as inputs to performance assessments. The working processes with the National Laboratories have been developed in procurements, communications, and reporting to support the necessary delivery-basedmore » technology support. The relationship continues to evolve from planning and technology development to support of ongoing operations and integration of multiple highly coordinated facilities.« less

Secondary Waste Cementitious Waste Form Data Package for the Integrated Disposal Facility Performance Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantrell, Kirk J.; Westsik, Joseph H.; Serne, R Jeffrey

A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at themore » Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.« less

Liquid secondary waste. Waste form formulation and qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cozzi, A. D.; Dixon, K. L.; Hill, K. A.

The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testingmore » to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.« less

Analysis of alternatives for immobilized low activity waste disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burbank, D.A.

This report presents a study of alternative disposal system architectures and implementation strategies to provide onsite near-surface disposal capacity to receive the immobilized low-activity waste produced by the private vendors. The analysis shows that a flexible unit strategy that provides a suite of design solutions tailored to the characteristics of the immobilized low-activity waste will provide a disposal system that best meets the program goals of reducing the environmental, health, and safety impacts; meeting the schedule milestones; and minimizing the life-cycle cost of the program.

Secondary Waste Form Down Selection Data Package – Ceramicrete

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantrell, Kirk J.; Westsik, Joseph H.

2011-08-31

As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratorymore » is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The

Melter feed viscosity during conversion to glass: Comparison between low-activity waste and high-level waste feeds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Tongan; Chun, Jaehun; Dixon, Derek R.

During nuclear waste vitrification, a melter feed (generally a slurry-like mixture of a nuclear waste and various glass forming and modifying additives) is charged into the melter where undissolved refractory constituents are suspended together with evolved gas bubbles from complex reactions. Knowledge of flow properties of various reacting melter feeds is necessary to understand their unique feed-to-glass conversion processes occurring within a floating layer of melter feed called a cold cap. The viscosity of two low-activity waste (LAW) melter feeds were studied during heating and correlated with volume fractions of undissolved solid phase and gas phase. In contrast to themore » high-level waste (HLW) melter feed, the effects of undissolved solid and gas phases play comparable roles and are required to represent the viscosity of LAW melter feeds. This study can help bring physical insights to feed viscosity of reacting melter feeds with different compositions and foaming behavior in nuclear waste vitrification.« less

Data Package for Secondary Waste Form Down-Selection—Cast Stone

DOE Office of Scientific and Technical Information (OSTI.GOV)

Serne, R. Jeffrey; Westsik, Joseph H.

2011-09-05

Available literature on Cast Stone and Saltstone was reviewed with an emphasis on determining how Cast Stone and related grout waste forms performed in relationship to various criteria that will be used to decide whether a specific type of waste form meets acceptance criteria for disposal in the Integrated Disposal Facility (IDF) at Hanford. After the critical review of the Cast Stone/Saltstone literature, we conclude that Cast Stone is a good candidate waste form for further consideration. Cast stone meets the target IDF acceptance criteria for compressive strength, no free liquids, TCLP leachate are below the UTS permissible concentrations andmore » leach rates for Na and Tc-99 are suiteably low. The cost of starting ingredients and equipment necessary to generate Cast Stone waste forms with secondary waste streams are low and the Cast Stone dry blend formulation can be tailored to accommodate variations in liquid waste stream compositions. The database for Cast Stone short-term performance is quite extensive compared to the other three candidate waste solidification processes. The solidification of liquid wastes in Cast Stone is a mature process in comparison to the other three candidates. Successful production of Cast Stone or Saltstone has been demonstrated from lab-scale monoliths with volumes of cm3 through m3 sized blocks to 210-liter sized drums all the way to the large pours into vaults at Savannah River. To date over 9 million gallons of low activity liquid waste has been solidified and disposed in concrete vaults at Savannah River.« less

Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hulse, R.A.

1991-08-01

Planning for storage or disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of that waste to estimate volumes, radionuclide activities, and waste forms. Data from existing literature, disposal records, and original research were used to estimate the characteristics and project volumes and radionuclide activities to the year 2035. GTCC LLW is categorized as: nuclear utilities waste, sealed sources waste, DOE-held potential GTCC LLW; and, other generator waste. It has been determined that the largest volume of those wastes, approximately 57%, is generated by nuclear power plants. The Other Generator waste category contributes approximately 10% of the totalmore » GTCC LLW volume projected to the year 2035. Waste held by the Department of Energy, which is potential GTCC LLW, accounts for nearly 33% of all waste projected to the year 2035; however, no disposal determination has been made for that waste. Sealed sources are less than 0.2% of the total projected volume of GTCC LLW.« less

Low-Activity Waste Pretreatment System Additional Engineering-Scale Integrated Test Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Landon, Matt R.; Wilson, Robert A.

Washington River Protections Solutions, LLC’s (WRPS) Low Activity Waste Pretreatment System (LAWPS) Project provides for the early production of immobilized low-activity waste (ILAW) by feeding LAW directly from Tank Farms to the Waste Treatment and Immobilization Plant (WTP) LAW Facility, bypassing the WTP Pretreatment Facility. Prior to the transfer of feed to the WTP LAW Vitrification Facility, tank supernatant waste will be pretreated in the LAWPS to meet the WTP LAW waste acceptance criteria (WAC). Full-scale and engineering-scale testing of critical technology elements, as part of the technology maturation process, are components of the overall LAWPS Project. WRPS awarded themore » engineering-scale integrated testing scope to AECOM via WRPS Subcontract 58349. This report is deliverable MSR-008 of the subcontract.« less

Candidate Low-Temperature Glass Waste Forms for Technetium-99 Recovered from Hanford Effluent Management Facility Evaporator Concentrate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ding, Mei; Tang, Ming; Rim, Jung Ho

Alternative treatment and disposition options may exist for technetium-99 (99Tc) in secondary liquid waste from the Hanford Direct-Feed Low-Activity Waste (DFLAW) process. One approach includes development of an alternate glass waste form that is suitable for on-site disposition of technetium, including salts and other species recovered by ion exchange or precipitation from the EMF evaporator concentrate. By recovering the Tc content from the stream, and not recycling the treated concentrate, the DFLAW process can potentially be operated in a more efficient manner that lowers the cost to the Department of Energy. This report provides a survey of candidate glass formulationsmore » and glass-making processes that can potentially incorporate technetium at temperatures <700 °C to avoid volatilization. Three candidate technetium feed streams are considered: (1) dilute sodium pertechnetate loaded on a non-elutable ion exchange resin; (2) dilute sodium-bearing aqueous eluent from ion exchange recovery of pertechnetate, or (3) technetium(IV) oxide precipitate containing Sn and Cr solids in an aqueous slurry. From the technical literature, promising candidate glasses are identified based on their processing temperatures and chemical durability data. The suitability and technical risk of three low-temperature glass processing routes (vitrification, encapsulation by sintering into a glass composite material, and sol-gel chemical condensation) for the three waste streams was assessed, based on available low-temperature glass data. For a subset of candidate glasses, their long-term thermodynamic behavior with exposure to water and oxygen was modeled using Geochemist’s Workbench, with and without addition of reducing stannous ion. For further evaluation and development, encapsulation of precipitated TcO2/Sn/Cr in a glass composite material based on lead-free sealing glasses is recommended as a high priority. Vitrification of pertechnetate in aqueous anion exchange eluent

Conceptual design statement of work for the immobilized low-activity waste interim storage facility project

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carlson, T.A., Fluor Daniel Hanford

1997-02-06

The Immobilized Low-Activity Waste Interim Storage subproject will provide storage capacity for immobilized low-activity waste product sold to the U.S. Department of Energy by the privatization contractor. This statement of work describes the work scope (encompassing definition of new installations and retrofit modifications to four existing grout vaults), to be performed by the Architect-Engineer, in preparation of a conceptual design for the Immobilized Low-Activity Waste Interim Storage Facility.

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

DOE Office of Scientific and Technical Information (OSTI.GOV)

Um, Wooyong; Williams, B. D.; Snyder, Michelle M. V.

This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting themore » U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99Tc desorption K d (distribution coefficient) values for 99Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99Tc(IV)-bearing solid phases for 99Tc transport in reducing conditions to support the IDF PA.« less

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herbst, A.K.; Rogers, A.Z.; McCray, J.A.

The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

Improving radioactive waste management: an overview of the Environmental Protection Agency's low-activity waste effort.

PubMed

Schultheisz, Daniel J; Czyscinski, Kenneth S; Klinger, Adam D

2006-11-01

Radioactive waste disposal in the United States is marked by a fragmented regulatory system, with requirements that often focus on the origin or statutory definition of the waste, rather than the hazard of the material in question. It may be possible to enhance public protection by moving toward a system that provides disposal options appropriate for the hazard presented by the waste in question. This paper summarizes aspects of an approach focusing on the potential use, with appropriate conditions, of Resource Conservation and Recovery Act Subtitle-C hazardous waste landfills for disposal of "low-activity" wastes and public comments on the suggested approach.

Performance Test on Polymer Waste Form - 12137

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, Se Yup

, biodegradation testing and water immersion testing, no degradation was observed in the waste forms. Also, by measuring the compressive strength after these tests, it was confirmed that the structural integrity was still retained. A leach test was performed by using non radioactive cobalt, cesium and strontium. The leaching of cobalt, cesium and strontium from the polymer waste forms was very low. Also, the polymer waste forms were found to possess adequate fire resistance. From the results of the performance tests, it is believed that the polymer waste form is very stable and can satisfy the acceptance criteria for permanent disposal. At present, Performance tests with full scale polymer waste forms are on-going in order to obtain qualification certificate by the regulatory institute in Korea. (authors)« less

Commercial high-level-waste management: Options and economics. A comparative analysis of the ceramic and glass waste forms

NASA Astrophysics Data System (ADS)

McKisson, R. L.; Grantham, L. F.; Guon, J.; Recht, H. L.

1983-02-01

Results of an estimate of the waste management costs of the commercial high level waste from a 3000 metric ton per year reprocessing plant show that the judicious use of the ceramic waste form can save about $2 billion during a 20 year operating campaign relative to the use of the glass waste form. This assumes PWR fuel is processed and the waste is encapsulated in 0.305-m-diam canisters with ultimate emplacement in a BWIP-type horizontal-borehole repository. Waste loading and waste form density are the driving factors in that the low waste loading (25%) and relatively low density (3.1 g cu cm) characteristic of the glass form require several times as many canisters to handle a given waste throughput than is needed for the ceramic waste form whose waste loading capability exceeds 60% and whose waste density is nominally 5.2 cu cm.

Scenarios for the Hanford immobilized Low-Activity waste (ILAW) performance assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

MANN, F.M.

The purpose of the next version of the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (ILAW PA) is to provide an updated estimate of the long-term human health and environmental impact of the disposal of ILAW and to compare these estimates against performance objectives displayed in Tables 1,2, and 3 (Mann 1999a). Such a radiological performance assessment is required by U.S. Department of Energy (DOE) Orders on radioactive waste management (DOE 1988a and DOE 1999a). This document defines the scenarios that will be used for the next update of the PA that is scheduled to be issued in 2001.more » Since the previous performance assessment (Mann 1998) was issued, considerable additional data on waste form behavior and site-specific soil geotechnical properties have been collected. In addition, the 2001 ILAW PA will benefit from improved computer models and the experience gained from the previous performance assessment. However, the scenarios (that is, the features, events, and processes analyzed in the Performance assessment) for the next PA are very similar to the ones in the 1998 PA.« less

Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo

2011-08-12

indicated that all the waste forms had leachability indices better than the target LI > 9 for technetium; (2) Rhenium diffusivity: Cast Stone 2M specimens, when tested using EPA 1315 protocol, had leachability indices better than the target LI > 9 for technetium based on rhenium as a surrogate for technetium. All other waste forms tested by ANSI/ANS 16.1, ASTM C1308, and EPA 1315 test methods had leachability indices that were below the target LI > 9 for Tc based on rhenium release. These studies indicated that use of Re(VII) as a surrogate for 99Tc(VII) in low temperature secondary waste forms containing reductants will provide overestimated diffusivity values for 99Tc. Therefore, it is not appropriate to use Re as a surrogate 99Tc in future low temperature waste form studies. (3) Iodine diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that the three waste forms had leachability indices that were below the target LI > 11 for iodine. Therefore, it may be necessary to use a more effective sequestering material than silver zeolite used in two of the waste forms (Ceramicrete and DuraLith); (4) Sodium diffusivity: All the waste form specimens tested by the three leach methods (ANSI/ANS 16.1, ASTM C1308, and EPA 1315) exceeded the target LI value of 6; (5) All three leach methods (ANS 16.1, ASTM C1308 and EPA 1315) provided similar 99Tc diffusivity values for both short-time transient diffusivity effects as well as long-term ({approx}90 days) steady diffusivity from each of the three tested waste forms (Cast Stone 2M, Ceramicrete and DuraLith). Therefore, any one of the three methods can be used to determine the contaminant diffusivities from a selected waste form.« less

Reductive capacity measurement of waste forms for secondary radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Um, Wooyong; Yang, Jung-Seok; Serne, R. Jeffrey

2015-12-01

The reductive capacities of dry ingredients and final solid waste forms were measured using both the Cr(VI) and Ce(IV) methods and the results were compared. Blast furnace slag (BFS), sodium sulfide, SnF2, and SnCl2 used as dry ingredients to make various waste forms showed significantly higher reductive capacities compared to other ingredients regardless of which method was used. Although the BFS exhibits appreciable reductive capacity, it requires greater amounts of time to fully react. In almost all cases, the Ce(IV) method yielded larger reductive capacity values than those from the Cr(VI) method and can be used as an upper boundmore » for the reductive capacity of the dry ingredients and waste forms, because the Ce(IV) method subjects the solids to a strong acid (low pH) condition that dissolves much more of the solids. Because the Cr(VI) method relies on a neutral pH condition, the Cr(VI) method can be used to estimate primarily the waste form surface-related and readily dissolvable reductive capacity. However, the Cr(VI) method does not measure the total reductive capacity of the waste form, the long-term reductive capacity afforded by very slowly dissolving solids, or the reductive capacity present in the interior pores and internal locations of the solids.« less

Assessment of the Cast Stone Low-Temperature Waste Form Technology Coupled with Technetium Removal - 14379

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Christopher F.; Rapko, Brian M.; Serne, R. Jeffrey

2014-03-03

The U.S. Department of Energy Office of Environmental Management (EM) is engaging the national laboratories to provide the scientific and technological rigor to support EM program and project planning, technology development and deployment, project execution, and assessment of program outcomes. As an early demonstration of this new responsibility, Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL) were chartered to implement a science and technology program addressing low-temperature waste forms for immobilization of DOE aqueous waste streams, including technetium removal as an implementing technology. As a first step, the laboratories examined the technical risks and uncertainties associated withmore » the Cast Stone waste immobilization and technetium removal projects at Hanford. Science and technology gaps were identified for work associated with 1) conducting performance assessments and risk assessments of waste form and disposal system performance, and 2) technetium chemistry in tank wastes and separation of technetium from waste processing streams. Technical approaches to address the science and technology gaps were identified and an initial sequencing priority was suggested. A subset of research was initiated in 2013 to begin addressing the most significant science and technology gaps. The purpose of this paper is to report progress made towards closing these gaps and provide notable highlights of results achieved to date.« less

LOW ACTIVITY WASTE FEED SOLIDS CARACTERIZATION AND FILTERABILITY TESTS

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCabe, D.; Crawford, C.; Duignan, M.

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant (WTP) that is currently under construction. The baseline plan for the WTP Pretreatment facility is to treat the waste, splitting it into High Level Waste (HLW) feed and Low Activity Waste (LAW) feed. Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed onsite in the Integrated Disposal Facility (IDF). There are currently no plans to treat the waste to remove technetium in the WTP Pretreatment facility, so itsmore » disposition path is the LAW glass. Options are being explored to immobilize the LAW portion of the tank waste, i.e., the LAW feed from the WTP Pretreatment facility. Removal of {sup 99}Tc from the LAW Feed, followed by off-site disposal of the {sup 99}Tc, would eliminate a key risk contributor for the IDF Performance Assessment (PA) for supplemental waste forms, and has potential to reduce treatment and disposal costs. Washington River Protection Solutions (WRPS) is developing some conceptual flow sheets for LAW treatment and disposal that could benefit from technetium removal. One of these flowsheets will specifically examine removing {sup 99}Tc from the LAW feed stream to supplemental immobilization. The conceptual flow sheet of the {sup 99}Tc removal process includes a filter to remove insoluble solids prior to processing the stream in an ion exchange column, but the characteristics and behavior of the liquid and solid phases has not previously been investigated. This report contains results of testing of a simulant that represents the projected composition of the feed to the Supplemental LAW process. This feed composition is not identical to the aqueous tank waste fed to the Waste Treatment Plant because it has been processed through WTP Pretreatment facility and therefore contains internal changes and recycle streams that will be generated within the WTP process

Getters for improved technetium containment in cementitious waste forms

DOE PAGES

Asmussen, R. Matthew; Pearce, Carolyn I.; Miller, Brian W.; ...

2017-07-26

A cementitious waste form, Cast Stone, is a possible candidate technology for the immobilization of low activity nuclear waste (LAW) at the Hanford site. This paper focuses on the addition of getter materials to Cast Stone that can sequester Tc from the LAW, and in turn, lower Tc release from the Cast Stone. Two getters which produce different products upon sequestering Tc from LAW were tested: Sn(II) apatite (Sn-A) that removes Tc as a Tc(IV)-oxide and potassium metal sulfide (KMS-2) that removes Tc as a Tc(IV)-sulfide species, allowing for a comparison of stability of the form of Tc upon enteringmore » the waste form. The Cast Stone with KMS-2 getter had the best performance with addition equivalent to ~0.08 wt% of the total waste form mass. The observed diffusion (D obs) of Tc decreased from 4.6 ± 0.2 × 10 -12 cm 2/s for Cast Stone that did not contain a getter to 5.4 ± 0.4 × 10 -13 cm 2/s for KMS-2 containing Cast Stone. Finally, it was found that Tc-sulfide species are more stable against re-oxidation within getter containing Cast Stone compared with Tc-oxide and is the origin of the decrease in Tc D obs when using the KMS-2.« less

Getters for improved technetium containment in cementitious waste forms.

PubMed

Asmussen, R Matthew; Pearce, Carolyn I; Miller, Brian W; Lawter, Amanda R; Neeway, James J; Lukens, Wayne W; Bowden, Mark E; Miller, Micah A; Buck, Edgar C; Serne, R Jeffery; Qafoku, Nikolla P

2018-01-05

A cementitious waste form, Cast Stone, is a possible candidate technology for the immobilization of low activity nuclear waste (LAW) at the Hanford site. This work focuses on the addition of getter materials to Cast Stone that can sequester Tc from the LAW, and in turn, lower Tc release from the Cast Stone. Two getters which produce different products upon sequestering Tc from LAW were tested: Sn(II) apatite (Sn-A) that removes Tc as a Tc(IV)-oxide and potassium metal sulfide (KMS-2) that removes Tc as a Tc(IV)-sulfide species, allowing for a comparison of stability of the form of Tc upon entering the waste form. The Cast Stone with KMS-2 getter had the best performance with addition equivalent to ∼0.08wt% of the total waste form mass. The observed diffusion (D obs ) of Tc decreased from 4.6±0.2×10 -12 cm 2 /s for Cast Stone that did not contain a getter to 5.4±0.4×10 -13 cm 2 /s for KMS-2 containing Cast Stone. It was found that Tc-sulfide species are more stable against re-oxidation within getter containing Cast Stone compared with Tc-oxide and is the origin of the decrease in Tc D obs when using the KMS-2. Copyright © 2017 Elsevier B.V. All rights reserved.

Greater-than-Class C low-level waste characterization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piscitella, R.R.

1991-12-31

In 1985, Public Law 99-240 (Low-Level Radioactive Waste Policy Amendments Act of 1985) made the Department of Energy (DOE) responsible for the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW). DOE strategies for storage and disposal of GTCC LLW required characterization of volumes, radionuclide activities, and waste forms. Data from existing literature, disposal records, and original research were used to estimate characteristics, project volumes, and determine radionuclide activities to the years 2035 and 2055. Twenty-year life extensions for 70% of the operating nuclear reactors were assumed to calculate the GTCC LLW available in 2055. The following categories of GTCCmore » LLW were addressed: Nuclear Utilities Waste; Potential Sealed Sources GTCC LLW; DOE-Held Potential GTCC LLW; and Other Generator Waste. It was determined that the largest volume of these wastes, approximately 57%, is generated by nuclear utilities. The Other Generator Waste category contributes approximately 10% of the total GTCC LLW volume projected to the year 2035. DOE-Held Potential GTCC LLW accounts for nearly 33% of all waste projected to the year 2035. Potential Sealed Sources GTCC LLW is less than 0.2% of the total projected volume. The base case total projected volume of GTCC LLW for all categories was 3,250 cubic meters. This was substantially less than previous estimates.« less

Redox-dependent solubility of technetium in low activity waste glass

NASA Astrophysics Data System (ADS)

Soderquist, Chuck Z.; Schweiger, Michael J.; Kim, Dong-Sang; Lukens, Wayne W.; McCloy, John S.

2014-06-01

The solubility of technetium was measured in a Hanford low activity waste (LAW) glass simulant, to investigate the extent that technetium solubility controls the incorporation of technetium into LAW glass. A series of LAW glass samples, spiked with 500-6000 ppm of Tc as potassium pertechnetate, were melted at 1000 °C in sealed fused quartz ampoules. Technetium solubility was determined in the quenched bulk glass to be 2000-2800 ppm, with slightly reducing conditions due to choice of milling media resulting in reductant contamination and higher solubility. The chemical form of technetium obtained by X-ray absorption near edge spectroscopy is mainly isolated, octahedrally-coordinated Tc(IV), with a minority of Tc(VII) in some glasses and TcO2 in two glasses. The concentration and speciation of technetium depends on glass redox and amount of technetium added. Salts formed at the top of higher technetium loaded glasses during the melt. The results of this study show that technetium solubility should not be a factor in technetium retention during melting of Hanford LAW glass.

U.S. Food Loss and Waste 2030 Champions Activity Form

EPA Pesticide Factsheets

To join the U.S. Food Loss and Waste 2030 Champions, organizations complete and submit the 2030 Champions form, in which they commit to reduce food loss and waste in their own operations and periodically report their progress on their website.

Iso standardization of theoretical activity evaluation method for low and intermediate level activated waste generated at nuclear power plants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Makoto Kashiwagi; Garamszeghy, Mike; Lantes, Bertrand

Disposal of low-and intermediate-level activated waste generated at nuclear power plants is being planned or carried out in many countries. The radioactivity concentrations and/or total quantities of long-lived, difficult-to-measure nuclides (DTM nuclides), such as C-14, Ni-63, Nb-94, α emitting nuclides etc., are often restricted by the safety case for a final repository as determined by each country's safety regulations, and these concentrations or amounts are required to be known and declared. With respect to waste contaminated by contact with process water, the Scaling Factor method (SF method), which is empirically based on sampling and analysis data, has been applied asmore » an important method for determining concentrations of DTM nuclides. This method was standardized by the International Organization for Standardization (ISO) and published in 2007 as ISO21238 'Scaling factor method to determine the radioactivity of low and intermediate-level radioactive waste packages generated at nuclear power plants' [1]. However, for activated metal waste with comparatively high concentrations of radioactivity, such as may be found in reactor control rods and internal structures, direct sampling and radiochemical analysis methods to evaluate the DTM nuclides are limited by access to the material and potentially high personnel radiation exposure. In this case, theoretical calculation methods in combination with empirical methods based on remote radiation surveys need to be used to best advantage for determining the disposal inventory of DTM nuclides while minimizing exposure to radiation workers. Pursuant to this objective a standard for the theoretical evaluation of the radioactivity concentration of DTM nuclides in activated waste, is in process through ISO TC85/SC5 (ISO Technical Committee 85: Nuclear energy, nuclear technologies, and radiological protection; Subcommittee 5: Nuclear fuel cycle). The project team for this ISO standard was formed in 2011 and is

RADIOACTIVE DEMONSTRATIONS OF FLUIDIZED BED STEAM REFORMING AS A SUPPLEMENTARY TREATMENT FOR HANFORD'S LOW ACTIVITY WASTE AND SECONDARY WASTES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jantzen, C.; Crawford, C.; Cozzi, A.

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in themore » time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. The Supplemental Treatment chosen will immobilize that portion of the retrieved LAW that is not sent to the WTP's LAW Vitrification facility into a solidified waste form. The solidified waste will then be disposed on the Hanford site in the Integrated Disposal Facility (IDF). In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as Cs-137, I-129, Tc-99, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap. The current waste disposal path for the WTP-SW is to recycle it to the supplemental LAW treatment to avoid a large steady state accumulation in the pretreatment-vitrification loop. Fluidized Bed Steam Reforming (FBSR) offers a moderate temperature (700-750 C) continuous method by which LAW and/or WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates

Production of activated carbons from waste tyres for low temperature NOx control.

PubMed

Al-Rahbi, Amal S; Williams, Paul T

2016-03-01

Waste tyres were pyrolysed in a bench scale reactor and the product chars were chemically activated with alkali chemical agents, KOH, K2CO3, NaOH and Na2CO3 to produce waste tyre derived activated carbons. The activated carbon products were then examined in terms of their ability to adsorb NOx (NO) at low temperature (25°C) from a simulated industrial process flue gas. This study investigates the influence of surface area and porosity of the carbons produced with the different alkali chemical activating agents on NO capture from the simulated flue gas. The influence of varying the chemical activation conditions on the porous texture and corresponding NO removal from the flue gas was studied. The activated carbon sorbents were characterized in relation to BET surface area, micropore and mesopore volumes and chemical composition. The highest NO removal efficiency for the waste tyre derived activated carbons was ∼75% which was obtained with the adsorbent treated with KOH which correlated with both the highest BET surface area and largest micropore volume. In contrast, the waste tyre derived activated carbons prepared using K2CO3, NaOH and Na2CO3 alkali activating agents appeared to have little influence on NO removal from the flue gases. The results suggest problematic waste tyres, have the potential to be converted to activated carbons with NOx removal efficiency comparable with conventionally produced carbons. Copyright © 2016 Elsevier Ltd. All rights reserved.

Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1994-09-01

The Department of Energy`s (DOE`s) planning for the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of the waste. This report estimates volumes, radionuclide activities, and waste forms of GTCC LLW to the year 2035. It groups the waste into four categories, representative of the type of generator or holder of the waste: Nuclear Utilities, Sealed Sources, DOE-Held, and Other Generator. GTCC LLW includes activated metals (activation hardware from reactor operation and decommissioning), process wastes (i.e., resins, filters, etc.), sealed sources, and other wastes routinely generated by users of radioactive material. Estimates reflect the possible effect thatmore » packaging and concentration averaging may have on the total volume of GTCC LLW. Possible GTCC mixed LLW is also addressed. Nuclear utilities will probably generate the largest future volume of GTCC LLW with 65--83% of the total volume. The other generators will generate 17--23% of the waste volume, while GTCC sealed sources are expected to contribute 1--12%. A legal review of DOE`s obligations indicates that the current DOE-Held wastes described in this report will not require management as GTCC LLW because of the contractual circumstances under which they were accepted for storage. This report concludes that the volume of GTCC LLW should not pose a significant management problem from a scientific or technical standpoint. The projected volume is small enough to indicate that a dedicated GTCC LLW disposal facility may not be justified. Instead, co-disposal with other waste types is being considered as an option.« less

Initial results of metal waste form development activities at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, D.D. Jr.; Westphal, B.R.; Hersbt, R.S.

1997-10-01

Argonne National Laboratory is developing a metal alloy to contain metallic waste constituents from the electrometallurgical treatment of spent nuclear fuel. This alloy will contain stainless steel (from stainless steel-clad fuel elements), {approximately}15 wt.% zirconium (from alloy fuel), fission products noble to the process (e.g., Ru, Pd, Tc, etc.), and minor amounts of actinides. The alloy will serve as a final waste form for these components and will be disposed of in a geologic repository. The alloy ingot is produced in an induction furnace situated in a hot cell using Ar cover gas. This paper discusses results from the meltingmore » campaigns that have been initiated at ANL-West to generate the metal waste form using actual process materials. In addition, metal waste form samples have been doped with Tc and selected actinides and are described in the context of how elements of interest partition between various phases in the alloy and how this distribution of elements in the alloy may affect the leaching behavior of the components in an aqueous environment. 3 refs.« less

Initial results of metal waste-form development activities at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Keiser, D.D. Jr.; Westphal, B.R.; Herbst, R.S.

1997-12-01

Argonne National Laboratory (ANL) is developing a metal alloy to contain metallic waste constituent residual from the electrometallurgical treatment of spent nuclear fuel. This alloy will contain stainless steel (from stainless steel-clad fuel elements), {approximately} 15 wt% zirconium (from alloy fuel), fission products noble to the process (e.g., ruthenium, palladium, technetium, etc.), and minor amounts of actinides. The alloy will serve as a final waste form for these components and will be disposed of in a geologic repository. The alloy ingot is produced in an induction furnace situated in a hot cell using argon cover gas. This paper discusses resultsmore » from the melting campaigns that have been initiated at ANL-West to generate the metal waste form using actual process materials. In addition, metal waste form samples have been doped with technetium and selected actinides and are described in the context of how elements of interest partition between various phases in the alloy and how this distribution of elements in the alloy may affect the leaching behavior of the components in an aqueous environment.« less

Membrane Treatment of Aqueous Film Forming Foam (AFFF) Wastes for Recovery of Its Active Ingredients

DTIC Science & Technology

1980-10-01

T ME1MBRANE TREATMENT OF AQUEOUS FILM FORMING FOAM~ (AFFF) WASTES FOR RECOVERY OFI Fts ACTIVE INGREDIENTS FINAL REPORT October 1980 by Edward S. K...OF THIS PAGEOPMn Date AVntr* d)__ ---- Ultrafiltration (UF) and Reverse Osmosis (RO) treatment of Aqueous Film Forming Foam (AFFF) solutions was...of Aqueous Film Forming Foam (AFFF) solutions was investigated to determine the feasibility of employing membrane processes to separate and recover

Hanford Immobilized Low Activity Waste (ILAW) Performance Assessment 2001 Version [Formerly DOE/RL-97-69] [SEC 1 & 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

MANN, F.M.

2000-08-01

The Hanford Immobilized Low-Activity Waste Performance Assessment examines the long-term environmental and human health effects associated with the planned disposal of the vitrified low-activity fraction of waste presently contained in Hanford Site tanks. The tank waste is the byproduct of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste is stored in underground single- and double-shell tanks. The tank waste is to be retrieved, separated into low-activity and high-level fractions, and then immobilized by vitrification. The US. Department of Energy (DOE) plans to dispose of the low-activity fraction in the Hanford Site 200 Eastmore » Area. The high-level fraction will be stored at the Hanford Site until a national repository is approved. This report provides the site-specific long-term environmental information needed by the DOE to modify the current Disposal Authorization Statement for the Hanford Site that would allow the following: construction of disposal trenches; and filling of these trenches with ILAW containers and filler material with the intent to dispose of the containers.« less

A Strategy for Maintenance of the Long-Term Performance Assessment of Immobilized Low-Activity Waste Glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryan, Joseph V.; Freedman, Vicky L.

2016-09-28

Approximately 50 million gallons of high-level radioactive mixed waste has accumulated in 177 buried single- and double-shell tanks at the Hanford Site in southeastern Washington State as a result of the past production of nuclear materials, primarily for defense uses. The United States Department of Energy (DOE) is proceeding with plans to permanently dispose of this waste. Plans call for separating the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions, which will be vitrified at the Hanford Waste Treatment and Immobilization Plant (WTP). Principal radionuclides of concern in LAW are 99Tc, 129I, and U, while non-radioactive contaminantsmore » of concern are Cr and nitrate/nitrite. HLW glass will be sent off-site to an undetermined federal site for deep geological disposal while the much larger volume of immobilized low-activity waste will be placed in the on-site, near-surface Integrated Disposal Facility (IDF).« less

Tc removal from the waste treatment and immobilization plant low-activity waste vitrification off-gas recycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor-Pashow, Kathryn M. L.; McCabe, Daniel J.; Nash, Charles A.

Vitrification of Low Activity Waste in the Hanford Waste Treatment and Immobilization Plant generates a condensate stream from the off-gas processes. Components in this stream are partially volatile and accumulate to high concentrations through recycling, which impacts the waste glass loading and facility throughput. The primary radionuclide that vaporizes and accumulates in the stream is 99Tc. This program is investigating Tc removal via reductive precipitation with stannous chloride to examine the potential for diverting this stream to an alternate disposition path. As a result, research has shown stannous chloride to be effective, and this paper describes results of recent experimentsmore » performed to further mature the technology.« less

Tc removal from the waste treatment and immobilization plant low-activity waste vitrification off-gas recycle

DOE PAGES

Taylor-Pashow, Kathryn M. L.; McCabe, Daniel J.; Nash, Charles A.

2017-03-16

Vitrification of Low Activity Waste in the Hanford Waste Treatment and Immobilization Plant generates a condensate stream from the off-gas processes. Components in this stream are partially volatile and accumulate to high concentrations through recycling, which impacts the waste glass loading and facility throughput. The primary radionuclide that vaporizes and accumulates in the stream is 99Tc. This program is investigating Tc removal via reductive precipitation with stannous chloride to examine the potential for diverting this stream to an alternate disposition path. As a result, research has shown stannous chloride to be effective, and this paper describes results of recent experimentsmore » performed to further mature the technology.« less

Sorption of mercury onto waste material derived low-cost activated carbon

NASA Astrophysics Data System (ADS)

Bhakta, Jatindra N.; Rana, Sukanta; Lahiri, Susmita; Munekage, Yukihiro

2017-03-01

The present study was performed to develop the low-cost activated carbon (AC) from some waste materials as potential mercury (Hg) sorbent to remove high amount of Hg from aqueous phase. The ACs were prepared from banana peel, orange peel, cotton fiber and paper wastes by pyrolysis and characterized by analyzing physico-chemical properties and Hg sorption capacity. The Brunauer Emmett and Teller surface areas (cotton 138 m2/g; paper 119 m2/g), micropore surface areas (cotton 65 m2/g; paper 54 m2/g) and major constituent carbon contents (cotton 95.04 %; paper 94.4 %) were higher in ACs of cotton fiber and paper wastes than the rest two ACs. The Hg sorption capacities and removal percentages were greater in cotton and paper wastes-derived ACs compared to those of the banana and orange peels. The results revealed that elevated Hg removal ability of cotton and paper wastes-derived ACs is largely regulated by their surface area, porosity and carbon content properties. Therefore, ACs of cotton and paper wastes were identified as potential sorbent among four developed ACs to remove high amount of Hg from aqueous phase. Furthermore, easily accessible precursor material, simple preparation process, favorable physico-chemical properties and high Hg sorption capacity indicated that cotton and paper wastes-derived ACs could be used as potential and low-cost sorbents of Hg for applying in practical field to control the severe effect of Hg contamination in the aquatic environment to avoid its human and environmental health risks.

Midwest Interstate Low-Level Radioactive Waste Commission annual report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1988-08-01

In 1980, Congress passed the Low-Level Radioactive Waste Policy Act. This Act provided for a new approach to the disposal of low-level radioactive waste. It assigned each state responsibility for the disposal of low-level radioactive waste generated within its borders, and it authorized states to enter into compacts for the purpose of operating regional disposal facilities. It also authorized compacts to restrict the use of regional disposal facilities to only member states. To meet their obligations under the Act, Indiana, Iowa, Michigan, Minnesota, Missouri, Ohio and Wisconsin formed the Midwest Interstate Low-Level Radioactive Waste Compact. The Compact was ratified bymore » each of the state legislatures and by Congress. The Compact established the Midwest Interstate Low-Level Radioactive Waste Commission, composed on one representative appointed by the Governor or Legislature of each member state. Article 3 of the compact requires that the Commission prepare an annual report regarding the activities and actions of the Commission. It also requires that the annual report be distributed to the Governors and legislative leaders in the member states. The Commission's Bylaw Article 12 requires the annual report to cover the preceding fiscal year, and to be distributed in August of each year. The Bylaw also requires that an annual audit, prepared by a certified public accountant, be included as part of the annual report. 3 figs.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamson, D.; Nash, C.; Mcclane, D.

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 tomore » 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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peeler, David K.; Kim, Dong-Sang; Vienna, John D.

2015-11-01

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, keymore » 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. The purpose of this advanced LAW glass research and development plan is to identify the near-term, mid-term, and longer-term research and development activities required to develop and validate advanced LAW glasses, property-composition models and their uncertainties, and an advanced glass algorithm to support WTP facility operations, including both Direct Feed LAW and full pretreatment flowsheets. Data are needed to develop, validate, and implement 1) new glass property-composition models and 2) a new glass formulation algorithm. Hence, this plan integrates specific studies associated with increasing the Na2O and SO3/halide concentrations in glass, because these components will ultimately dictate waste loadings for LAW vitrification. Of equal importance is the development of an efficient and economic strategy for 99Tc management. Specific and detailed studies are being implemented to understand the fate of Tc

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kruger, A. A.; Peeler, D. K.; Kim, D. S.

2015-11-23

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, keymore » 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.« less

A U-bearing composite waste form for electrochemical processing wastes

NASA Astrophysics Data System (ADS)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

2018-04-01

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases.

DuraLith geopolymer waste form for Hanford secondary waste: correlating setting behavior to hydration heat evolution.

PubMed

Xu, Hui; Gong, Weiliang; Syltebo, Larry; Lutze, Werner; Pegg, Ian L

2014-08-15

The binary furnace slag-metakaolin DuraLith geopolymer waste form, which has been considered as one of the candidate waste forms for immobilization of certain Hanford secondary wastes (HSW) from the vitrification of nuclear wastes at the Hanford Site, Washington, was extended to a ternary fly ash-furnace slag-metakaolin system to improve workability, reduce hydration heat, and evaluate high HSW waste loading. A concentrated HSW simulant, consisting of more than 20 chemicals with a sodium concentration of 5 mol/L, was employed to prepare the alkaline activating solution. Fly ash was incorporated at up to 60 wt% into the binder materials, whereas metakaolin was kept constant at 26 wt%. The fresh waste form pastes were subjected to isothermal calorimetry and setting time measurement, and the cured samples were further characterized by compressive strength and TCLP leach tests. This study has firstly established quantitative linear relationships between both initial and final setting times and hydration heat, which were never discovered in scientific literature for any cementitious waste form or geopolymeric material. The successful establishment of the correlations between setting times and hydration heat may make it possible to efficiently design and optimize cementitious waste forms and industrial wastes based geopolymers using limited testing results. Copyright © 2014 Elsevier B.V. All rights reserved.

A U-bearing composite waste form for electrochemical processing wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, X.; Ebert, W. L.; Indacochea, J. E.

Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phasesmore » that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.« less

SUBGRADE MONOLITHIC ENCASEMENT STABILIZATION OF CATEGORY 3 LOW LEVEL WASTE (LLW)

DOE Office of Scientific and Technical Information (OSTI.GOV)

PHILLIPS, S.J.

2004-02-03

A highly efficient and effective technology has been developed and is being used for stabilization of Hazard Category 3 low-level waste at the U.S. Department of Energy's Hanford Site. Using large, structurally interconnected monoliths, which form one large monolith that fills a waste disposal trench, the patented technology can be used for final internment of almost any hazardous, radioactive, or toxic waste or combinations of these waste materials packaged in a variety of sizes, shapes, and volumes within governmental regulatory limits. The technology increases waste volumetric loading by 100 percent, area use efficiency by 200 percent, and volumetric configuration efficiencymore » by more than 500 percent over past practices. To date, in excess of 2,010 m{sup 3} of contact-handled and remote-handled low-level radioactive waste have been interned using this patented technology. Additionally, in excess of 120 m{sup 3} of low-level radioactive waste requiring stabilization in low-diffusion coefficient waste encasement matrix has been disposed using this technology. Greater than five orders of magnitude in radiation exposure reduction have been noted using this method of encasement of Hazard Category 3 waste. Additionally, exposure monitored at all monolith locations produced by the slip form technology is less than 1.29 x E-07 C {center_dot} kg{sup -1}. Monolithic encasement of Hazard Category 3 low-level waste and other waste category materials may be successfully accomplished using this technology at nominally any governmental or private sector waste disposal facility. Additionally, other waste materials consisting of hazardous, radioactive, toxic, or mixed waste materials can be disposed of using the monolithic slip form encasement technology.« less

Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

NASA Astrophysics Data System (ADS)

Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

2014-09-01

The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.

Development of iron phosphate ceramic waste form to immobilize radioactive waste solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Choi, Jongkwon; Um, Wooyong; Choung, Sungwook

The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions weremore » 16 MPa and 19 MPa for LiCl-KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4,136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl-KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl-KCl eutectic salt and off-gas scrubber solutions with significant cost savings.« less

Demonstration of sulfur solubility determinations in high waste loading, low-activity waste glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K. M.

2016-04-25

A method recommended by Pacific Northwest National Laboratory (PNNL) for sulfate solubility determinations in simulated low-activity waste glasses was demonstrated using three compositions from a recent Hanford high waste loading glass study. Sodium and sulfate concentrations in the glasses increased after each re-melting step. Visual observations of the glasses during the re-melting process reflected the changes in composition. The measured compositions showed that the glasses met the targeted values. The amount of SO 3 retained in the glasses after washing was relatively high, ranging from 1.6 to 2.6 weight percent (wt %). Measured SnO 2 concentrations were notably low inmore » all of the study glasses. The composition of the wash solutions should be measured in future work to determine whether SnO 2 is present with the excess sulfate washed from the glass. Increases in batch size and the amount of sodium sulfate added did not have a measureable impact on the amount of sulfate retained in the glass, although this was tested for only a single glass composition. A batch size of 250 g and a sodium sulfate addition targeting 7 wt %, as recommended by PNNL, will be used in future experiments.« less

Iodosodalite Waste Forms from Low-Temperature Aqueous Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nam, Junghune; Chong, Saehwa; Riley, Brian J.

ABSTRACT Nuclear energy is one option to meet rising electricity demands, although one concern of this technology is the proper capture and storage of radioisotopes produced during fission processes. One of the more difficult radioisotopes is 129I due to its volatility and poor solubility in traditional waste forms such as borosilicate glass. Iodosodalite has been previously proposed as a viable candidate to immobilize iodine due to high iodine loading and good chemical durability. Iodosodalite was traditionally synthesized using solid state and hydrothermal techniques, but this paper discusses an aqueous synthesis approach to optimize and maximize the iodosodalite yield. Products weremore » pressed into pellets and fired with glass binders. Chemical durability and iodine retention results are included.« less

Preparation and evaporation of Hanford Waste treatment plant direct feed low activity waste effluent management facility simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamson, D.; Nash, C.; Howe, A.

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 involves concentrating the condensate in a new evaporator at the Effluent Management Facility (EMF) and returning it to themore » LAW melter. The LMOGC stream will contain components, e.g. halides and sulfates, that are volatile at melter temperatures, have limited solubility in glass waste forms, and present a material corrosion concern. Because this stream will recycle within WTP, these components are expected to accumulate in the LMOGC 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 glass and is a key objective of this program. In order to determine the disposition path, it is key to experimentally determine the fate of contaminants. To do this, testing is needed to account for the buffering chemistry of the components, determine the achievable evaporation end point, identify insoluble solids that form, determine the formation and distribution of key regulatoryimpacting constituents, and generate an aqueous stream that can be used in testing of the subsequent immobilization step. This overall program examines the potential treatment and immobilization of the LMOGC stream to enable alternative disposal. The objective of this task was to (1) prepare a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, (2) demonstrate evaporation in order to predict the final composition of the effluents from

Immobilization of Technetium in a Metallic Waste Form

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.M. Frank; D. D. Keiser, Jr.; K. C. Marsden

Fission-product technetium accumulated during treatment of spent nuclear fuel will ultimately be disposed of in a geological repository. The exact form of Tc for disposal has yet to be determined; however, a reasonable solution is to incorporate elemental Tc into a metallic waste form similar to the waste form produced during the pyrochemical treatment of spent, sodium-bonded fuel. This metal waste form, produced at the Idaho National Laboratory, has undergone extensive qualification examination and testing for acceptance to the Yucca Mountain geological repository. It is from this extensive qualification effort that the behavior of Tc and other fission products inmore » the waste form has been elucidated, and that the metal waste form is extremely robust in the retention of fission products, such as Tc, in repository like conditions. This manuscript will describe the metal waste form, the behavior of Tc in the waste form; and current research aimed at determining the maximum possible loading of Tc into the metal waste and subsequent determination of the performance of high Tc loaded metal waste forms.« less

Development of Alternative Technetium Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Czerwinski, Kenneth

2013-09-13

The UREX+1 process is under consideration for the separation of transuranic elements from spent nuclear fuel. The first steps of this process extract the fission product technicium-99 ({sup 99}Tc) into an organic phase containing tributylphosphate together with uranium. Treatment of this stream requires the separation of Tc from U and placement into a suitable waste storage form. A potential candidate waste form involves immobilizing the Tc as an alloy with either excess metallic zirconium or stainless steel. Although Tc-Zr alloys seem to be promising waste forms, alternative materials must be investigated. Innovative studies related to the synthesis and behavior ofmore » a different class of Tc materials will increase the scientific knowledge related to development of Tc waste forms. These studies will also provide a better understanding of the behavior of {sup 99}Tc in repository conditions. A literature survey has selected promising alternative waste forms for further study: technetium metallic alloys, nitrides, oxides, sulfides, and pertechnetate salts. The goals of this project are to 1) synthesize and structurally characterize relevant technetium materials that may be considered as waste forms, 2) investigate material behavior in solution under different conditions of temperature, electrochemical potential, and radiation, and 3) predict the long-term behavior of these materials.« less

Spent fuel treatment and mineral waste form development at Argonne National Laboratory-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Benedict, R.W.; Bateman, K.

1996-07-01

At Argonne National Laboratory-West (ANL-West) there are several thousand kilograms of metallic spent nuclear fuel containing bond sodium. This fuel will be treated in the Fuel Conditioning Facility (FCF) at ANL-West to produce stable waste forms for storage and disposal. Both mineral and metal high-level waste forms will be produced. The mineral waste form will contain the active metal fission products and the transuranics. Cold small-scale waste form testing has been on-going at Argonne in Illinois. Large-scale testing is commencing at ANL-West.

Iron-phosphate ceramics for solidification of mixed low-level waste

DOEpatents

Aloy, Albert S.; Kovarskaya, Elena N.; Koltsova, Tatiana I.; Macheret, Yevgeny; Medvedev, Pavel G.; Todd, Terry

2000-01-01

A method of immobilizing mixed low-level waste is provided which uses low cost materials and has a relatively long hardening period. The method includes: forming a mixture of iron oxide powders having ratios, in mass %, of FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 equal to 25-40:40-10:35-50, or weighing a definite amount of magnetite powder. Metallurgical cinder can also be used as the source of iron oxides. A solution of the orthophosphoric acid, or a solution of the orthophosphoric acid and ferric oxide, is formed and a powder phase of low-level waste and the mixture of iron oxide powders or cinder (or magnetite powder) is also formed. The acid solution is mixed with the powder phase to form a slurry with the ratio of components (mass %) of waste:iron oxide powders or magnetite:acid solution=30-60:15-10:55-30. The slurry is blended to form a homogeneous mixture which is cured at room temperature to form the final product.

Physical, Hydraulic, and Transport Properties of Sediments and Engineered Materials Associated with Hanford Immobilized Low-Activity Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rockhold, Mark L.; Zhang, Z. F.; Meyer, Philip D.

2015-02-28

Current plans for treatment and disposal of immobilized low-activity waste (ILAW) from Hanford’s underground waste storage tanks include vitrification and storage of the glass waste form in a nearsurface disposal facility. This Integrated Disposal Facility (IDF) is located in the 200 East Area of the Hanford Central Plateau. Performance assessment (PA) of the IDF requires numerical modeling of subsurface flow and reactive transport processes over very long periods (thousands of years). The models used to predict facility performance require parameters describing various physical, hydraulic, and transport properties. This report provides updated estimates of physical, hydraulic, and transport properties and parametersmore » for both near- and far-field materials, intended for use in future IDF PA modeling efforts. Previous work on physical and hydraulic property characterization for earlier IDF PA analyses is reviewed and summarized. For near-field materials, portions of this document and parameter estimates are taken from an earlier data package. For far-field materials, a critical review is provided of methodologies used in previous data packages. Alternative methods are described and associated parameters are provided.« less

Material Recover and Waste Form Development--2016 Accomplishments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todd, Terry A.; Vienna, John; Paviet, Patricia

The Material Recovery and Waste Form Development (MRWFD) Campaign under the U.S. Department of Energy (DOE) Fuel Cycle Technologies (FCT) Program is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress (April 2010). This MRWFD accomplishments report summarizes the results of the research and development (R&D) efforts performed within MRWFD in Fiscal Year (FY) 2016. Each section of the report contains an overview of the activities, results, technical point of contact, applicable references, and documents produced during the FY. Thismore » report briefly outlines campaign management and integration activities but primarily focuses on the many technical accomplishments of FY 2016. The campaign continued to use an engineering-driven, science-based approach to maintain relevance and focus.« less

Special Analysis for the Disposal of the Lawrence Livermore National Laboratory Low Activity Beta/Gamma Sources Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shott, Gregory J.

This special analysis (SA) evaluates whether the Lawrence Livermore National Laboratory (LLNL) Low Activity Beta/Gamma Sources waste stream (BCLALADOEOSRP, Revision 0) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada National Security Site (NNSS). The LLNL Low Activity Beta/Gamma Sources waste stream consists of sealed sources that are no longer needed. The LLNL Low Activity Beta/Gamma Sources waste stream required a special analysis because cobalt-60 (60Co), strontium-90 (90Sr), cesium-137 (137Cs), and radium-226 (226Ra) exceeded the NNSS Waste Acceptance Criteria (WAC) Action Levels (U.S. Department of Energy, National Nuclearmore » Security Administration Nevada Field Office [NNSA/NFO] 2015). The results indicate that all performance objectives can be met with disposal of the LLNL Low Activity Beta/Gamma Sources in a SLB trench. The LLNL Low Activity Beta/Gamma Sources waste stream is suitable for disposal by SLB at the Area 5 RWMS. However, the activity concentration of 226Ra listed on the waste profile sheet significantly exceeds the action level. Approval of the waste profile sheet could potentially allow the disposal of high activity 226Ra sources. To ensure that the generator does not include large 226Ra sources in this waste stream without additional evaluation, a control is need on the maximum 226Ra inventory. A limit based on the generator’s estimate of the total 226Ra inventory is recommended. The waste stream is recommended for approval with the control that the total 226Ra inventory disposed shall not exceed 5.5E10 Bq (1.5 Ci).« less

Effect of Technetium-99 sources on its retention in low activity waste glass

NASA Astrophysics Data System (ADS)

Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong; Wang, Guohui; Schweiger, Michael J.; Soderquist, Chuck Z.; Lukens, Wayne; Kruger, Albert A.

2018-05-01

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2•2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.

Hanford Low-Activity Waste Processing: Demonstration of the Off-Gas Recycle Flowsheet - 13443

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramsey, William G.; Esparza, Brian P.

2013-07-01

Vitrification of Hanford Low-Activity Waste (LAW) is nominally the thermal conversion and incorporation of sodium salts and radionuclides into borosilicate glass. One key radionuclide present in LAW is technetium-99. Technetium-99 is a low energy, long-lived beta emitting radionuclide present in the waste feed in concentrations on the order of 1-10 ppm. The long half-life combined with a high solubility in groundwater results in technetium-99 having considerable impact on performance modeling (as potential release to the environment) of both the waste glass and associated secondary waste products. The current Hanford Tank Waste Treatment and Immobilization Plant (WTP) process flowsheet calls formore » the recycle of vitrification process off-gas condensates to maximize the portion of technetium ultimately immobilized in the waste glass. This is required as technetium acts as a semi-volatile specie, i.e. considerable loss of the radionuclide to the process off-gas stream can occur during the vitrification process. To test the process flowsheet assumptions, a prototypic off-gas system with recycle capability was added to a laboratory melter (on the order of 1/200 scale) and testing performed. Key test goals included determination of the process mass balance for technetium, a non-radioactive surrogate (rhenium), and other soluble species (sulfate, halides, etc.) which are concentrated by recycling off-gas condensates. The studies performed are the initial demonstrations of process recycle for this type of liquid-fed melter system. This paper describes the process recycle system, the waste feeds processed, and experimental results. Comparisons between data gathered using process recycle and previous single pass melter testing as well as mathematical modeling simulations are also provided. (authors)« less

Equilibrium Temperature Profiles within Fission Product Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaminski, Michael D.

2016-10-01

We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.

Alternative High-Performance Ceramic Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sundaram, S. K.

This final report (M5NU-12-NY-AU # 0202-0410) summarizes the results of the project titled “Alternative High-Performance Ceramic Waste Forms,” funded in FY12 by the Nuclear Energy University Program (NEUP Project # 12-3809) being led by Alfred University in collaboration with Savannah River National Laboratory (SRNL). The overall focus of the project is to advance fundamental understanding of crystalline ceramic waste forms and to demonstrate their viability as alternative waste forms to borosilicate glasses. We processed single- and multiphase hollandite waste forms based on simulated waste streams compositions provided by SRNL based on the advanced fuel cycle initiative (AFCI) aqueous separation process developed in the Fuel Cycle Research and Development (FCR&D). For multiphase simulated waste forms, oxide and carbonate precursors were mixed together via ball milling with deionized water using zirconia media in a polyethylene jar for 2 h. The slurry was dried overnight and then separated from the media. The blended powders were then subjected to melting or spark plasma sintering (SPS) processes. Microstructural evolution and phase assemblages of these samples were studied using x-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersion analysis of x-rays (EDAX), wavelength dispersive spectrometry (WDS), transmission electron spectroscopy (TEM), selective area x-ray diffraction (SAXD), and electron backscatter diffraction (EBSD). These results showed that the processing methods have significant effect on the microstructure and thus the performance of these waste forms. The Ce substitution into zirconolite and pyrochlore materials was investigated using a combination of experimental (in situ XRD and x-ray absorption near edge structure (XANES)) and modeling techniques to study these single phases independently. In zirconolite materials, a transition from the 2M to the 4M polymorph was observed with increasing Ce content. The resulting

Waste Treatment Technology Process Development Plan For Hanford Waste Treatment Plant Low Activity Waste Recycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

2013-08-29

The purpose of this Process Development Plan is to summarize the objectives and plans for the technology development activities for an alternative path for disposition of the recycle stream that will be generated in the Hanford Waste Treatment Plant Low Activity Waste (LAW) vitrification facility (LAW Recycle). This plan covers the first phase of the development activities. The baseline plan for disposition of this stream is to recycle it to the WTP Pretreatment Facility, where it will be concentrated by evaporation and returned to the LAW vitrification facility. Because this stream contains components that are volatile at melter temperatures andmore » are also problematic for the glass waste form, they accumulate in the Recycle stream, exacerbating their impact on the number of LAW glass containers. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and reducing the halides in the Recycle is a key component of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, this stream does not have a proven disposition path, and resolving this gap becomes vitally important. This task seeks to examine the impact of potential future disposition of this stream in the Hanford tank farms, and to develop a process that will remove radionuclides from this stream and allow its diversion to another disposition path, greatly decreasing the LAW vitrification mission duration and quantity of glass waste. The origin of this LAW Recycle stream will be from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover or precipitates of scrubbed components (e.g. carbonates). The

Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.

2010-09-23

In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development ofmore » a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both

Effect of Technetium-99 sources on its retention in low activity waste glass

DOE PAGES

Luksic, Steven A.; Kim, Dong Sang; Um, Wooyong; ...

2018-03-02

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO 2∙2H 2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. Here, we postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Finally,more » additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.« less

Effect of Technetium-99 sources on its retention in low activity waste glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO2∙2H2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with hexavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Additional studies are neededmore » to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from glass melt.« less

Effect of Technetium-99 sources on its retention in low activity waste glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luksic, Steven A.; Kim, Dong Sang; Um, Wooyong

Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO 2∙2H 2O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. Here, we postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generally accepted idea. Finally,more » additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.« less

Effect of Technetium-99 sources on its retention in low activity waste glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong

© 2018 Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO 2 ∙2H 2 O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generallymore » accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.« less

Effect of Technetium-99 sources on its retention in low activity waste glass

DOE PAGES

Luksic, Steven A.; Kim, Dong-Sang; Um, Wooyong; ...

2018-05-01

© 2018 Small-scale crucible melting tests on simulated waste glass were performed with technetium-99 (Tc-99) introduced as different species in a representative low activity waste simulant. The glass saw an increase in Tc-99 retention when TcO 2 ∙2H 2 O and various Tc-minerals containing reduced tetravalent Tc were used compared to tests in which pertechnetate with heptavalent Tc was used. We postulate that the increase of Tc retention is likely caused by different reaction paths for Tc incorporation into glass during early stages of melting, rather than the low volatility of reduced tetravalent Tc compounds, which has been a generallymore » accepted idea. Additional studies are needed to clarify the exact mechanisms relevant to the effect of reduced Tc compounds on Tc incorporation into or volatilization from the glass melt.« less

Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steven Frank; Hwan Seo Park; Yung Zun Cho

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration betweenmore » US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.« less

I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

DOE Office of Scientific and Technical Information (OSTI.GOV)

S.M. Frank

Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomicmore » Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration

Engineering-Scale Demonstration of DuraLith and Ceramicrete Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Josephson, Gary B.; Westsik, Joseph H.; Pires, Richard P.

2011-09-23

To support the selection of a waste form for the liquid secondary wastes from the Hanford Waste Immobilization and Treatment Plant, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing on four candidate waste forms. Two of the candidate waste forms have not been developed to scale as the more mature waste forms. This work describes engineering-scale demonstrations conducted on Ceramicrete and DuraLith candidate waste forms. Both candidate waste forms were successfully demonstrated at an engineering scale. A preliminary conceptual design could be prepared for full-scale production of the candidate waste forms. However, both waste forms are stillmore » too immature to support a detailed design. Formulations for each candidate waste form need to be developed so that the material has a longer working time after mixing the liquid and solid constituents together. Formulations optimized based on previous lab studies did not have sufficient working time to support large-scale testing. The engineering-scale testing was successfully completed using modified formulations. Further lab development and parametric studies are needed to optimize formulations with adequate working time and assess the effects of changes in raw materials and process parameters on the final product performance. Studies on effects of mixing intensity on the initial set time of the waste forms are also needed.« less

Development of chemically bonded phosphate ceramics for stabilizing low-level mixed wastes

NASA Astrophysics Data System (ADS)

Jeong, Seung-Young

1997-11-01

Novel chemically bonded phosphate ceramics have been developed by acid-base reactions between magnesium oxide and an acid phosphate at room temperature for stabilizing U.S. Department of Energy's low-level mixed waste streams that include hazardous chemicals and radioactive elements. Newberyite (MgHPOsb4.3Hsb2O)-rich magnesium phosphate ceramic was formed by an acid-base reaction between phosphoric acid and magnesium oxide. The reaction slurry, formed at room-temperature, sets rapidly and forms stable mineral phases of newberyite, lunebergite, and residual MgO. Rapid setting also generates heat due to exothermic acid-base reaction. The reaction was retarded by partially neutralizing the phosphoric acid solution by adding sodium or potassium hydroxide. This reduced the rate of reaction and heat generation and led to a practical way of producing novel magnesium potassium phosphate ceramic. This ceramic was formed by reacting stoichiometric amount of monopotassium dihydrogen phosphate crystals, MgO, and water, forming pure-phase of MgKPOsb4.6Hsb2O (MKP) with moderate exothermic reaction. Using this chemically bonded phosphate ceramic matrix, low-level mixed waste streams were stabilized, and superior waste forms in a monolithic structure were developed. The final waste forms showed low open porosity and permeability, and higher compression strength than the Land Disposal Requirements (LDRs). The novel MKP ceramic technology allowed us to develop operational size waste forms of 55 gal with good physical integrity. In this improved waste form, the hazardous contaminants such as RCRA heavy metals (Hg, Pb, Cd, Cr, Ni, etc) were chemically fixed by their conversion into insoluble phosphate forms and physically encapsulated by the phosphate ceramic. In addition, chemically bonded phosphate ceramics stabilized radioactive elements such U and Pu. This was demonstrated with a detailed stabilization study on cerium used as a surrogate (chemically equivalent but nonradioactive

I-NERI Annual Technical Progress Report 2007-004-K Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

S. Frank

The current method for the immobilization of fission products that accumulate in electrorefiner salt during the electrochemical processing of used metallic nuclear fuel is to encapsulate the electrorefiner salt in a glass-bonded sodalite ceramic waste form. This process was developed by Argonne National Laboratory in the USA and is currently performed at the Idaho National Laboratory for the treatment of Experimental Breeder Reactor-II (EBR-II) used fuel. This process utilizes a “once-through” option for the disposal of spent electrorefiner salt; where, after the treatment of the EBR-II fuel, the electrorefiner salt containing the active fission products will be disposed of inmore » the ceramic waste form (CWF). The CWF produced will have low fission product loading of approximately 2 to 5 weight percent due to the limited fuel inventory currently being processed. However; the design and implementation of advanced electrochemical processing facilities to treat used fuel would process much greater quantities fuel. With an advanced processing facility, it would be necessary to selectively remove fission products from the electrorefiner salt for salt recycle and to concentrate the fission products to reduce the volume of high-level waste from the treatment facility. The Korean Atomic Energy Research Institute and the Idaho National Laboratory have been collaborating on I-NERI research projects for a number of years to investigate both aspects of selective fission product separation from electrorefiner salt, and to develop advanced waste forms for the immobilization of the collected fission products. The first joint KAERI/INL I-NERI project titled: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels, was successfully completed in 2009 by concentrating and isolating fission products from actual electrorefiner salt used for the treated used EBR-II fuel. Two separation methods were tested and from these tests

Apatite and sodalite based glass-bonded waste forms for immobilization of 129I and mixed halide radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goel, Ashutosh; McCloy, John S.; Riley, Brian J.

The goal of the project was to utilize the knowledge accumulated by the team, in working with minerals for chloride wastes and biological apatites, toward the development of advanced waste forms for immobilizing 129I and mixed-halide wastes. Based on our knowledge, experience, and thorough literature review, we had selected two minerals with different crystal structures and potential for high chemical durability, sodalite and CaP/PbV-apatite, to form the basis of this project. The focus of the proposed effort was towards: (i) low temperature synthesis of proposed minerals (iodine containing sodalite and apatite) leading to the development of monolithic waste forms, (ii)more » development of a fundamental understanding of the atomic-scale to meso-scale mechanisms of radionuclide incorporation in them, and (iii) understanding of the mechanism of their chemical corrosion, alteration mechanism, and rates. The proposed work was divided into four broad sections. deliverables. 1. Synthesis of materials 2. Materials structural and thermal characterization 3. Design of glass compositions and synthesis glass-bonded minerals, and 4. Chemical durability testing of materials.« less

Laboratory Testing of Bulk Vitrified Low-Activity Waste Forms to Support the 2005 Integrated Disposal Facility Performance Assessment. Erratum

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Gary L.

2016-09-06

This report refers to or contains K g values for glasses LAWA44, LAWB45 and LAWC22 affected by calculations errors as identified by Papathanassiu et al. (2011). The corrected K g values are reported in an erratum included in the revised version of the original report. The revised report can be referenced as follows: Pierce E. M. et al. (2004) Waste Form Release Data Package for the 2005 Integrated Disposal Facility Performance Assessment. PNNL-14805 Rev. 0 Erratum. Pacific Northwest National Laboratory, Richland, WA, USA.

Final report on cermet high-level waste forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kobisk, E.H.; Quinby, T.C.; Aaron, W.S.

1981-08-01

Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures.

Secondary Waste Form Screening Test Results—THOR® Fluidized Bed Steam Reforming Product in a Geopolymer Matrix

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pires, Richard P.; Westsik, Joseph H.; Serne, R. Jeffrey

2011-07-14

Screening tests are being conducted to evaluate waste forms for immobilizing secondary liquid wastes from the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Plans are underway to add a stabilization treatment unit to the Effluent Treatment Facility to provide the needed capacity for treating these wastes from WTP. The current baseline is to use a Cast Stone cementitious waste form to solidify the wastes. Through a literature survey, DuraLith alkali-aluminosilicate geopolymer, fluidized-bed steam reformation (FBSR) granular product encapsulated in a geopolymer matrix, and a Ceramicrete phosphate-bonded ceramic were identified both as candidate waste forms and alternatives to the baseline.more » These waste forms have been shown to meet waste disposal acceptance criteria, including compressive strength and universal treatment standards for Resource Conservation and Recovery Act (RCRA) metals (as measured by the toxicity characteristic leaching procedure [TCLP]). Thus, these non-cementitious waste forms should also be acceptable for land disposal. Information is needed on all four waste forms with respect to their capability to minimize the release of technetium. Technetium is a radionuclide predicted to be in the secondary liquid wastes in small quantities, but the Integrated Disposal Facility (IDF) risk assessment analyses show that technetium, even at low mass, produces the largest contribution to the estimated IDF disposal impacts to groundwater.« less

Performance of a Steel/Oxide Composite Waste Form for Combined Waste Steams from Advanced Electrochemical Processes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indacochea, J. E.; Gattu, V. K.; Chen, X.

The results of electrochemical corrosion tests and modeling activities performed collaboratively by researchers at the University of Illinois at Chicago and Argonne National Laboratory as part of workpackage NU-13-IL-UIC-0203-02 are summarized herein. The overall objective of the project was to develop and demonstrate testing and modeling approaches that could be used to evaluate the use of composite alloy/ceramic materials as high-level durable waste forms. Several prototypical composite waste form materials were made from stainless steels representing fuel cladding, reagent metals representing metallic fuel waste streams, and reagent oxides representing oxide fuel waste streams to study the microstructures and corrosion behaviorsmore » of the oxide and alloy phases. Microelectrodes fabricated from small specimens of the composite materials were used in a series of electrochemical tests to assess the corrosion behaviors of the constituent phases and phase boundaries in an aggressive acid brine solution at various imposed surface potentials. The microstructures were characterized in detail before and after the electrochemical tests to relate the electrochemical responses to changes in both the electrode surface and the solution composition. The results of microscopic, electrochemical, and solution analyses were used to develop equivalent circuit and physical models representing the measured corrosion behaviors of the different materials pertinent to long-term corrosion behavior. This report provides details regarding (1) the production of the composite materials, (2) the protocol for the electrochemical measurements and interpretations of the responses of multi-phase alloy and oxide composites, (3) relating corrosion behaviors to microstructures of multi-phase alloys based on 316L stainless steel and HT9 (410 stainless steel was used as a substitute) with added Mo, Ni, and/or Mn, and (4) modeling the corrosion behaviors and rates of several alloy/oxide composite

Waste forms, packages, and seals working group summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sridhar, N.

1995-09-01

This article is a summary of the proceedings of a group discussion which took place at the Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste in San Antonio, Texas on July 22-25, 1991. The working group concentrated on the subject of radioactive waste forms and packaging. Also included is a description of the use of natural analogs in waste packaging, container materials and waste forms.

Radionuclide Retention Mechanisms in Secondary Waste-Form Testing: Phase II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Um, Wooyong; Valenta, Michelle M.; Chung, Chul-Woo

2011-09-26

This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate candidate stabilization technologies that have the potential to successfully treat liquid secondary waste stream effluents produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). WRPS is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF, a multi-waste, treatment-and-storage unit that has been permitted under the Resource Conservation and Recovery Act (RCRA), can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needsmore » to be operational by 2018 to receive secondary liquid waste generated during operation of the WTP. The STU will provide the additional capacity needed for ETF to process the increased volume of secondary waste expected to be produced by WTP. This report on radionuclide retention mechanisms describes the testing and characterization results that improve understanding of radionuclide retention mechanisms, especially for pertechnetate, {sup 99}TcO{sub 4}{sup -} in four different waste forms: Cast Stone, DuraLith alkali aluminosilicate geopolymer, encapsulated fluidized bed steam reforming (FBSR) product, and Ceramicrete phosphate bonded ceramic. These data and results will be used to fill existing data gaps on the candidate technologies to support a decision-making process that will identify a subset of the candidate waste forms that are most promising and should undergo further performance testing.« less

40 CFR 761.205 - Notification of PCB waste activity (EPA Form 7710-53).

Code of Federal Regulations, 2011 CFR

2011-07-01

... (CONTINUED) TOXIC SUBSTANCES CONTROL ACT POLYCHLORINATED BIPHENYLS (PCBs) MANUFACTURING, PROCESSING, DISTRIBUTION IN COMMERCE, AND USE PROHIBITIONS PCB Waste Disposal Records and Reports § 761.205 Notification of... 40 Protection of Environment 31 2011-07-01 2011-07-01 false Notification of PCB waste activity...

40 CFR 761.205 - Notification of PCB waste activity (EPA Form 7710-53).

Code of Federal Regulations, 2014 CFR

2014-07-01

... (CONTINUED) TOXIC SUBSTANCES CONTROL ACT POLYCHLORINATED BIPHENYLS (PCBs) MANUFACTURING, PROCESSING, DISTRIBUTION IN COMMERCE, AND USE PROHIBITIONS PCB Waste Disposal Records and Reports § 761.205 Notification of... 40 Protection of Environment 31 2014-07-01 2014-07-01 false Notification of PCB waste activity...

40 CFR 761.205 - Notification of PCB waste activity (EPA Form 7710-53).

Code of Federal Regulations, 2013 CFR

2013-07-01

... (CONTINUED) TOXIC SUBSTANCES CONTROL ACT POLYCHLORINATED BIPHENYLS (PCBs) MANUFACTURING, PROCESSING, DISTRIBUTION IN COMMERCE, AND USE PROHIBITIONS PCB Waste Disposal Records and Reports § 761.205 Notification of... 40 Protection of Environment 32 2013-07-01 2013-07-01 false Notification of PCB waste activity...

40 CFR 761.205 - Notification of PCB waste activity (EPA Form 7710-53).

Code of Federal Regulations, 2012 CFR

2012-07-01

... (CONTINUED) TOXIC SUBSTANCES CONTROL ACT POLYCHLORINATED BIPHENYLS (PCBs) MANUFACTURING, PROCESSING, DISTRIBUTION IN COMMERCE, AND USE PROHIBITIONS PCB Waste Disposal Records and Reports § 761.205 Notification of... 40 Protection of Environment 32 2012-07-01 2012-07-01 false Notification of PCB waste activity...

National low-level waste management program radionuclide report series, Volume 15: Uranium-238

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adams, J.P.

1995-09-01

This report, Volume 15 of the National Low-Level Waste Management Program Radionuclide Report Series, discusses the radiological and chemical characteristics of uranium-238 ({sup 238}U). The purpose of the National Low-Level Waste Management Program Radionuclide Report Series is to provide information to state representatives and developers of low-level radioactive waste disposal facilities about the radiological, chemical, and physical characteristics of selected radionuclides and their behavior in the waste disposal facility environment. This report also includes discussions about waste types and forms in which {sup 238}U can be found, and {sup 238}U behavior in the environment and in the human body.

Microbial activity of trench leachates from shallow-land, low-level radioactive waste disposal sites.

PubMed Central

Francis, A J; Dobbs, S; Nine, B J

1980-01-01

Trench leachate samples collected anoxically from shallow-land, low-level radioactive waste disposal sites were analyzed for total aerobic and anaerobic populations, sulfate reducers, denitrifiers, and methanogens. Among the several aerobic and anaerobic bacteria isolated, only Bacillus sp., Pseudomonas sp., Citrobacter sp., and Clostridium sp. were identified. Mixed bacterial cultures isolated from the trench leachates were able to grow anaerobically in trench leachates, which indicates that the radionuclides and organic chemicals present were not toxic to these bacteria. Changes in concentrations of several of the organic constituents of the waste leachate samples were observed due to anaerobic microbial activity. Growth of a mixed culture of trench-water bacteria in media containing a mixture of radionuclides, 60Co, 85Sr, and 134,137Cs, was not affected at total activity concentrations of 2.6 X 10(2) and 2.7 X 10(3) pCi/ml. PMID:7406490

Annual waste reduction activities report. Issue 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1991-03-18

This report discusses the waste minimization activities for the Pinellas Plant. The Pinellas Plant deals with low-level radioactive wastes, solvents, scrap metals and various other hazardous materials. This program has realized cost savings through recycling and reuse of materials.

Localized chemistry of 99Tc in simulated low activity waste glass

NASA Astrophysics Data System (ADS)

Weaver, Jamie L.

A priority of the United States Department of Energy (DOE) is to dispose of the nuclear waste accumulated in the underground tanks at the Hanford Nuclear Reservation in Richland, WA. Incorporation and stabilization of technetium (99Tc) from these tanks into vitrified waste forms is a concern to the waste glass community and DOE due to 99Tc's long half-life ( 2.13˙105 y), and its high mobility in the subsurface environment under oxidizing conditions. Working in collaboration with researchers at Pacific Northwest National Laboratory (PNNL) and other national laboratories, plans were formulated to obtain first-of-a-kind chemical structure determination of poorly understood and environmentally relevant technetium compounds that relate to the chemistry of the Tc in nuclear waste glasses. Knowledge of the structure and spectral signature of these compounds aid in refining the understanding of 99Tc incorporation into and release from oxide based waste glass. In this research a first-of-its kind mechanism for the behavior of 99Tc during vitrification is presented, and the structural role of Tc(VII) and (IV) in borosilicate waste glasses is readdressed.

Mississippi State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

The Mississippi State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state an federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Mississippi. The profile is the result of a survey of NRC licensees in Mississippi. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed throughmore » personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Mississippi.« less

Wyoming State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The Wyoming State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Wyoming. The profile is the result of a survey of NRC licensees in Wyoming. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed throughmore » personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Wyoming.« less

Massachusetts State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-03-12

The Massachusetts State Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist State and Federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Massachusetts. The profile is the result of a survey of NRC licensees in Massachusetts. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed throughmore » personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Massachusetts.« less

Texas State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-08-01

The Texas State Briefing Book is one of a series of state briefing books on low-level radioactivee waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Texas. The profile is the result of a survey of NRC licensees in Texas. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed throughmore » personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Texas.« less

Ohio State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The Ohio State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Ohio. The profile is the result of a survey of NRC licensees in Ohio. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed throughmore » personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Ohio.« less

SEPARATIONS AND WASTE FORMS CAMPAIGN IMPLEMENTATION PLAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vienna, John D.; Todd, Terry A.; Peterson, Mary E.

2012-11-26

This Separations and Waste Forms Campaign Implementation Plan provides summary level detail describing how the Campaign will achieve the objectives set-forth by the Fuel Cycle Reasearch and Development (FCRD) Program. This implementation plan will be maintained as a living document and will be updated as needed in response to changes or progress in separations and waste forms research and the FCRD Program priorities.

Glass composite waste forms for iodine confined in bismuth-embedded SBA-15

NASA Astrophysics Data System (ADS)

Yang, Jae Hwan; Park, Hwan Seo; Ahn, Do-Hee; Yim, Man-Sung

2016-11-01

The aim of this study was to stabilize bismuth-embedded SBA-15 that captured iodine gas by fabrication of monolithic waste forms. The iodine containing waste was mixed with Bi2O3 (a stabilizing additive) and low-temperature sintering glass followed by pelletizing and the sintering process to produce glass composite materials. Iodine volatility during the sintering process was significantly affected by the ratio of Bi2O3 and the glass composition. It was confirmed that BiI3, the main iodine phase within bismuth-embedded SBA-15, was effectively transformed to the mixed phases of Bi5O7I and BiOI. The initial leaching rates of iodine from the glass composite waste forms ranged 10-3-10-2 g/m2 day, showing the stability of the iodine phases encapsulated by the glassy networks. It was also observed that common groundwater anions (e.g., chloride, carbonate, sulfite, and fluoride) elevated the iodine leaching rate by anion exchange reactions. The present results suggest that the glass composite waste form of bismuth-embedded SBA-15 could be a candidate material for stable storage of 129I.

LEACHING BOUNDARY MOVEMENT IN SOLIDIFIED/STABILIZED WASTE FORMS

EPA Science Inventory

Investigation of the leaching of cement-based waste forms in acetic acid solutions found that acids attacked the waste form from the surface toward the center. A sharp leaching boundary was identified in every leached sample, using pH color indica- tors. The movement of the leach...

Bench scale experiments for the remediation of Hanford Waste Treatment Plant low activity waste melter off-gas condensate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor-Pashow, Kathryn M.L.; Poirier, Michael; McCabe, Daniel J.

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the off-gas system. The plan for disposition of this stream during baseline 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 again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter,more » so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.« less

Summary of Uranium Solubility Studies in Concrete Waste Forms and Vadose Zone Environments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Golovich, Elizabeth C.; Wellman, Dawn M.; Serne, R. Jeffrey

2011-09-30

One of the methods being considered for safely disposing of Category 3 low-level radioactive wastes is to encase the waste in concrete. Concrete encasement would contain and isolate the waste packages from the hydrologic environment and act as an intrusion barrier. The current plan for waste isolation consists of stacking low-level waste packages on a trench floor, surrounding the stacks with reinforced steel, and encasing these packages in concrete. These concrete-encased waste stacks are expected to vary in size with maximum dimensions of 6.4 m long, 2.7 m wide, and 4 m high. The waste stacks are expected to havemore » a surrounding minimum thickness of 15 cm of concrete encasement. These concrete-encased waste packages are expected to withstand environmental exposure (solar radiation, temperature variations, and precipitation) until an interim soil cover or permanent closure cover is installed and to remain largely intact thereafter. Any failure of concrete encasement may result in water intrusion and consequent mobilization of radionuclides from the waste packages. This report presents the results of investigations elucidating the uranium mineral phases controlling the long-term fate of uranium within concrete waste forms and the solubility of these phases in concrete pore waters and alkaline, circum-neutral vadose zone environments.« less

Separations and Waste Forms Research and Development FY 2013 Accomplishments Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

None, None

The Separations and Waste Form Campaign (SWFC) under the U.S. Department of Energy (DOE) Fuel Cycle Research and Development Program (FCRD) is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The fiscal year (FY) 2013 accomplishments report provides a highlight of the results of the research and development (R&D) efforts performed within SWFC in FY 2013. Each section contains a high-level overview of the activities, results, technical point of contact, applicable references, and documents produced during themore » fiscal year. This report briefly outlines campaign management and integration activities, but the intent of the report is to highlight the many technical accomplishments made during FY 2013.« less

Kansas State Briefing Book on low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-07-01

The Kansas State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Kansas. The profile is the result of a survey of radioactive material licensees in Kansas. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developedmore » through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may affect waste management practices in Kansas.« less

Vermont State Briefing Book on low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-07-01

The Vermont State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Vermont. The profile is the result of a survey of Nuclear Regulatory Commission licensees in Vermont. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment wasmore » developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may affect waste management practices in Vermont.« less

Glass binder development for a glass-bonded sodalite ceramic waste form

NASA Astrophysics Data System (ADS)

Riley, Brian J.; Vienna, John D.; Frank, Steven M.; Kroll, Jared O.; Peterson, Jacob A.; Canfield, Nathan L.; Zhu, Zihua; Zhang, Jiandong; Kruska, Karen; Schreiber, Daniel K.; Crum, Jarrod V.

2017-06-01

This paper discusses work to develop Na2O-B2O3-SiO2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. Here, five new glasses with ∼20 mass% Na2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. These improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.

Glass binder development for a glass-bonded sodalite ceramic waste form

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Vienna, John D.; Frank, Steven M.

This paper discusses work to develop Na 2O-B 2O 3-SiO 2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na 2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion formore » the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.« less

Glass binder development for a glass-bonded sodalite ceramic waste form

DOE PAGES

Riley, Brian J.; Vienna, John D.; Frank, Steven M.; ...

2017-06-01

This paper discusses work to develop Na 2O-B 2O 3-SiO 2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. In this paper, five new glasses with ~20 mass% Na 2O were designed to generate waste forms with high sodalite. The glasses were then used to produce ceramic waste forms with a surrogate salt waste. The waste forms made using these new glasses were formulated to generate more sodalite than those made with previous baseline glasses for this type of waste. The coefficients of thermal expansion formore » the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature than previous binder glasses used. Finally, these improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability.« less

Formulation and preparation of Hanford Waste Treatment Plant direct feed low activity waste Effluent Management Facility core simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCabe, Daniel J.; Nash, Charles A.; Adamson, Duane J.

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 tomore » 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 enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other problems such a recycle stream present. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate. Because this stream will recycle within WTP, these components 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 sulfate 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 formulate and prepare a simulant of the LAW

Low activation ferritic alloys

DOEpatents

Gelles, David S.; Ghoniem, Nasr M.; Powell, Roger W.

1986-01-01

Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

Low activation ferritic alloys

DOEpatents

Gelles, D.S.; Ghoniem, N.M.; Powell, R.W.

1985-02-07

Low activation ferritic alloys, specifically bainitic and martensitic stainless steels, are described for use in the production of structural components for nuclear fusion reactors. They are designed specifically to achieve low activation characteristics suitable for efficient waste disposal. The alloys essentially exclude molybdenum, nickel, nitrogen and niobium. Strength is achieved by substituting vanadium, tungsten, and/or tantalum in place of the usual molybdenum content in such alloys.

Maine State Briefing Book on low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-08-01

The Maine State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and Federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Maine. The profile is the result of a survey of radioactive material licensees in Maine. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested partices including industry, government, the media, and interest groups. The assessment was developedmore » through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant goverment agencies and activities, all of which may impact management practices in Maine.« less

Leaching boundary movement in solidified/stabilized waste forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kuang Ye Cheng; Bishop, P.L.

1992-02-01

Investigation of the leaching of cement-based waste forms in acetic acid solutions found that acids attacked the waste form from the surface toward the center. A sharp leaching boundary was identified in every leached sample, using pH color indicators. The movement of the leaching boundary was found to be a single diffusion-controlled process.

Topic I: Induced changes in hydrology at low-level radioactive waste repository sites: A section in Safe disposal of radionuclides in low-level radioactive-waste repository sites; Low-level radioactive-waste disposal workshop, U.S. Geological Survey, July 11-16, 1987, Big Bear Lake, Calif., Proceedings (Circular 1036)

USGS Publications Warehouse

Prudic, David E.; Dennehy, Kevin F.; Bedinger, Marion S.; Stevens, Peter R.

1990-01-01

Engineering practices, including the excavation of trenches, placement of waste, nature of waste forms, backfilling procedures and materials, and trench-cover construction and materials at low-level radioactive-waste repository sites greatly affect the geohydrology of the sites. Engineering practices are dominant factors in eventual stability and isolation of the waste. The papers presented relating to Topic I were discussions of the hydrogeologic setting at existing low-level radioactive-waste repository sites and changes in the hydrology induced by site operations. Papers summarizing detailed studies presented at this workshop include those at sites near Sheffield, Ill.; Oak Ridge National Laboratory, Tenn.; West Valley, N.Y.; Maxey Flats, Ky.; Barnwell, S.C.; and Beatty, Nev. 

Setting and Stiffening of Cementitious Components in Cast Stone Waste Form for Disposal of Secondary Wastes from the Hanford waste treatment and immobilization plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, Chul-Woo; Chun, Jaehun; Um, Wooyong

2013-04-01

Cast stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from Hanford vitrification plant. While the strength and radioactive technetium leaching of different waste form candidates have been reported, no study has been performed to understand the flow and stiffening behavior of Cast Stone, which is essential to ensure the proper workability, especially considering necessary safety as a nuclear waste form in a field scale application. The rheological and ultrasonic wave reflection (UWR) measurements were used to understand the setting and stiffening Cast Stone batches. X-ray diffraction (XRD) was used to find themore » correlation between specific phase formation and the stiffening of the paste. Our results showed good correlation between rheological properties of the fresh Cast Stone mixture and phase formation during hydration of Cast Stone. Secondary gypsum formation originating from blast furnace slag was observed in Cast Stone made with low concentration simulants. The formation of gypsum was suppressed in high concentration simulants. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration.« less

Forming artificial soils from waste materials for mine site rehabilitation

NASA Astrophysics Data System (ADS)

Yellishetty, Mohan; Wong, Vanessa; Taylor, Michael; Li, Johnson

2014-05-01

Surface mining activities often produce large volumes of solid wastes which invariably requires the removal of significant quantities of waste rock (overburden). As mines expand, larger volumes of waste rock need to be moved which also require extensive areas for their safe disposal and containment. The erosion of these dumps may result in landform instability, which in turn may result in exposure of contaminants such as trace metals, elevated sediment delivery in adjacent waterways, and the subsequent degradation of downstream water quality. The management of solid waste materials from industrial operations is also a key component for a sustainable economy. For example, in addition to overburden, coal mines produce large amounts of waste in the form of fly ash while sewage treatment plants require disposal of large amounts of compost. Similarly, paper mills produce large volumes of alkaline rejected wood chip waste which is usually disposed of in landfill. These materials, therefore, presents a challenge in their use, and re-use in the rehabilitation of mine sites and provides a number of opportunities for innovative waste disposal. The combination of solid wastes sourced from mines, which are frequently nutrient poor and acidic, with nutrient-rich composted material produced from sewage treatment and alkaline wood chip waste has the potential to lead to a soil suitable for mine rehabilitation and successful seed germination and plant growth. This paper presents findings from two pilot projects which investigated the potential of artificial soils to support plant growth for mine site rehabilitation. We found that pH increased in all the artificial soil mixtures and were able to support plant establishment. Plant growth was greatest in those soils with the greatest proportion of compost due to the higher nutrient content. These pot trials suggest that the use of different waste streams to form an artificial soil can potentially be used in mine site rehabilitation

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamson, Duane J.; Nash, Charles A.; McCabe, Daniel J.

2014-01-01

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 again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrificationmore » mission duration and quantity of canistered glass waste forms. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to be within acceptable concentration ranges in the LAW glass. Diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the impact of potential future disposition of this stream in the Hanford tank farms, and investigates auxiliary evaporation to enable another disposition path. Unless an auxiliary evaporator is used, returning the stream to the tank farms would require evaporation in the 242-A evaporator. This stream is expected to be unusual because it will be very high in corrosive species that are volatile in the

Radioactive Demonstration Of Mineralized Waste Forms Made From Hanford Low Activity Waste (Tank SX-105, Tank AN-103, And AZ-101/102) By Fluidized Bed Steam Reformation (FBSR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jantzen, C. M.; Crawford, C. L.; Bannochie, C. J.

Fluidized Bed Steam Reforming (FBSR) is a robust technology for the immobilization of a wide variety of radioactive wastes. Applications have been tested at the pilot scale for the high sodium, sulfate, halide, organic and nitrate wastes at the Hanford site, the Idaho National Laboratory (INL), and the Savannah River Site (SRS). Due to the moderate processing temperatures, halides, sulfates, and technetium are retained in mineral phases of the feldspathoid family (nepheline, sodalite, nosean, carnegieite, etc). The feldspathoid minerals bind the contaminants such as Tc-99 in cage (sodalite, nosean) or ring (nepheline) structures to surrounding aluminosilicate tetrahedra in the feldspathoidmore » structures. The granular FBSR mineral waste form that is produced has a comparable durability to LAW glass based on the short term PCT testing in this study, the INL studies, SPFT and PUF testing from previous studies as given in the columns in Table 1-3 that represent the various durability tests. Monolithing of the granular product was shown to be feasible in a separate study. Macro-encapsulating the granular product provides a decrease in leaching compared to the FBSR granular product when the geopolymer is correctly formulated.« less

Material Recovery and Waste Form Development FY 2015 Accomplishments Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todd, Terry Allen; Braase, Lori Ann

The Material Recovery and Waste Form Development (MRWFD) Campaign under the U.S. Department of Energy (DOE) Fuel Cycle Technologies (FCT) Program is responsible for developing advanced separation and waste form technologies to support the various fuel cycle options defined in the DOE Nuclear Energy Research and Development Roadmap, Report to Congress, April 2010. The FY 2015 Accomplishments Report provides a highlight of the results of the research and development (R&D) efforts performed within the MRWFD Campaign in FY-14. Each section contains a high-level overview of the activities, results, technical point of contact, applicable references, and documents produced during the fiscalmore » year. This report briefly outlines campaign management and integration activities, but primarily focuses on the many technical accomplishments made during FY-15. The campaign continued to utilize an engineering driven-science-based approach to maintain relevance and focus. There was increased emphasis on development of technologies that support near-term applications that are relevant to the current once-through fuel cycle.« less

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, task 17208: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

Melt processed crystalline ceramic waste forms for advanced nuclear fuel cycles: CRP T21027 1813: Processing technologies for high level waste, formulation of matrices and characterization of waste forms, Task 17208: Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amoroso, J. W.; Marra, J. C.

2015-08-26

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics)more » over geologic timescales of interest for nuclear waste immobilization [1]. A durable multi-phase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing. This report summarizes results from three years of work on the IAEA Coordinated Research Project on “Processing technologies for high level waste, formulation of matrices and characterization of waste forms” (T21027), and specific task “Melt Processed Crystalline Ceramic Waste Forms for Advanced Nuclear Fuel Cycles” (17208).« less

New Jersey State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The New Jersey state Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in New Jersey. The profile is the result of a survey of NRC licensees in New Jersey. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessmentmore » was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in New Jersey.« less

North Carolina State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The North Carolina State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in North Carolina. The profile is the result of a survey of NRC licensees in North Carolina. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessmentmore » was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in North Carolina.« less

Puerto Rico State Briefing Book for low-level radioactive waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

The Puerto Rico State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Puerto Rico. The profile is the result of a survey of NRC licensees in Puerto Rico. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessmentmore » was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Puerto Rico.« less

EXAFS/XANES studies of plutonium-loaded sodalite/glass waste forms

NASA Astrophysics Data System (ADS)

Richmann, Michael K.; Reed, Donald T.; Kropf, A. Jeremy; Aase, Scott B.; Lewis, Michele A.

2001-09-01

A sodalite/glass ceramic waste form is being developed to immobilize highly radioactive nuclear wastes in chloride form, as part of an electrochemical cleanup process. Two types of simulated waste forms were studied: where the plutonium was alone in an LiCl/KCl matrix and where simulated fission-product elements were added representative of the electrometallurgical treatment process used to recover uranium from spent nuclear fuel also containing plutonium and a variety of fission products. Extended X-ray absorption fine structure spectroscopy (EXAFS) and X-ray absorption near-edge spectroscopy (XANES) studies were performed to determine the location, oxidation state, and particle size of the plutonium within these waste form samples. Plutonium was found to segregate as plutonium(IV) oxide with a crystallite size of at least 4.8 nm in the non-fission-element case and 1.3 nm with fission elements present. No plutonium was observed within the sodalite in the waste form made from the plutonium-loaded LiCl/KCl eutectic salt. Up to 35% of the plutonium in the waste form made from the plutonium-loaded simulated fission-product salt may be segregated with a heavy-element nearest neighbor other than plutonium or occluded internally within the sodalite lattice.

TWRS Retrieval and Storage Mission and Immobilized Low Activity Waste (ILAW) Disposal Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

BURBANK, D.A.

This project plan has a twofold purpose. First, it provides a waste stream project plan specific to the River Protection Project (RPP) (formerly the Tank Waste Remediation System [TWRS] Project) Immobilized Low-Activity Waste (LAW) Disposal Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-90-01 (Ecology et al. 1994) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan (Ecology et al. 1998). Second, it provides an upper tier document that can be used as themore » basis for future subproject line-item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 [DOE 1992] and 430.1 [DOE 1995a]). The format and content of this project plan are designed to accommodate the requirements mentioned by the Tri-Party Agreement and the DOE orders. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.« less

RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Burket, P.; Cozzi, A.

2012-02-02

The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in themore » time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as {sup 137}Cs, {sup 129}I, {sup 99}Tc, Cl, F, and SO{sub 4} that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline

Three-dimensional mapping of crystalline ceramic waste form materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cocco, Alex P.; DeGostin, Matthew B.; Wrubel, Jacob A.

Here, we demonstrate the use of synchrotron-based, transmission X-ray microscopy (TXM) and scanning electron microscopy to image the 3-D morphologies and spatial distributions of Ga-doped phases within model, single- and two-phase waste form material systems. Gallium doping levels consistent with those commonly used for nuclear waste immobilization (e.g., Ba 1.04Cs 0.24Ga 2.32Ti 5.68O 16) could be readily imaged. This analysis suggests that a minority phase with different stoichiometry/composition from the primary hollandite phase can be formed by the solid-state ceramic processing route with varying morphology (globular vs. cylindrical) as a function of Cs content. Our results represent a crucial stepmore » in developing the tools necessary to gain an improved understanding of the microstructural and chemical properties of waste form materials that influence their resistance to aqueous corrosion. This understanding will aid in the future design of higher durability waste form materials.« less

Three-dimensional mapping of crystalline ceramic waste form materials

DOE PAGES

Cocco, Alex P.; DeGostin, Matthew B.; Wrubel, Jacob A.; ...

2017-04-21

Here, we demonstrate the use of synchrotron-based, transmission X-ray microscopy (TXM) and scanning electron microscopy to image the 3-D morphologies and spatial distributions of Ga-doped phases within model, single- and two-phase waste form material systems. Gallium doping levels consistent with those commonly used for nuclear waste immobilization (e.g., Ba 1.04Cs 0.24Ga 2.32Ti 5.68O 16) could be readily imaged. This analysis suggests that a minority phase with different stoichiometry/composition from the primary hollandite phase can be formed by the solid-state ceramic processing route with varying morphology (globular vs. cylindrical) as a function of Cs content. Our results represent a crucial stepmore » in developing the tools necessary to gain an improved understanding of the microstructural and chemical properties of waste form materials that influence their resistance to aqueous corrosion. This understanding will aid in the future design of higher durability waste form materials.« less

78 FR 1155 - Low-Level Waste Disposal

Federal Register 2010, 2011, 2012, 2013, 2014

2013-01-08

... NUCLEAR REGULATORY COMMISSION 10 CFR Part 61 [NRC-2011-0012] RIN 3150-AI92 Low-Level Waste... correcting a document appearing in the Federal Register on December 7, 2012 entitled, ``Low-Level Waste... and Submitting Comments, ``Regulatory Analysis for Proposed Revisions to Low-Level Waste Disposal...

Low Activity Waste Pretreatment System Bench-Scale Testing: Supporting Integrated Testing and Facility Safety Analyses - 17171

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schonewill, Philip P.; Russell, Renee L.; Daniel, Richard C.

The Low Activity Waste Pretreatment System (LAWPS) is being designed to enable the direct feed of waste to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) facility to be immobilized. Prior to construction of the LAWPS, pilot-scale integrated testing of the key unit operations (crossflow filtration, ion exchange using spherical resorcinol-formaldehyde (sRF) resin) will be conducted by a team led by Washington River Protection Solutions (WRPS) to increase the technology maturation level of the facility’s critical technology elements. As a part of this effort, Pacific Northwest National Laboratory (PNNL) has conducted a series of bench-scalemore » (or engineering-scale) tests to perform two major objectives: (1) support pilot-scale integrated testing of the LAWPS by supplying information or performance data in advance of operating the pilot-scale facility; and (2) collect data needed to establish or confirm assumptions/approaches planned for implementation in the LAWPS safety basis. The first objective was focused in two technical areas: developing simulants that are representative of expected waste feed and can be produced at larger scales, and using these simulants in a bench-scale crossflow filter to establish expected solid-liquid separation performance. The crossflow filter was also used to observe the efficacy (with respect to filter production rate) of selected operational strategies. The second objective also included two technical areas: measuring the effect of sRF resin on hydrogen generation rate under irradiation, and demonstrating that the planned hydrogen management approach is effective and robust. The hydrogen management strategy involves fluidization of the sRF resin bed in the ion exchange columns and recirculating the liquid, a scenario that is planned for testing at full column height. The full height tests at PNNL also supported full-scale IX column testing conducted as part of the technology maturation

Glass-bonded iodosodalite waste form for immobilization of 129I

NASA Astrophysics Data System (ADS)

Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.; Tabada, Diana; Wall, Donald; Corkhill, Claire L.; McCloy, John S.

2018-06-01

Immobilization of radioiodine is an important requirement for current and future nuclear fuel cycles. Iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Dried unwashed sodalite powders were used to synthesize glass-bonded iodosodalite waste forms (glass composite materials) by heating pressed pellets at 650, 750, or 850 °C with two types of sodium borosilicate glass binders. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. For the best waste form produced (pellets mixed with 10 mass% of glass binder and heat-treated at 750 °C), the maximum possible elemental iodine loading was 19.8 mass%, but only ∼8-9 mass% waste loading of iodine was retained in the waste form after thermal processing. Other pellets with higher iodine retention either contained higher porosity or were incompletely sintered. ASTM C1308 and C1285 (product consistency test, PCT) experiments were performed to understand chemical durability under diffusive and static conditions. The C1308 test resulted in significantly higher normalized loss compared to the C1285 test, most likely because of the strong effect of neutral pH solution renewal and prevention of ion saturation in solution. Both experiments indicated that release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix or from initial sodalite synthesis; however the C1308 test result indicated that congruent dissolution of iodosodalite occurred. The average release rates of iodine obtained from C1308 were 0.17 and 1.29 g m-2 d-1 for 80 or 8 m-1, respectively, and the C1285 analysis gave a value of 2 × 10-5 g m-2 d-1, which is comparable to or better than the durability of

40 CFR 761.345 - Form of the waste to be sampled.

Code of Federal Regulations, 2013 CFR

2013-07-01

... Characterization for PCB Disposal in Accordance With § 761.62, and Sampling PCB Remediation Waste Destined for Off-Site Disposal, in Accordance With § 761.61 § 761.345 Form of the waste to be sampled. PCB bulk product waste and PCB remediation waste destined for off-site disposal must be in the form of either flattened...

Supplemental Immobilization Cast Stone Technology Development and Waste Form Qualification Testing Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westsik, Joseph H.; Serne, R. Jeffrey; Pierce, Eric M.

2013-05-31

The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). The pretreatment facility will have the capacity to separate all of the tank wastes into the HLW and LAW fractions, and the HLW Vitrification Facility will have the capacity to vitrifymore » all of the HLW. However, a second immobilization facility will be needed for the expected volume of LAW requiring immobilization. A number of alternatives, including Cast Stone—a cementitious waste form—are being considered to provide the additional LAW immobilization capacity.« less

78 FR 65390 - Exemption From Licensing for Disposal of Low-Activity Radioactive Waste at the US Ecology Idaho...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-31

... Disposal of Low-Activity Radioactive Waste at the US Ecology Idaho Resource Conservation and Recovery Act..., Pennsylvania, at the US Ecology Idaho (USEI) Resource Conservation and Recovery Act (RCRA) Subtitle C disposal... from the US Ecology, Inc. (US Ecology), dated July 7, 2013 (ADAMS Accession No. ML13198A017), for...

Glass binder development for a glass-bonded sodalite ceramic waste form

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Vienna, John D.; Frank, Steven M.

This paper discusses work to develop Na2O-B2O3-SiO2 glass binders for immobilizing LiCl-KCl eutectic salt waste in a glass-bonded sodalite waste form following electrochemical reprocessing of used metallic nuclear fuel. Here, five new glasses with high Na2O contents were designed to generate waste forms having higher sodalite contents and fewer stress fractures. The structural, mechanical, and thermal properties of the new glasses were measured using variety of analytical techniques. The glasses were then used to produce ceramic waste forms with surrogate salt waste. The materials made using the glasses developed during this study were formulated to generate more sodalite than materialsmore » made with previous baseline glasses used. The coefficients of thermal expansion for the glass phase in the glass-bonded sodalite waste forms made with the new binder glasses were closer to the sodalite phase in the critical temperature region near and below the glass transition temperature. These improvements should result in lower probability of cracking in the full-scale monolithic ceramic waste form, leading to better long-term chemical durability. Additionally, a model generated during this study for predicting softening temperature of silicate binder glasses is presented.« less

Recharge Data Package for the Immobilized Low-Activity Waste 2001 Performance Assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

MJ Fayer; EM Murphy; JL Downs

2000-01-18

Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are currently stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as immobilized low-activity waste (ILAW) is to vitrify the waste and place the product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is known as the Hanford ILAW Performance Assessment (PA) Activity, hereafter called the ILAW PA project. The goal of this project is to provide a reasonable expectation thatmore » the disposal of the waste is protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require predictions of contaminant migration from the facility. To make such predictions will require estimates of the fluxes of water moving through the sediments within the vadose zone around and beneath the disposal facility. These fluxes, loosely called recharge rates, are the primary mechanism for transporting contaminants to the groundwater. Pacific Northwest National Laboratory (PNNL) assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of recharge rates for current conditions and long-term scenarios involving the shallow-land disposal of ILAW. Specifically, recharge estimates are needed for a filly functional surface cover; the cover sideslope, and the immediately surrounding terrain. In addition, recharge estimates are needed for degraded cover conditions. The temporal scope of the analysis is 10,000 years, but could be longer if some contaminant peaks occur after 10,000 years. The elements of this report compose the Recharge Data Package, which provides estimates of recharge rates for the scenarios being considered in the 2001 PA. Table S.1 identifies the surface

77 FR 72997 - Low-Level Waste Disposal

Federal Register 2010, 2011, 2012, 2013, 2014

2012-12-07

...-2011-0012] RIN 3150-AI92 Low-Level Waste Disposal AGENCY: Nuclear Regulatory Commission. ACTION... Regulatory Commission (NRC) is proposing to amend its regulations that govern low-level radioactive waste... development of criteria for waste acceptance based on the results of these analyses. These amendments will...

Nuclear waste forms for actinides

PubMed Central

Ewing, Rodney C.

1999-01-01

The disposition of actinides, most recently 239Pu from dismantled nuclear weapons, requires effective containment of waste generated by the nuclear fuel cycle. Because actinides (e.g., 239Pu and 237Np) are long-lived, they have a major impact on risk assessments of geologic repositories. Thus, demonstrable, long-term chemical and mechanical durability are essential properties of waste forms for the immobilization of actinides. Mineralogic and geologic studies provide excellent candidate phases for immobilization and a unique database that cannot be duplicated by a purely materials science approach. The “mineralogic approach” is illustrated by a discussion of zircon as a phase for the immobilization of excess weapons plutonium. PMID:10097054

National low-level waste management program radionuclide report series, Volume 14: Americium-241

DOE Office of Scientific and Technical Information (OSTI.GOV)

Winberg, M.R.; Garcia, R.S.

1995-09-01

This report, Volume 14 of the National Low-Level Waste Management Program Radionuclide Report Series, discusses the radiological and chemical characteristics of americium-241 ({sup 241}Am). This report also includes discussions about waste types and forms in which {sup 241}Am can be found and {sup 241}Am behavior in the environment and in the human body.

Measurements of Mercury Released from Solidified/Stabilized Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mattus, C.H.

2001-04-19

This report covers work performed during FY 1999-2000 in support of treatment demonstrations conducted for the Mercury Working Group of the U.S. Department of Energy (DOE) Mixed Waste Focus Area. In order to comply with the requirements of the Resource Conservation and Recovery Act, as implemented by the U.S. Environmental Protection Agency (EPA), DOE must use one of these procedures for wastes containing mercury at levels above 260 ppm: a retorting/roasting treatment or an incineration treatment (if the wastes also contain organics). The recovered radioactively contaminated mercury must then be treated by an amalgamation process prior to disposal. The DOEmore » Mixed Waste Focus Area and Mercury Working Group are working with the EPA to determine if some alternative processes could treat these types of waste directly, thereby avoiding for DOE the costly recovery step. They sponsored a demonstration in which commercial vendors applied their technologies for the treatment of two contaminated waste soils from Brookhaven National Laboratory. Each soil was contaminated with {approx}4500 ppm mercury; however, one soil had as a major radioelement americium-241, while the other contained mostly europium-152. The project described in this report addressed the need for data on the mercury vapor released by the solidified/stabilized mixed low-level mercury wastes generated during these demonstrations as well as the comparison between the untreated and treated soils. A related work began in FY 1998, with the measurement of the mercury released by amalgamated mercury, and the results were reported in ORNL/TM-13728. Four treatments were performed on these soils. The baseline was obtained by thermal treatment performed by SepraDyne Corp., and three forms of solidification/stabilization were employed: one using sulfur polymer cement (Brookhaven National Laboratory), one using portland cement [Allied Technology Group (ATG)], and a third using proprietary additives (Nuclear Fuel

Chemical composition measurements of the low activity waste (LAW) EPA-Series glasses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fox, K.; Edwards, T. B.

2016-03-01

In this report, the Savannah River National Laboratory provides chemical analysis results for a series of simulated low activity waste glasses provided by Pacific Northwest National Laboratory as part of an ongoing development task. The measured chemical composition data are reported and compared with the targeted values for each component for each glass. A detailed review showed no indications of errors in the preparation or measurement of the study glasses. All of the measured sums of oxides for the study glasses fell within the interval of 100.2 to 100.8 wt %, indicating recovery of all components. Comparisons of the targetedmore » and measured chemical compositions showed that the measured values for the glasses met the targeted concentrations within 10% for those components present at more than 5 wt %.« less

Special Analysis for Disposal of High-Concentration I-129 Waste in the Intermediate-Level Vaults at the E-Area Low-Level Waste Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collard, L.B.

2000-09-26

This revision was prepared to address comments from DOE-SR that arose following publication of revision 0. This Special Analysis (SA) addresses disposal of wastes with high concentrations of I-129 in the Intermediate-Level (IL) Vaults at the operating, low-level radioactive waste disposal facility (the E-Area Low-Level Waste Facility or LLWF) on the Savannah River Site (SRS). This SA provides limits for disposal in the IL Vaults of high-concentration I-129 wastes, including activated carbon beds from the Effluent Treatment Facility (ETF), based on their measured, waste-specific Kds.

National profile on commercially generated low-level radioactive mixed waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Klein, J.A.; Mrochek, J.E.; Jolley, R.L.

1992-12-01

This report details the findings and conclusions drawn from a survey undertaken as part of a joint US Nuclear Regulatory Commission and US Environmental Protection Agency-sponsored project entitled ``National Profile on Commercially Generated Low-Level Radioactive Mixed Waste.`` The overall objective of the work was to compile a national profile on the volumes, characteristics, and treatability of commercially generated low-level mixed waste for 1990 by five major facility categories-academic, industrial, medical, and NRC-/Agreement State-licensed goverment facilities and nuclear utilities. Included in this report are descriptions of the methodology used to collect and collate the data, the procedures used to estimate themore » mixed waste generation rate for commercial facilities in the United States in 1990, and the identification of available treatment technologies to meet applicable EPA treatment standards (40 CFR Part 268) and, if possible, to render the hazardous component of specific mixed waste streams nonhazardous. The report also contains information on existing and potential commercial waste treatment facilities that may provide treatment for specific waste streams identified in the national survey. The report does not include any aspect of the Department of Energy`s (DOES) management of mixed waste and generally does not address wastes from remedial action activities.« less

Iodine Solubility in Low-Activity Waste Borosilicate Glass at 1000 °C

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Schweiger, Michael J.; Kim, Dong-Sang

2014-04-30

The purpose of this study was to determine the solubility of iodine in a low-activity waste borosilicate glass when heated inside an evacuated and sealed fused quartz ampoule. The iodine was added to glass frit as KI in quantities of 100–24000 ppm iodine (by mass), each mixture was added to an ampoule, the ampoule was heated at 1000 °C for 2 h and then air quenched. In samples with ≥12000 ppm iodine, low viscosity salt phases were observed on the surface of the melts during cooling that solidified into a white coating upon cooling. These salts were identified as mixturesmore » of KI, NaI, and Na2SO4 with X-ray diffraction (XRD). The iodine concentrations in glass specimens were analyzed with inductively-coupled plasma mass spectrometry and the overall iodine solubility was determined to be 10000 ppm by mass. Several crystalline inclusions of iodine sodalite, Na8(AlSiO4)6I2, were observed in the 24000 ppm specimen and were verified with micro-XRD and wavelength dispersive spectroscopy.« less

Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

NASA Astrophysics Data System (ADS)

Williams, Benjamin D.; Neeway, James J.; Snyder, Michelle M. V.; Bowden, Mark E.; Amonette, James E.; Arey, Bruce W.; Pierce, Eric M.; Brown, Christopher F.; Qafoku, Nikolla P.

2016-05-01

Mitigation of hazardous and radioactive waste can be improved through conversion of existing waste to a more chemically stable and physically robust waste form. One option for waste conversion is the fluidized bed steam reforming (FBSR) process. The resulting FBSR granular material was encapsulated in a geopolymer matrix referred to here as Geo-7. This provides mechanical strength for ease in transport and disposal. However, it is necessary to understand the phase assemblage evolution as a result of geopolymer encapsulation. In this study, we examine the mineral assemblages formed during the synthesis of the multiphase ceramic waste form. The FBSR granular samples were created from waste simulant that was chemically adjusted to resemble Hanford tank waste. Another set of samples was created using Savannah River Site Tank 50 waste simulant in order to mimic a blend of waste collected from 68 Hanford tank. Waste form performance tests were conducted using the product consistency test (PCT), the Toxicity Characteristic Leaching Procedure (TCLP), and the single-pass flow-through (SPFT) test. X-ray diffraction analyses revealed the structure of a previously unreported NAS phase and indicate that monolith creation may lead to a reduction in crystallinity as compared to the primary FBSR granular product.

Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

DOE PAGES

Williams, Benjamin D.; Neeway, James J.; Snyder, Michelle M. V.; ...

2015-12-23

We can improve mitigation of hazardous and radioactive waste through conversion of existing waste to a more chemically stable and physically robust waste form. One option for waste conversion is the fluidized bed steam reforming (FBSR) process. The resulting FBSR granular material was encapsulated in a geopolymer matrix referred to here as Geo-7. This provides mechanical strength for ease in transport and disposal. But, it is necessary to understand the phase assemblage evolution as a result of geopolymer encapsulation. In this study, we examine the mineral assemblages formed during the synthesis of the multiphase ceramic waste form. The FBSR granularmore » samples were created from waste simulant that was chemically adjusted to resemble Hanford tank waste. Another set of samples was created using Savannah River Site Tank 50 waste simulant in order to mimic a blend of waste collected from 68 Hanford tank. Waste form performance tests were conducted using the product consistency test (PCT), the Toxicity Characteristic Leaching Procedure (TCLP), and the single-pass flow-through (SPFT) test. Finally, X-ray diffraction analyses revealed the structure of a previously unreported NAS phase and indicate that monolith creation may lead to a reduction in crystallinity as compared to the primary FBSR granular product.« less

Final waste forms project: Performance criteria for phase I treatability studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilliam, T.M.; Hutchins, D.A.; Chodak, P. III

1994-06-01

This document defines the product performance criteria to be used in Phase I of the Final Waste Forms Project. In Phase I, treatability studies will be performed to provide {open_quotes}proof-of-principle{close_quotes} data to establish the viability of stabilization/solidification (S/S) technologies. This information is required by March 1995. In Phase II, further treatability studies, some at the pilot scale, will be performed to provide sufficient data to allow treatment alternatives identified in Phase I to be more fully developed and evaluated, as well as to reduce performance uncertainties for those methods chosen to treat a specific waste. Three main factors influence themore » development and selection of an optimum waste form formulation and hence affect selection of performance criteria. These factors are regulatory, process-specific, and site-specific waste form standards or requirements. Clearly, the optimum waste form formulation will require consideration of performance criteria constraints from each of the three categories. Phase I will focus only on the regulatory criteria. These criteria may be considered the minimum criteria for an acceptable waste form. In other words, a S/S technology is considered viable only if it meet applicable regulatory criteria. The criteria to be utilized in the Phase I treatability studies were primarily taken from Environmental Protection Agency regulations addressed in 40 CFR 260 through 265 and 268; and Nuclear Regulatory Commission regulations addressed in 10 CFR 61. Thus the majority of the identified criteria are independent of waste form matrix composition (i.e., applicable to cement, glass, organic binders etc.).« less

Colloid formation during waste form reaction: Implications for nuclear waste disposal

USGS Publications Warehouse

Bates, J. K.; Bradley, J.; Teetsov, A.; Bradley, C. R.; Buchholtz ten Brink, Marilyn R.

1992-01-01

Insoluble plutonium- and americium-bearing colloidal particles formed during simulated weathering of a high-level nuclear waste glass. Nearly 100 percent of the total plutonium and americium in test ground water was concentrated in these submicrometer particles. These results indicate that models of actinide mobility and repository integrity, which assume complete solubility of actinides in ground water, underestimate the potential for radionuclide release into the environment. A colloid-trapping mechanism may be necessary for a waste repository to meet long-term performance specifications.

Fundamental Aspects of Zeolite Waste Form Production by Hot Isostatic Pressing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jubin, Robert Thomas; Bruffey, Stephanie H.; Jordan, Jacob A.

The direct conversion of iodine-bearing sorbents into a stable waste form is a research topic of interest to the US Department of Energy. The removal of volatile radioactive 129I from the off-gas of a nuclear fuel reprocessing facility will be necessary in order to comply with the regulatory requirements that apply to facilities sited within the United States (Jubin et al., 2012a), and any iodine-containing media or solid sorbents generated by this process would contain 129I and would be destined for eventual geological disposal. While recovery of iodine from some sorbents is possible, a method to directly convert iodineloaded sorbentsmore » to a durable waste form with little or no additional waste materials being formed and a potentially reduced volume would be beneficial. To this end, recent studies have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by hot isostatic pressing (HIPing) (Bruffey and Jubin, 2015). Silver mordenite (AgZ), of the zeolite class of minerals, is under consideration for use in adsorbing iodine from nuclear reprocessing off-gas streams. Direct conversion of I-AgZ by HIPing may provide the following benefits: (1) a waste form of high density that is tolerant to high temperatures, (2) a waste form that is not significantly chemically hazardous, and (3) a robust conversion process that requires no pretreatment.« less

Technical and economic evaluation of controlled disposal options for very low level radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robinson, P.J.; Vance, J.N.

1990-08-01

Over the past several years, there has been considerable interest by the nuclear industry in the Nuclear Regulatory Commission (NRC) explicitly defined an activity level in plant waste materials at which the radiological impacts would be so low as to be considered Below Regulatory Concern (BRC). In January 1989, Electric Power Research Institute (EPRI) completed an extensive industry research effort to develop the technical bases for establishing criteria for the disposal of very low activity wastes in ordinary disposal facilities. The Nuclear Management and Resources Council (NUMARC), with assistance from the Edison Electric Institute (EEI) and the Electric Power Researchmore » Institute (EPRI), drafted a petition titled: Petition for Rulemaking Regarding Disposal of Below Regulatory Concern Radioactive Wastes from Commercial Nuclear Power Plants.'' Subsequent to the industry making a final decision for submittal of the drafted BRC petition, EPRI was requested to evaluate the technical and economic impact of six BRC options. These options are: take no action in pursuing a BRC waste exemption, petition the NRC for authorization to disposal of any BRC waste in any ordinary disposal facility, limit disposal of BRC waste to the nuclear power plant site, limit disposal of BRC waste to the nuclear power plant site and other utility owned property, petition for a mixed waste exemption, and petition for single waste stream exemptions in sequence (i.e. soil, followed by sewage sludge, etc.). The petition and technical bases were written to support the disposal of any BRC waste type in any ordinary disposal facility. These documents do not provide all of the technical and economic information needed to completely assessment the BRC options. This report provides the technical and economic basis for a range of options concerning disposal of very low activity wastes. 3 figs., 20 tabs.« less

Geochemical data package for the Hanford immobilized low-activity tank waste performance assessment (ILAW PA)

DOE Office of Scientific and Technical Information (OSTI.GOV)

DI Kaplan; RJ Serne

Lockheed Martin Hanford Company (LMHC) is designing and assessing the performance of disposal facilities to receive radioactive wastes that are stored in single- and double-shell tanks at the Hanford Site. The preferred method of disposing of the portion that is classified as low-activity waste is to vitrify the liquid/slurry and place the solid product in near-surface, shallow-land burial facilities. The LMHC project to assess the performance of these disposal facilities is the Hanford Immobilized Low-Activity Tank Waste (ILAW) Performance Assessment (PA) activity. The goal of this project is to provide a reasonable expectation that the disposal of the waste ismore » protective of the general public, groundwater resources, air resources, surface-water resources, and inadvertent intruders. Achieving this goal will require prediction of contaminant migration from the facilities. This migration is expected to occur primarily via the movement of water through the facilities, and the consequent transport of dissolved contaminants in the porewater of the vadose zone. Pacific Northwest National Laboratory assists LMHC in their performance assessment activities. One of the PNNL tasks is to provide estimates of the geochemical properties of the materials comprising the disposal facility, the disturbed region around the facility, and the physically undisturbed sediments below the facility (including the vadose zone sediments and the aquifer sediments in the upper unconfined aquifer). The geochemical properties are expressed as parameters that quantify the adsorption of contaminants and the solubility constraints that might apply for those contaminants that may exceed solubility constraints. The common parameters used to quantify adsorption and solubility are the distribution coefficient (K{sub d}) and the thermodynamic solubility product (K{sub sp}), respectively. In this data package, the authors approximate the solubility of contaminants using a more simplified

Controlling mechanisms of metals release form cement-based waste form in acetic acid solution

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, Kuang Ye.

1991-01-01

The purpose of this dissertation is to identify the individual leaching mechanisms of metals by knowing the pH profile within the leached specimen and the physical and chemical properties of the leached material. Leaching of cement-based waste form in acetic acid solutions with different acidic strengths has been investigated in this work. The pH profile along the acid penetration route in the cement-based waste form was identified by various pH color indicators. The pH in the surface altered layer varies from 5.0 to 6.0, which is very close to the pH in the bulk leachate. A reacting zone, where themore » pH abruptly changes from 6 to 12, sharply divides the altered surface layer from the remaining unleached waste form or kernel. Leaching of metals is controlled by the acidity available in the leachant. Dissolution of alkaline materials leaves a silica-rich layer on the surface of the cement-based waste form. This surface layer exhibits different properties than those of the unleached material. The surface layer has a higher water content, is lighter weight, and is soft and friable. Furthermore, the abundant silicate content on the solid surface detains portion of the leached metals, while they are moving through the leached layer into bulk solution. The leaching of metals is a consequence of acid penetration. The distance from the solid/solution interface to the front of the leaching boundary can be regarded as the depth of leaching zone, where the metals dissolve and diffuse out of the waste form. The metal ions diffuse through the leached layer may be retarded on the solid surface by the pH-dependent adsorption reactions. It is found that the leaching process through the leached layer is diffusion-controlled for calcium and cadmium, whereas diffusion and adsorption occur simultaneously in the leached layer for lead and arsenic.« less

Investigation of variable compositions on the removal of technetium from Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor-Pashow, Kathryn M. L.; McCabe, Daniel J.; Pareizs, John M.

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the offgas system. The plan for disposition of this stream during baseline 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 again. The primary reason to recycle this stream is so that the semi-volatile 99Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter,more » so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the 99Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.« less

A new approach to characterize very-low-level radioactive waste produced at hadron accelerators.

PubMed

Zaffora, Biagio; Magistris, Matteo; Chevalier, Jean-Pierre; Luccioni, Catherine; Saporta, Gilbert; Ulrici, Luisa

2017-04-01

Radioactive waste is produced as a consequence of preventive and corrective maintenance during the operation of high-energy particle accelerators or associated dismantling campaigns. Their radiological characterization must be performed to ensure an appropriate disposal in the disposal facilities. The radiological characterization of waste includes the establishment of the list of produced radionuclides, called "radionuclide inventory", and the estimation of their activity. The present paper describes the process adopted at CERN to characterize very-low-level radioactive waste with a focus on activated metals. The characterization method consists of measuring and estimating the activity of produced radionuclides either by experimental methods or statistical and numerical approaches. We adapted the so-called Scaling Factor (SF) and Correlation Factor (CF) techniques to the needs of hadron accelerators, and applied them to very-low-level metallic waste produced at CERN. For each type of metal we calculated the radionuclide inventory and identified the radionuclides that most contribute to hazard factors. The methodology proposed is of general validity, can be extended to other activated materials and can be used for the characterization of waste produced in particle accelerators and research centres, where the activation mechanisms are comparable to the ones occurring at CERN. Copyright © 2017 Elsevier Ltd. All rights reserved.

Active and passive computed tomography mixed waste focus area final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roberson, G P

1998-08-19

The Mixed Waste Focus Area (MWFA) Characterization Development Strategy delineates an approach to resolve technology deficiencies associated with the characterization of mixed wastes. The intent of this strategy is to ensure the availability of technologies to support the Department of Energy's (DOE) mixed waste low-level or transuranic (TRU) contaminated waste characterization management needs. To this end the MWFA has defined and coordinated characterization development programs to ensure that data and test results necessary to evaluate the utility of non-destructive assay technologies are available to meet site contact handled waste management schedules. Requirements used as technology development project benchmarks are basedmore » in the National TRU Program Quality Assurance Program Plan. These requirements include the ability to determine total bias and total measurement uncertainty. These parameters must be completely evaluated for waste types to be processed through a given nondestructive waste assay system constituting the foundation of activities undertaken in technology development projects. Once development and testing activities have been completed, Innovative Technology Summary Reports are generated to provide results and conclusions to support EM-30, -40, or -60 end user/customer technology selection. The Active and Passive Computed Tomography non-destructive assay system is one of the technologies selected for development by the MWFA. Lawrence Livermore National Laboratory's (LLNL) is developing the Active and Passive Computed Tomography (A&PCT) nondestructive assay (NDA) technology to identify and accurately quantify all detectable radioisotopes in closed containers of waste. This technology will be applicable to all types of waste regardless of .their classification; low level, transuranic or provide results and conclusions to support EM-30, -40, or -60 end user/customer technology selection. The Active and Passive Computed Tomography non

NEVADA TEST SITE WASTE ACCEPTANCE CRITERIA

DOE Office of Scientific and Technical Information (OSTI.GOV)

U.S. DEPARTMENT OF ENERGY, NATIONAL NUCLEAR SECURITY ADMINISTRATION, NEVADA SITE OFFICE

This document establishes the U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site will accept low-level radioactive and mixed waste for disposal. Mixed waste generated within the State of Nevada by NNSA/NSO activities is accepted for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the Nevada Test Site Area 3 and Area 5 Radioactive Waste Management Site for storage or disposal.

Secondary Waste Form Development and Optimization—Cast Stone

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sundaram, S. K.; Parker, Kent E.; Valenta, Michelle M.

2011-07-14

Washington River Protection Services is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF is a Resource Conservation and Recovery Act-permitted, multi-waste, treatment and storage unit and can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needs to be operational by 2018 to receive secondary liquid wastes generated during operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The STU to ETF will provide the additional capacity needed for ETF to process the increased volume of secondary wastes expected to be produced by WTP.

Test plan for formulation and evaluation of grouted waste forms with shine process wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ebert, W. L.; Jerden, J. L.

2015-09-01

The objective of this experimental project is to demonstrate that waste streams generated during the production of Mo99 by the SHINE Medical Technologies (SHINE) process can be immobilized in cement-based grouted waste forms having physical, chemical, and radiological stabilities that meet regulatory requirements for handling, storage, transport, and disposal.

Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tyacke, M.

1993-08-01

This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placedmore » in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.« less

77 FR 64361 - Report on Waste Burial Charges: Changes in Decommissioning Waste Disposal Costs at Low-Level...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-19

... Decommissioning Waste Disposal Costs at Low-Level Waste Burial Facilities AGENCY: Nuclear Regulatory Commission... 15, ``Report on Waste Burial Charges: Changes in Decommissioning Waste Disposal Costs at Low-Level... for low-level waste. DATES: Submit comments by November 15, 2012. Comments received after this date...

Glasses for immobilization of low- and intermediate-level radioactive waste

NASA Astrophysics Data System (ADS)

Laverov, N. P.; Omel'yanenko, B. I.; Yudintsev, S. V.; Stefanovsky, S. V.; Nikonov, B. S.

2013-03-01

Reprocessing of spent nuclear fuel (SNF) for recovery of fissionable elements is a precondition of long-term development of nuclear energetics. Solution of this problem is hindered by the production of a great amount of liquid waste; 99% of its volume is low- and intermediate-level radioactive waste (LILW). The volume of high-level radioactive waste (HLW), which is characterized by high heat release, does not exceed a fraction of a percent. Solubility of glasses at an elevated temperature makes them unfit for immobilization of HLW, the insulation of which is ensured only by mineral-like matrices. At the same time, glasses are a perfect matrix for LILW, which are distinguished by low heat release. The solubility of borosilicate glass at a low temperature is so low that even a glass with relatively low resistance enables them to retain safety of under-ground LILW depositories without additional engineering barriers. The optimal technology of liquid confinement is their concentration and immobilization in borosilicate glasses, which are disposed in shallow-seated geological repositories. The vitrification of 1 m3 liquid LILW with a salt concentration of ˜300 kg/m3 leaves behind only 0.2 m3 waste, that is, 4-6 times less than by bitumen impregnation and 10 times less than by cementation. Environmental and economic advantages of LILW vitrification result from (1) low solubility of the vitrified LILW in natural water; (2) significant reduction of LILW volume; (3) possibility to dispose the vitrified waste without additional engineering barriers under shallow conditions and in diverse geological media; (4) the strength of glass makes its transportation and storage possible; and finally (5) reliable longterm safety of repositories. When the composition of the glass matrix for LILW is being chosen, attention should be paid to the factors that ensure high technological and economic efficiency of vitrification. The study of vitrified LILW from the Kursk nuclear power plant

Cast Stone Formulation for Nuclear Waste Immobilization at Higher Sodium Concentrations

DOE PAGES

Fox, Kevin; Cozzi, Alex; Roberts, Kimberly; ...

2014-11-01

Low activity radioactive waste at U.S. Department of Energy sites can be immobilized for permanent disposal using cementitious waste forms. This study evaluated waste forms produced with simulated wastes at concentrations up to twice that of currently operating processes. The simulated materials were evaluated for their fresh properties, which determine processability, and cured properties, which determine waste form performance. The results show potential for greatly reducing the volume of material. Fresh properties were sufficient to allow for processing via current practices. Cured properties such as compressive strength meet disposal requirements. Leachability indices provide an indication of expected long-term performance.

40 CFR 761.345 - Form of the waste to be sampled.

Code of Federal Regulations, 2010 CFR

2010-07-01

... Characterization for PCB Disposal in Accordance With § 761.62, and Sampling PCB Remediation Waste Destined for Off... waste and PCB remediation waste destined for off-site disposal must be in the form of either flattened...

LEACHING BOUNDARY IN CEMENT-BASED WASTE FORMS

EPA Science Inventory

Cement-based fixation systems are among the most commonly employed stabilization/solidification techniques. These cement haste mixtures, however, are vulnerable to ardic leaching solutions. Leaching of cement-based waste forms in acetic acid solutions with different acidic streng...

Pathways for Disposal of Commercially-Generated Tritiated Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Halverson, Nancy V.

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two ofmore » these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste

Performance of NDA techniques on a vitrified waste form

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hurd, J.R.; Veazey, G.W.; Prettyman, T.H.

1997-11-01

Rocky Flats Environmental Technology Site (RFETS) is currently considering the use of vitrified transuranic (TRU)-waste forms for the final disposition of several waste materials. To date, however, little nondestructive assay (NDA) data have been acquired in the general NDA community to assist in this endeavor. This paper describes the efforts to determine constraints and operating parameters for using NDA instrumentation on vitrified waste. The present study was conducted on a sample composed of a plutonium-contaminated ash, similar to that found in the RFETS inventory, and a borosilicate-based glass. The vitrified waste item was fabricated at Los Alamos National Laboratory (LANL)more » using methods and equipment similar to those being proposed by RFETS to treat their ash material. The focus of this study centered on the segmented gamma scanner (SGS) with 1/2-inch collimation, a technique that is presently available at RFETS. The accuracy and precision of SGS technology was evaluated, with particular attention to bias issues involving matrix geometry, homogeneity, and attenuation. Tomographic gamma scanning was utilized in the determination of the waste form homogeneity. A thermal neutron technique was also investigated and comparisons made with the gamma results.« less

Iodine solubility in a low-activity waste borosilicate glass at 1000°C

DOE PAGES

Riley, Brian J.; Schweiger, Michael J.; Kim, Dong-Sang; ...

2014-04-30

The purpose of this study was to determine the solubility of iodine in a low-activity waste borosilicate glass when heated inside an evacuated and sealed fused quartz ampoule. The iodine was added to glass frit as KI in quantities of 100–24000 ppm iodine (by mass), each mixture was added to an ampoule, the ampoules were heated at 1000 °C for 2h, and then air quenched. In samples with ≥12000 ppm iodine, low viscosity salt phases were observed on the surface of the melts during cooling that solidified into a white coating upon cooling. These salts were identified as mixtures ofmore » KI, NaI, and Na 2SO 4 with X-ray diffraction (XRD). The iodine concentrations in glass specimens were analyzed with inductively-coupled plasma mass spectrometry and the overall iodine solubility was determined to be 10000 ppm by mass. Several crystalline inclusions of iodine sodalite, Na 8(AlSiO 4) 6I 2, were observed in the 24000 ppm specimen as determined by micro-XRD and wavelength dispersive spectroscopy.« less

Effect of electric signal frequency and form on physical-chemical oxidation of organic wastes

NASA Astrophysics Data System (ADS)

Morozov, Yegor; Tikhomirov, Alexander A.; Trifonov, Sergey V.; Kudenko, D.. Yurii A.

The behavior conditions of physical-chemical reactions securing organic wastes’ oxidation in H _{2}O _{2} aqueous medium aimed at an increase of mass exchange processes in a life support system (LSS) for a space purpose have been under study. The character of dependence of organic wastes oxidation rate in H _{2}O _{2} aqueous medium, activated with alternating current of different frequency and form have been considered. Ways of those parameters optimization for the purpose to efficiently increase the physical-chemical decomposition of organic wastes in LSS have been proposed. Specifically, power consumption and reaction time of wastes mineralization have been determined to reduce more than twice. Involvement ways of mineralized organic wastes received in intrasystem mass exchange have been shown. Application feasibility of the obtained results both for space and terrestrial purpose has been discussed. Key words: life support sustem, mineralization, turnover, frequency, organic wastes

Mercury in aqueous tank waste at the Savannah River Site: Facts, forms, and impacts

DOE PAGES

Bannochie, C. J.; Fellinger, T. L.; Garcia-Strickland, P.; ...

2017-03-28

Over the past two years, there has been an intense effort to understand the chemistry of mercury across the Savannah River Site’s high-level liquid waste system to determine the impacts of various mercury species. This effort started after high concentrations of mercury were measured in the leachates from a toxicity characteristic leaching procedure (TCLP) test on the low-level cementitious waste form produced in the Savannah River Saltstone facility. Speciation showed the dominant form of leached mercury to be the methylmercury cation. Neither the source of the methylmercury nor its concentration in the Saltstone feed was well established at the timemore » of the testing. Finally, this assessment of mercury was necessary to inform points in the process operations that may be subject to new separation technologies for the removal of mercury.« less

Mercury in aqueous tank waste at the Savannah River Site: Facts, forms, and impacts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bannochie, C. J.; Fellinger, T. L.; Garcia-Strickland, P.

Over the past two years, there has been an intense effort to understand the chemistry of mercury across the Savannah River Site’s high-level liquid waste system to determine the impacts of various mercury species. This effort started after high concentrations of mercury were measured in the leachates from a toxicity characteristic leaching procedure (TCLP) test on the low-level cementitious waste form produced in the Savannah River Saltstone facility. Speciation showed the dominant form of leached mercury to be the methylmercury cation. Neither the source of the methylmercury nor its concentration in the Saltstone feed was well established at the timemore » of the testing. Finally, this assessment of mercury was necessary to inform points in the process operations that may be subject to new separation technologies for the removal of mercury.« less

Underground waste barrier structure

DOEpatents

Saha, Anuj J.; Grant, David C.

1988-01-01

Disclosed is an underground waste barrier structure that consists of waste material, a first container formed of activated carbonaceous material enclosing the waste material, a second container formed of zeolite enclosing the first container, and clay covering the second container. The underground waste barrier structure is constructed by forming a recessed area within the earth, lining the recessed area with a layer of clay, lining the clay with a layer of zeolite, lining the zeolite with a layer of activated carbonaceous material, placing the waste material within the lined recessed area, forming a ceiling over the waste material of a layer of activated carbonaceous material, a layer of zeolite, and a layer of clay, the layers in the ceiling cojoining with the respective layers forming the walls of the structure, and finally, covering the ceiling with earth.

Method for forming microspheres for encapsulation of nuclear waste

DOEpatents

Angelini, Peter; Caputo, Anthony J.; Hutchens, Richard E.; Lackey, Walter J.; Stinton, David P.

1984-01-01

Microspheres for nuclear waste storage are formed by gelling droplets containing the waste in a gelation fluid, transferring the gelled droplets to a furnace without the washing step previously used, and heating the unwashed gelled droplets in the furnace under temperature or humidity conditions that result in a substantially linear rate of removal of volatile components therefrom.

Epsilon Metal Waste Form for Immobilization of Noble Metals from Used Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crum, Jarrod V.; Strachan, Denis M.; Rohatgi, Aashish

2013-10-01

Epsilon metal (ε-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass and thus the processing problems related there insolubility in glass. This work focused on the processing aspects of the epsilonmore » metal waste form development. Epsilon metal is comprised of refractory metals resulting in high reaction temperatures to form the alloy, expected to be 1500 - 2000°C making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).« less

Disposal of low-level radioactive waste. Impact on the medical profession

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brill, D.R.; Allen, E.W.; Lutzker, L.G.

1985-11-01

During 1985, low-level radioactive waste disposal has become a critical concern. The issue has been forced by the threatened closure of the three commercial disposal sites. The medical community has used radioactive isotopes for decades in nuclear medicine, radiation therapy, radioimmunoassay, and biomedical research. Loss of disposal capacity for radioactive wastes generated by these activities, by the suppliers of radioisotopes, and by pharmaceutical companies will have a profound impact on the medical profession.

Crystallization of rhenium salts in a simulated low-activity waste borosilicate glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; McCloy, John S.; Goel, Ashutosh

2013-04-01

This study presents a new method for looking at the solubility of volatile species in simulated low-activity waste glass. The present study looking at rhenium salts is also applicable to real applications involving radioactive technetium salts. In this synthesis method, oxide glass powder is mixed with the volatiles species, vacuum-sealed in a fused quartz ampoule, and then heat-treated under vacuum in a furnace. This technique restricts the volatile species to the headspace above the melt but still within the sealed ampoule, thus maximizing the volatile concentration in contact with the glass. Various techniques were used to measure the solubility ofmore » rhenium in glass and include energy dispersive spectroscopy, wavelength dispersive spectroscopy, laser ablation inductively-coupled plasma mass spectroscopy, and inductively-coupled plasma optical emission spectroscopy. The Re-solubility in this glass was determined to be ~3004 parts per million Re atoms. Above this concentration, the salts separated out of the melt as inclusions and as a low viscosity molten salt phase on top of the melt observed during and after cooling. This salt phase was analyzed with X-ray diffraction, scanning electron microscopy as well as some of the other aforementioned techniques and identified to be composed of alkali perrhenate and alkali sulfate.« less

40 CFR 761.345 - Form of the waste to be sampled.

Code of Federal Regulations, 2012 CFR

2012-07-01

....345 Section 761.345 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) TOXIC... Characterization for PCB Disposal in Accordance With § 761.62, and Sampling PCB Remediation Waste Destined for Off... waste and PCB remediation waste destined for off-site disposal must be in the form of either flattened...

40 CFR 761.345 - Form of the waste to be sampled.

Code of Federal Regulations, 2011 CFR

2011-07-01

....345 Section 761.345 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) TOXIC... Characterization for PCB Disposal in Accordance With § 761.62, and Sampling PCB Remediation Waste Destined for Off... waste and PCB remediation waste destined for off-site disposal must be in the form of either flattened...

40 CFR 761.345 - Form of the waste to be sampled.

Code of Federal Regulations, 2014 CFR

2014-07-01

....345 Section 761.345 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) TOXIC... Characterization for PCB Disposal in Accordance With § 761.62, and Sampling PCB Remediation Waste Destined for Off... waste and PCB remediation waste destined for off-site disposal must be in the form of either flattened...

Testing and Performance Validation of a Shielded Waste Segregation and Clearance Monitor Designed for the Measurement of Low Level Waste-13043

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mason, John A.; Burke, Kevin J.; Towner, Antony C.N.

This paper describes the development, testing and validation of a shielded waste segregation and clearance monitor designed for the measurement of low-density low-level waste (LLW). The monitor is made of a measurement chamber surrounded by detectors and a shielded outer frame. The shielded chamber consists of a steel frame, which contains typically 1.5 inches (3.81 cm) of lead and 0.5 inches (1.27 cm) of steel shielding. Inside the shielding are plastic scintillator panels, which serve as gross gamma ray detectors. The detector panels, with embedded photomultipliers, completely surround the internal measurement chamber on all 6 sides. Care has been takenmore » to distribute the plastic scintillator detectors in order to optimise both the efficiency for gamma ray detection and at the same time achieve a volumetric sensitivity, which is as uniform as possible. A common high voltage power supply provides the bias voltage for each of the six photomultipliers. The voltage signals arising from the detectors and photomultipliers are amplified by six sensitive amplifiers. Each amplifier incorporates a single channel analyser with both upper and lower thresholds and the digitised counts from each detector are recorded on six scalars. Operation of the device is by means of a microprocessor from which the scalars are controlled. An internal load cell linked to the microprocessor determines the weight of the waste object, and this information is used to calculate the specific activity of the waste. The monitor makes background measurements when the shielded door is closed and a sample, usually a bag of low-density waste, is not present in the measurement chamber. Measurements of the minimum detectable activity (MDA) of an earlier large volume prototype instrument are reported as part of the development of the Waste Segregation and Clearance Monitor (WSCM) described in the paper. For the optimised WSCM a detection efficiency of greater than 32% was measured using a small Cs

1995 solid waste 30-year characteristics volume summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Templeton, K.J.; DeForest, T.J.; Rice, G.I.

1995-10-01

The Hanford Site has been designated by the US Department of Energy (DOE) to store, treat, and dispose of solid waste received from both onsite and offsite generators. This waste is currently or planned to be generated from ongoing operations, maintenance and deactivation activities, decontamination and decommissioning (D&D) of facilities, and environmental restoration (ER) activities. This document, prepared by Pacific Northwest Laboratory (PNL) under the direction of Westinghouse Hanford Company (WHC), describes the characteristics of the waste to be shipped to Hanford`s SWOC. The physical waste forms and hazardous constituents are described for the low-level mixed waste (LLMW) and themore » transuranic - transuranic mixed waste (TW{underscore}TRUM).« less

Microbial activity in argillite waste storage cells for the deep geological disposal of French bituminous medium activity long lived nuclear waste: Impact on redox reaction kinetics and potential

NASA Astrophysics Data System (ADS)

Albrecht, A.; Leone, L.; Charlet, L.

2009-04-01

for the initial development of microbial colonies is the high pH controlled by the cement solution. Archea are able to survive at high pH, when hydrogen gas is available as an energy sources; they are therefore likely candidates for an initial biofilm formation. It can not be excluded that other micro-organisms such as fungi may develop as well in such conditions. It also needs to be evaluated how conditions change with time and how this affects microbial ecology. The following is known about the impact of microbes on the waste cell biogeochemistry: • enhancement of redox reaction kinetics (particularly involving nitrates, sulphate, selenate, pertechnetate, organic matter and H2), thus a faster move towards reducing conditions, important to guarantee the low mobility of critical RN, • increased retardation of mobile RN in biofilms (i.e. adsorption on microbial cell surfaces and products of possible biomineralization); complexation by embedded extracellular polymeric substances, • secretion of organic substances (i.e siderophores) known to complex RN and to enhance their mobility, • biodegradation of dissolved organic substances, such as those released form the waste (organic acids) or generated by microbes, • production of CO2 or other gases that may affect cement integrity. Quantification of microbial activity has been implemented into biogeochemical models but the important parameters describing their evolution and metabolism in the real system (ecology, mass, energy sources, metabolites) need to be obtained via specific empirical studies. Such studies require a particular trans-disciplinary approach that brings together the competence of chemical and environmental engineers, microbiologists and system modellers.

Ground-water protection, low-level waste, and below regulatory concern: What`s the connection?

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gruhlke, J.M.; Galpin, F.L.

1991-12-31

The Environmental Protection Agency (EPA) has a responsibility to protect ground water and drinking water under a wide variety of statutes. Each statute establishes different but specific requirements for EPA and applies to diverse environmental contaminants. Radionuclides are but one of the many contaminants subject to this regulatory matrix. Low-level radioactive waste (LLW) and below regulatory concern (BRC) are but two of many activities falling into this regulatory structure. The nation`s ground water serves as a major source of drinking water, supports sensitive ecosystems, and supplies the needs of agriculture and industry. Ground water can prove enormously expensive to cleanmore » up. EPA policy for protecting ground water has evolved considerably over the last ten years. The overall goal is to prevent adverse effects to human health, both now and in the future, and to protect the integrity of the nation`s ground-water resources. The Agency uses the Maximum Contaminant Levels (MCLs) under the Safe Drinking Water Act as reference points for protection in both prevention and remediation activities. What`s the connection? Both low-level waste management and disposal activities and the implementation of below regulatory concern related to low-level waste disposal have the potential for contaminating ground water. EPA is proposing to use the MCLs as reference points for low-level waste disposal and BRC disposal in order to define limits to the environmental contamination of ground water that is, or may be, used for drinking water.« less

A Strategy to Conduct an Analysis of the Long-Term Performance of Low-Activity Waste Glass in a Shallow Subsurface Disposal System at Hanford

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neeway, James J.; Pierce, Eric M.; Freedman, Vicky L.

2014-08-04

The federal facilities located on the Hanford Site in southeastern Washington State have been used extensively by the U.S. government to produce nuclear materials for the U.S. strategic defense arsenal. Currently, the Hanford Site is under the stewardship of the U.S. Department of Energy (DOE) Office of Environmental Management (EM). A large inventory of radioactive and mixed waste resulting from the production of nuclear materials has accumulated, mainly in 177 underground single- and double-shell tanks located in the central plateau of the Hanford Site (Mann et al., 2001). The DOE-EM Office of River Protection (ORP) is proceeding with plans tomore » immobilize and permanently dispose of the low-activity waste (LAW) fraction onsite in a shallow subsurface disposal facility (the Integrated Disposal Facility [IDF]). Pacific Northwest National Laboratory (PNNL) was contracted to provide the technical basis for estimating radionuclide release from the engineered portion of the IDF (the source term) as part of an immobilized low-activity waste (ILAW) glass testing program to support future IDF performance assessments (PAs).« less

Characterization and Leaching Tests of the Fluidized Bed Steam Reforming (FBSR) Waste Form for LAW Immobilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neeway, James J.; Qafoku, Nikolla; Brown, Christopher F.

2013-10-01

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) have been evaluated. One such immobilization technology is the Fluidized Bed Steam Reforming (FBSR) granular product. The FBSR granular product is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals. Production of the FBSR mineral product has been demonstrated both at the industrial and laboratory scale. Pacific Northwest National Laboratory (PNNL) was involved in an extensive characterization campaign. This goal of this campaign was study the durability of the FBSR mineral product and the mineral product encapsulated in a monolith to meet compressive strength requirements. This paper gives anmore » overview of results obtained using the ASTM C 1285 Product Consistency Test (PCT), the EPA Test Method 1311 Toxicity Characteristic Leaching Procedure (TCLP), and the ASTMC 1662 Single-Pass Flow-Through (SPFT) test. Along with these durability tests an overview of the characteristics of the waste form has been collected using Scanning Electron Microscopy (SEM), X-ray Diffraction (XRD), microwave digestions for chemical composition, and surface area from Brunauer, Emmett, and Teller (BET) theory.« less

Assessment and evaluation of engineering options at a low-level radioactive waste storage site

NASA Astrophysics Data System (ADS)

Kanehiro, B. Y.; Guvanasen, V.

1982-09-01

Solutions to hydrologic and geotechnical problems associated with existing disposal sites were sought and the efficiency of engineering options that were proposed to improve the integrity of such sites were evaluated. The Weldon Spring site is generally like other low-level nuclear waste sites, except that the wastes are primarily in the form of residues and contaminated rubble from the processing of uranium and thorium ores rather than industrial isotopes or mill tailings.

Minerals and design of new waste forms for conditioning nuclear waste

NASA Astrophysics Data System (ADS)

Montel, Jean-Marc

2011-02-01

Safe storage of radioactive waste is a major challenge for the nuclear industry. Mineralogy is a good basis for designing ceramics, which could eventually replace nuclear glasses. This requires a new storage concept: separation-conditioning. Basic rules of crystal chemistry allow one to select the most suitable structures and natural occurrences allow assessing the long-term performance of ceramics in a geological environment. Three criteria are of special interest: compatibility with geological environment, resistance to natural fluids, and effects of self-irradiation. If mineralogical information is efficient for predicting the behaviour of common, well-known minerals, such as zircon, monazite or apatite, more research is needed to rationalize the long-term behaviour of uncommon waste form analogs.

National Low-Level Waste Management Program Radionuclide Report Series. Volume 10, Nickel-63

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carboneau, M.L.; Adams, J.P.

1995-02-01

This report outlines the basic radiological, chemical, and physical characteristics of nickel-63 ({sup 63}Ni) and examines how these characteristics affect the behavior of {sup 63}Ni in various environmental media, such as soils, groundwater, plants, animals, the atmosphere, and the human body. Discussions also include methods of {sup 63}Ni production, waste types, and waste forms that contain {sup 63}Ni. The primary source of {sup 63}Ni in the environment has been low-level radioactive waste material generated as a result of neutron activation of stable {sup 62}Ni that is present in the structural components of nuclear reactor vessels. {sup 63}Ni enters the environmentmore » from the dismantling activities associated with nuclear reactor decommissioning. However, small amounts of {sup 63}Ni have been detected in the environment following the testing of thermonuclear weapons in the South Pacific. Concentrations as high as 2.7 Bq{sup a} per gram of sample (or equivalently 0.0022 parts per billion) were observed on Bikini Atoll (May 1954). {sup 63}Ni was not created as a fission product species (e.g., from {sup 235}U or {sup 239}Pu fissions), but instead was produced as a result of neutron capture in {sup 63}Ni, a common nickel isotope present in the stainless steel components of nuclear weapons (e.g., stainless-304 contains {approximately}9% total Ni or {approximately}0.3% {sup 63}Ni).« less

Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Riley, Brian J.; Mccloy, John S.; Crum, Jarrod V.

2014-01-17

The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for themore » Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.« less

Low Carbon Footprint mortar from Pozzolanic Waste Material

NASA Astrophysics Data System (ADS)

Mehmannavaz, Taha; Mehman navaz, Hossein Ali; Moayed Zefreh, Fereshteh; Aboata, Zahra

2017-04-01

Nowadays, Portland cement clinker leads to emission of CO2 into the atmosphere and therefore causes greenhouse effect. Incorporating of Palm Oil Fuel Ash (POFA) and Pulverized Fuel Ash (PFA) as partial cement replacement materials into mix of low carbon mortar decreases the amount of cement use and reduces high dependence on cements compared to ordinary mortar. The result of this research supported use of the new concept in preparing low carbon mortar for industrial constructions. Strength of low carbon mortar with POFA and PFA replacement in cement was affected and changed by replacing percent finesse, physical and chemical properties and pozzolanic activity of these wastes. Waste material replacement instead of Ordinary Portland Cement (OPC) was used in this study. This in turn was useful for promoting better quality of construction and innovative systems in construction industry, especially in Malaysia. This study was surely a step forward to achieving quality products which were affordable, durable and environmentally friendly. Disposing ash contributes to shortage of landfill space in Malaysia. Besides, hazard of ash might be another serious issue for human health. The ash disposal area also might create a new problem, which is the area's sedimentation and erosion.

Epsilon metal waste form for immobilization of noble metals from used nuclear fuel

NASA Astrophysics Data System (ADS)

Crum, Jarrod V.; Strachan, Denis; Rohatgi, Aashish; Zumhoff, Mac

2013-10-01

Epsilon metal (ɛ-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass, thus the processing problems related to their insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high alloying temperatures, expected to be 1500-2000 °C, making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

Gravitational sedimentation of flocculated waste activated sludge.

PubMed

Chu, C P; Lee, D J; Tay, J H

2003-01-01

The sedimentation characteristics of flocculated wastewater sludge have not been satisfactorily explored using the non-destructive techniques, partially owing to the rather low solid content (ca. 1-2%) commonly noted in the biological sediments. This paper investigated, for the first time, the spatial-temporal gravitational settling characteristics of original and polyelectrolyte flocculated waste activated sludge using Computerized Axial Tomography Scanner. The waste activated sludge possessed a distinct settling characteristic from the kaolin slurries. The waste activated sludges settled more slowly and reached a lower solid fraction in the final sediment than the latter. Flocculation markedly enhanced the settleability of both sludges. Although the maximum achievable solid contents for the kaolin slurries were reduced, flocculation had little effects on the activated sludge. The purely plastic rheological model by Buscall and White (J Chem Soc Faraday Trans 1(83) (1987) 873) interpreted the consolidating sediment data, while the purely elastic model by Tiller and Leu (J. Chin. Inst. Chem. Eng. 11 (1980) 61) described the final equilibrated sediment. Flocculation produced lower yield stress during transient settling, thereby resulting in the more easily consolidated sludge than the original sample. Meanwhile, the flocculated activated sludge was stiffer in the final sediment than in the original sample. The data reported herein are valuable to the theories development for clarifier design and operation.

Finite element analysis of ion transport in solid state nuclear waste form materials

NASA Astrophysics Data System (ADS)

Rabbi, F.; Brinkman, K.; Amoroso, J.; Reifsnider, K.

2017-09-01

Release of nuclear species from spent fuel ceramic waste form storage depends on the individual constituent properties as well as their internal morphology, heterogeneity and boundary conditions. Predicting the release rate is essential for designing a ceramic waste form, which is capable of effectively storing the spent fuel without contaminating the surrounding environment for a longer period of time. To predict the release rate, in the present work a conformal finite element model is developed based on the Nernst Planck Equation. The equation describes charged species transport through different media by convection, diffusion, or migration. And the transport can be driven by chemical/electrical potentials or velocity fields. The model calculates species flux in the waste form with different diffusion coefficient for each species in each constituent phase. In the work reported, a 2D approach is taken to investigate the contributions of different basic parameters in a waste form design, i.e., volume fraction, phase dispersion, phase surface area variation, phase diffusion co-efficient, boundary concentration etc. The analytical approach with preliminary results is discussed. The method is postulated to be a foundation for conformal analysis based design of heterogeneous waste form materials.

Radiologic safety assessment for low level waste storage on TRU pads

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryan, J.P.

1986-03-17

The reference document (TA 2-1118) proposes to store Low Level Radioactive Solid Waste in B-25 boxes on concrete pads at the 643-G burial ground site, pending resolution of policy concernig its ultimate disposal. This analysis verifies that the reference proposal is safe, as long as it is applied to a limited material quantity of low specific activity, as described in the reference document. The predominant concern in the safety analysis is the emission of airborne activity as a result of tornados and fires. However, containment provided by B-25 boxes is sufficient to mitigate the consequences of these events sufficiently. Nevertheless,more » it is strongly recommended that any above-ground storage procedures include provisions for covering the waste containment boxes to prevent exposure to rainwater and subsequent corrosion if the storage period is to extend beyond one year.« less

Description of Defense Waste Processing Facility reference waste form and canister. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baxter, R.G.

1983-08-01

The Defense Waste Processing Facility (DWPF) will be located at the Savannah River Plant in Aiken, SC, and is scheduled for construction authorization during FY-1984. The reference waste form is borosilicate glass containing approx. 28 wt % sludge oxides, with the balance glass frit. Borosilicate glass was chosen because of its high resistance to leaching by water, its relatively high solubility for nuclides found in the sludge, and its reasonably low melting temperature. The glass frit contains about 58% SiO/sub 2/ and 15% B/sub 2/O/sub 3/. Leachabilities of SRP waste glasses are expected to approach 10/sup -8/ g/m/sup 2/-day basedmore » upon 1000-day tests using glasses containing SRP radioactive waste. Tests were performed under a wide variety of conditions simulating repository environments. The canister is filled with 3260 lb of glass which occupies about 85% of the free canister volume. The filled canister will generate approx. 470 watts when filled with oxides from 5-year-old sludge and 15-year-old supernate from the sludge and supernate processes. The radionuclide content of the canister is about 177,000 ci, with a radiation level of 5500 rem/h at canister surface contact. The reference canister is fabricated of standard 24-in.-OD, Schedule 20, 304L stainless steel pipe with a dished bottom, domed head, and a combined lifting and welding flange on the head neck. The overall canister length is 9 ft 10 in. with a 3/8-in. wall thickness. The 3-m canister length was selected to reduce equipment cell height in the DWPF to a practical size. The canister diameter was selected as an optimum size from glass quality considerations, a logical size for repository handling and to ensure that a filled canister with its double containment shipping cask could be accommodated on a legal-weight truck. The overall dimensions and weight appear to be compatible with preliminary assessments of repository requirements. 10 references.« less

Setting and stiffening of cementitious components in Cast Stone waste form for disposal of secondary wastes from the Hanford waste treatment and immobilization plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chung, Chul-Woo; Chun, Jaehun, E-mail: jaehun.chun@pnnl.gov; Um, Wooyong

2013-04-01

Cast Stone is a cementitious waste form, a viable option to immobilize secondary nuclear liquid wastes generated from the Hanford Waste Treatment and Immobilization Plant. However, no study has been performed to understand the flow and stiffening behavior, which is essential to ensure proper workability and is important to safety in a nuclear waste field-scale application. X-ray diffraction, rheology, and ultrasonic wave reflection methods were used to understand the specific phase formation and stiffening of Cast Stone. Our results showed a good correlation between rheological properties of the fresh mixture and phase formation in Cast Stone. Secondary gypsum formation wasmore » observed with low concentration simulants, and the formation of gypsum was suppressed in high concentration simulants. A threshold concentration for the drastic change in stiffening was found at 1.56 M Na concentration. It was found that the stiffening of Cast Stone was strongly dependent on the concentration of simulant. Highlights: • A combination of XRD, UWR, and rheology gives a better understanding of Cast Stone. • Stiffening of Cast Stone was strongly dependent on the concentration of simulant. • A drastic change in stiffening of Cast Stone was found at 1.56 M Na concentration.« less

Improved methane production from waste activated sludge with low organic content by alkaline pretreatment at pH 10.

PubMed

Feng, L Y; Yang, L Q; Zhang, L X; Chen, H L; Chen, J

2013-01-01

Sludge with low organic content always results in an unsatisfactory performance, even failure of anaerobic digestion. The alkaline pretreatment effect on anaerobic digestion of sludge with low organic content has seldom been studied although it gives many benefits for sludge with high organic content. In this study the influence of alkaline pretreatment (pH 10, an effective alkaline pH) on the solubilization and methane production from waste activated sludge (WAS) with low organic content was investigated. Results from biochemical methane potential (BMP) experiments showed that anaerobic biodegradability of WAS was greatly improved by alkaline pretreatment at pH 10. Methane production from the current WAS under conditions of pretreatment time 4 h and digestion time 15 d was 139.6 mL/g VS (volatile solids), much higher than that from the unpretreated WAS with digestion time of 20 d (75.2 mL/g VS). Also, the solubilization of WAS was significantly accelerated by alkaline pretreatment. Mechanism exploration indicated that the general activities of anaerobic microorganisms, specific activities of key enzymes and the amounts of methanogens were enhanced by alkaline pretreatment at pH 10, showing good agreement with methane production.

Electrochemical Corrosion Studies for Modeling Metallic Waste Form Release Rates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poineau, Frederic; Tamalis, Dimitri

The isotope 99Tc is an important fission product generated from nuclear power production. Because of its long half-life (t 1/2 = 2.13 ∙ 10 5 years) and beta-radiotoxicity (β⁻ = 292 keV), it is a major concern in the long-term management of spent nuclear fuel. In the spent nuclear fuel, Tc is present as an alloy with Mo, Ru, Rh, and Pd called the epsilon-phase, the relative amount of which increases with fuel burn-up. In some separation schemes for spent nuclear fuel, Tc would be separated from the spent fuel and disposed of in a durable waste form. Technetium wastemore » forms under consideration include metallic alloys, oxide ceramics and borosilicate glass. In the development of a metallic waste form, after separation from the spent fuel, Tc would be converted to the metal, incorporated into an alloy and the resulting waste form stored in a repository. Metallic alloys under consideration include Tc–Zr alloys, Tc–stainless steel alloys and Tc–Inconel alloys (Inconel is an alloy of Ni, Cr and iron which is resistant to corrosion). To predict the long-term behavior of the metallic Tc waste form, understanding the corrosion properties of Tc metal and Tc alloys in various chemical environments is needed, but efforts to model the behavior of Tc metallic alloys are limited. One parameter that should also be considered in predicting the long-term behavior of the Tc waste form is the ingrowth of stable Ru that occurs from the radioactive decay of 99Tc ( 99Tc → 99Ru + β⁻). After a geological period of time, significant amounts of Ru will be present in the Tc and may affect its corrosion properties. Studying the effect of Ru on the corrosion behavior of Tc is also of importance. In this context, we studied the electrochemical behavior of Tc metal, Tc-Ni alloys (to model Tc-Inconel alloy) and Tc-Ru alloys in acidic media. The study of Tc-U alloys has also been performed in order to better understand the nature of Tc in metallic spent fuel. Computational

Co-conditioning and dewatering of chemical sludge and waste activated sludge.

PubMed

Chang, G R; Liu, J C; Lee, D J

2001-03-01

The conditioning and dewatering behaviors of chemical and waste activated sludges from a tannery were studied. Capillary suction time (CST), specific resistance to filtration (SRF), and bound water content were used to evaluate the sludge dewatering behaviors. Zeta potentials were also measured. Experiments were conducted on each sludge conditioned and dewatered separately, and on the sludge mixed at various ratios. Results indicate that the chemical sludge was relatively difficult to be dewatered, even in the presence of polyelectrolyte. When the waste activated sludge was mixed with the chemical sludge at ratios of 1:1 and 2:1, respectively, the dewaterability of chemical sludge improved remarkably while the relatively better dewaterability of the waste activated sludge deteriorated only to a limited extent. As the mixing ratios became 4:1 and 8:1, the dewaterability of the mixed sludge was equal to that of the waste activated sludge. The optimal polyelectrolyte dosage for the mixed sludge was equal to or less than that of the waste activated sludge. It is proposed that the chemical sludges act as skeleton builders that reduce the compressibility of the mixed sludge whose dewaterability is enhanced. Bound water contents of sludge decreased at low polyelectrolyte dosage and were not significantly affected as polyelectrolyte dosage increased. Advantages and disadvantages of co-conditioning and dewatering chemical sludge and waste activated sludge were discussed.

Secondary Waste Form Screening Test Results—Cast Stone and Alkali Alumino-Silicate Geopolymer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pierce, Eric M.; Cantrell, Kirk J.; Westsik, Joseph H.

2010-06-28

PNNL is conducting screening tests on the candidate waste forms to provide a basis for comparison and to resolve the formulation and data needs identified in the literature review. This report documents the screening test results on the Cast Stone cementitious waste form and the Geopolymer waste form. Test results suggest that both the Cast Stone and Geopolymer appear to be viable waste forms for the solidification of the secondary liquid wastes to be treated in the ETF. The diffusivity for technetium from the Cast Stone monoliths was in the range of 1.2 × 10-11 to 2.3 × 10-13 cm2/smore » during the 63 days of testing. The diffusivity for technetium from the Geopolymer was in the range of 1.7 × 10-10 to 3.8 × 10-12 cm2/s through the 63 days of the test. These values compare with a target of 1 × 10-9 cm2/s or less. The Geopolymer continues to show some fabrication issues with the diffusivities ranging from 1.7 × 10-10 to 3.8 × 10-12 cm2/s for the better-performing batch to from 1.2 × 10-9 to 1.8 × 10-11 cm2/s for the poorer-performing batch. In the future more comprehensive and longer term performance testing will be conducted, to further evaluate whether or not these waste forms will meet the regulation and performance criteria needed to cost-effectively dispose of secondary wastes.« less

Conversion of transuranic waste to low level waste by decontamination: a site specific update

DOE Office of Scientific and Technical Information (OSTI.GOV)

Allen, R.P.; Hazelton, R.F.

1985-09-01

As a followup to an FY-1984 cost/benefit study, a program was conducted in FY-1985 to transfer to the relevant DOE sites the information and technology for the direct conversion of transuranic (TRU) waste to low-level waste (LLW) by decontamination. As part of this work, the economic evaluation of the various TRUW volume reduction and conversion options was updated and expanded to include site-specific factors. The results show, for the assumptions used, that size reduction, size reduction followed by decontamination, or in situ decontamination are cost effective compared with the no-processing option. The technology transfer activities included site presentations and discussionsmore » with operations and waste management personnel to identify application opportunities and site-specific considerations and constraints that could affect the implementation of TRU waste conversion principles. These discussions disclosed definite potential for the beneficial application of these principles at most of the sites, but also confirmed the existence of site-specific factors ranging from space limitations to LLW disposal restrictions that could preclude particular applications or diminish expected benefits. 8 refs., 2 figs., 4 tabs.« less

Effect of Sulfate on Rhenium Partitioning during Melting of Low-Activity Waste Glass Feeds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jin, Tongan; Kim, Dong-Sang; Schweiger, Michael J.

2015-10-01

The volatile loss of technetium-99 (99Tc) is a major concern of the low-activity waste (LAW) vitrification at Hanford. We investigated the incorporation and volatile loss of Re (a nonradioactive surrogate for 99Tc) during batch-to-glass conversion up to 1100°C. The AN-102 feed, which is one of the representative Hanford LAW feeds, containing 0.59 wt% of SO3 (in glass if 100% retained) was used. The modified sulfate-free AN-102_0S feed was also tested to investigate the effect of sulfate on Re partitioning and retention during melting. After heating of the dried melter feed (mixture of LAW simulant and glass forming/modifying additives) to differentmore » temperatures, the heat-treated samples were quenched. For each heat-treated sample, the salts (soluble components in room temperature leaching), early glass forming melt (soluble components in 80°C leaching), and insoluble solids were separated by a two-step leaching and the chemical compositions of each phase were quantitatively analyzed. The final retention ratio of AN-102 and AN-102_0S in glass (insoluble solids) are 32% and 63% respectively. The presence of sulfate in the salt phase between 600 and 800°C leads to a significantly higher Re loss via volatilization from the salt layer. At ≥800°C, for both samples, there is no more incorporation of Re into the insoluble phase because: for AN-102_0S there is no salt left i.e., the split into the insoluble and gas phases is complete by 800°C and for AN-102 all the Re contained in the remaining salt phase is lost through volatilization. The present results on the effect of sulfate, although not directly applicable to LAW vitrification in the melter, will be used to understand the mechanism of Re incorporation into glass to eventually develop the methods that can increase the 99Tc retention during LAW vitrification at Hanford.« less

Morphology and pH changes in leached solidified/stabilized waste forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cheng, K.Y.; Bishop, P.L.

1996-12-31

Leaching of cement-based waste forms in acetic acid solutions with different acidic strengths has been investigated in this work. The examination of the morphology and pH profile along the acid penetration route by an optical microscope and various pH color indicators is reported. A clear-cut leaching boundary, where the pH changes from below 6 in the leached surface layers to above 12 in the unleached waste form, was observed in every leached sample.

Solid Waste Activity Packet for Teachers.

ERIC Educational Resources Information Center

Illinois Univ., Urbana. Cooperative Extension Service.

This solid waste activity packet introduces students to the solid waste problem in Illinois. Topics explore consumer practices in the market place, packaging, individual and community garbage generation, and disposal practices. The activities provide an integrated approach to incorporating solid waste management issues into subject areas. The…

Comparative risk assessments for the production and interim storage of glass and ceramic waste forms: Defense waste processing facility

NASA Astrophysics Data System (ADS)

Huang, J. C.; Wright, W. V.

1982-04-01

The Defense Waste Processing Facility (DWPF) for immobilizing nuclear high level waste (HLW) is scheduled to be built. High level waste is produced when reactor components are subjected to chemical separation operations. Two candidates for immobilizing this HLW are borosilicate glass and crystalline ceramic, either being contained in weld sealed stainless steel canisters. A number of technical analyses are being conducted to support a selection between these two waste forms. The risks associated with the manufacture and interim storage of these two forms in the DWPF are compared. Process information used in the risk analysis was taken primarily from a DWPF processibility analysis. The DWPF environmental analysis provided much of the necessary environmental information.

Characterization and Leaching Tests of the Fluidized Bed Steam Reforming (FBSR) Waste Form for LAW Immobilization - 13400

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neeway, James J.; Qafoku, Nikolla P.; Peterson, Reid A.

2013-07-01

Several supplemental technologies for treating and immobilizing Hanford low activity waste (LAW) have been evaluated. One such immobilization technology is the Fluidized Bed Steam Reforming (FBSR) granular product. The FBSR granular product is composed of insoluble sodium aluminosilicate (NAS) feldspathoid minerals. Production of the FBSR mineral product has been demonstrated both at the industrial and laboratory scale. Pacific Northwest National Laboratory (PNNL) was involved in an extensive characterization campaign. The goal of this campaign was to study the durability of the FBSR mineral product and the encapsulated FBSR product in a geo-polymer monolith. This paper gives an overview of resultsmore » obtained using the ASTM C 1285 Product Consistency Test (PCT), the EPA Test Method 1311 Toxicity Characteristic Leaching Procedure (TCLP), and the ASTMC 1662 Single-Pass Flow-Through (SPFT) test. Along with these durability tests an overview of the characteristics of the waste form has been collected using Scanning Electron Microscopy (SEM), X-ray Diffraction (XRD), microwave digestions for chemical composition, and surface area from Brunauer, Emmett, and Teller (BET) theory. (authors)« less

FY16 Annual Accomplishments - Waste Form Development and Performance: Evaluation Of Ceramic Waste Forms - Comparison Of Hot Isostatic Pressed And Melt Processed Fabrication Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amoroso, J.; Dandeneau, C.

FY16 efforts were focused on direct comparison of multi-phase ceramic waste forms produced via melt processing and HIP methods. Based on promising waste form compositions previously devised at SRNL, simulant material was prepared at SRNL and a portion was sent to the Australian Nuclear Science and Technology Organization (ANSTO) for HIP treatments, while the remainder of the material was melt processed at SRNL. The microstructure, phase formation, elemental speciation, and leach behavior, and radiation stability of the fabricated ceramics was performed. In addition, melt-processed ceramics designed with different fractions of hollandite, zirconolite, perovskite, and pyrochlore phases were investigated. for performancemore » and properties.« less

Operating Experience and Lessons Learned in the Use of Soft-Sided Packaging for Transportation and Disposal of Low Activity Radioactive Waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kapoor, A.; Gordon, S.; Goldston, W.

2013-07-08

This paper describes the operating experience and lessons learned at U.S. Department of Energy (DOE) sites as a result of an evaluation of potential trailer contamination and soft-sided packaging integrity issues related to the disposal of low-level and mixed low-level (LLW/MLLW) radioactive waste shipments. Nearly 4.3 million cubic meters of LLW/MLLW will have been generated and disposed of during fiscal year (FY) 2010 to FY 2015—either at commercial disposal sites or disposal sites owned by DOE. The LLW/MLLW is packaged in several different types of regulatory compliant packaging and transported via highway or rail to disposal sites safely and efficientlymore » in accordance with federal, state, and local regulations and DOE orders. In 1999, DOE supported the development of LLW containers that are more volumetrically efficient, more cost effective, and easier to use as compared to metal or wooden containers that existed at that time. The DOE Idaho National Engineering and Environmental Laboratory (INEEL), working in conjunction with the plastic industry, tested several types of soft-sided waste packaging systems that meet U.S. Department of Transportation requirements for transport of low specific activity and surface contaminated objects. Since then, soft-sided packaging of various capacities have been used successfully by the decontamination and decommissioning (D&D) projects to package, transport, and dispose D&D wastes throughout the DOE complex. The joint team of experts assembled by the Energy Facility Contractors Group from DOE waste generating sites, DOE and commercial waste disposal facilities, and soft-sided packaging suppliers conducted the review of soft-sided packaging operations and transportation of these packages to the disposal sites. As a result of this evaluation, the team developed several recommendations and best practices to prevent or minimize the recurrences of equipment contamination issues and proper use of soft-sided packaging for

Methods of vitrifying waste with low melting high lithia glass compositions

DOEpatents

Jantzen, Carol M.; Pickett, John B.; Cicero-Herman, Connie A.; Marra, James C.

2001-01-01

The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods. The invention is useful for stabilization of numerous types of waste materials, including aqueous waste streams, sludge solids, mixtures of aqueous supernate and sludge solids, combinations of spent filter aids from waste water treatment and waste sludges, supernate alone, incinerator ash, incinerator offgas blowdown, or combinations thereof, geological mine tailings and sludges, asbestos, inorganic filter media, cement waste forms in need of remediation, spent or partially spent ion exchange resins or zeolites, contaminated soils, lead paint, etc. The decrease in melting point achieved by the present invention desirably prevents volatilization of hazardous or radioactive species during vitrification.

Dilute condition corrosion behavior of glass-ceramic waste form

DOE PAGES

Crum, Jarrod V.; Neeway, James J.; Riley, Brian J.; ...

2016-08-11

Borosilicate glass-ceramics are being developed to immobilize high-level waste generated by aqueous reprocessing into a stable waste form. The corrosion behavior of this multiphase waste form is expected to be complicated by multiple phases and crystal-glass interfaces. A modified single-pass flow-through test was performed on polished monolithic coupons at a neutral pH (25 °C) and 90 °C for 33 d. The measured glass corrosion rates by micro analysis in the samples ranged from 0.019 to 0.29 g m -2 d -1 at a flow rate per surface area = 1.73 × 10 -6 m s -1. The crystal phases (oxyapatitemore » and Ca-rich powellite) corroded below quantifiable rates, by micro analysis. While, Ba-rich powellite corroded considerably in O10 sample. The corrosion rates of C1 and its replicate C20 were elevated an order of magnitude by mechanical stresses at crystal-glass interface caused by thermal expansion mismatch during cooling and unique morphology (oxyapatite clustering).« less

Dilute condition corrosion behavior of glass-ceramic waste form

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crum, Jarrod V.; Neeway, James J.; Riley, Brian J.

Borosilicate glass-ceramics are being developed to immobilize high-level waste generated by aqueous reprocessing into a stable waste form. The corrosion behavior of this multiphase waste form is expected to be complicated by multiple phases and crystal-glass interfaces. A modified single-pass flow-through test was performed on polished monolithic coupons at a neutral pH (25 °C) and 90 °C for 33 d. The measured glass corrosion rates by micro analysis in the samples ranged from 0.019 to 0.29 g m -2 d -1 at a flow rate per surface area = 1.73 × 10 -6 m s -1. The crystal phases (oxyapatitemore » and Ca-rich powellite) corroded below quantifiable rates, by micro analysis. While, Ba-rich powellite corroded considerably in O10 sample. The corrosion rates of C1 and its replicate C20 were elevated an order of magnitude by mechanical stresses at crystal-glass interface caused by thermal expansion mismatch during cooling and unique morphology (oxyapatite clustering).« less

Safe disposal of radionuclides in low-level radioactive-waste repository sites; Low-level radioactive-waste disposal workshop, U.S. Geological Survey, July 11-16, 1987, Big Bear Lake, Calif., Proceedings

USGS Publications Warehouse

Bedinger, Marion S.; Stevens, Peter R.

1990-01-01

In the United States, low-level radioactive waste is disposed by shallow-land burial. Low-level radioactive waste generated by non-Federal facilities has been buried at six commercially operated sites; low-level radioactive waste generated by Federal facilities has been buried at eight major and several minor Federally operated sites (fig. 1). Generally, low-level radioactive waste is somewhat imprecisely defined as waste that does not fit the definition of high-level radioactive waste and does not exceed 100 nCi/g in the concentration of transuranic elements. Most low-level radioactive waste generated by non-Federal facilities is generated at nuclear powerplants; the remainder is generated primarily at research laboratories, hospitals, industrial facilities, and universities. On the basis of half lives and concentrations of radionuclides in low-level radioactive waste, the hazard associated with burial of such waste generally lasts for about 500 years. Studies made at several of the commercially and Federally operated low-level radioactive-waste repository sites indicate that some of these sites have not provided containment of waste nor the expected protection of the environment.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wall, Nathalie A.; Neeway, James J.; Qafoku, Nikolla P.

2015-09-30

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,more » 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

Toward understanding the effect of low-activity waste glass composition on sulfur solubility

DOE PAGES

Vienna, John D.; Kim, Dong -Sang; Muller, Isabelle S.; ...

2014-07-24

The concentration of sulfur in nuclear waste glass melter feed must be maintained below the point where salt accumulates on the melt surface. The allowable concentrations may range from 0.37 to over 2.05 weight percent (of SO 3 on a calcined oxide basis) depending on the composition of the melter feed and processing conditions. If the amount of sulfur exceeds the melt tolerance level, a molten salt will accumulate, which may upset melter operations and potentially shorten the useful life of the melter. At the Hanford site, relatively conservative limits have been placed on sulfur loading in melter feed, whichmore » in turn significantly increases the amount of glass that will be produced. Crucible-scale sulfur solubility data and scaled melter sulfur tolerance data have been collected on simulated Hanford waste glasses over the last 15 years. These data were compiled and analyzed. A model was developed to predict the solubility of SO 3 in glass based on 252 simulated Hanford low-activity waste (LAW) glass compositions. This model represents the data well, accounting for over 85% of the variation in data, and was well validated. The model was also found to accurately predict the tolerance for sulfur in melter feed for 13 scaled melter tests of simulated LAW glasses. The model can be used to help estimate glass volumes and make informed decisions on process options. The model also gives quantitative estimates of component concentration effects on sulfur solubility. The components that most increase sulfur solubility are Li 2O > V 2O 5> CaO ≈ P 2O 5 > Na 2O ≈ B 2O 3 > K 2O. The components that most decrease sulfur solubility are Cl > Cr 2O 3 > Al 2O 3 > ZrO 2 ≈ SnO 2 > Others ≈ SiO 2. As a result, the order of component effects is similar to previous literature data, in most cases.« less

Durability and degradation of HT9 based alloy waste forms with variable Ni and Cr content

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, L.

2016-12-31

Short-term electrochemical and long-term hybrid electrochemical corrosion tests were performed on alloy waste forms in reference aqueous solutions that bound postulated repository conditions. The alloy waste forms investigated represent candidate formulations that can be produced with advanced electrochemical treatment of used nuclear fuel. The studies helped to better understand the alloy waste form durability with differing concentrations of nickel and chromium, species that can be added to alloy waste forms to potentially increase their durability and decrease radionuclide release into the environment.

Application of Molecular Techniques To Elucidate the Influence of Cellulosic Waste on the Bacterial Community Structure at a Simulated Low-Level-Radioactive-Waste Site▿ †

PubMed Central

Field, Erin K.; D'Imperio, Seth; Miller, Amber R.; VanEngelen, Michael R.; Gerlach, Robin; Lee, Brady D.; Apel, William A.; Peyton, Brent M.

2010-01-01

Low-level-radioactive-waste (low-level-waste) sites, including those at various U.S. Department of Energy sites, frequently contain cellulosic waste in the form of paper towels, cardboard boxes, or wood contaminated with heavy metals and radionuclides such as chromium and uranium. To understand how the soil microbial community is influenced by the presence of cellulosic waste products, multiple soil samples were obtained from a nonradioactive model low-level-waste test pit at the Idaho National Laboratory. Samples were analyzed using 16S rRNA gene clone libraries and 16S rRNA gene microarray (PhyloChip) analyses. Both methods revealed changes in the bacterial community structure with depth. In all samples, the PhyloChip detected significantly more operational taxonomic units, and therefore relative diversity, than the clone libraries. Diversity indices suggest that diversity is lowest in the fill and fill-waste interface (FW) layers and greater in the wood waste and waste-clay interface layers. Principal-coordinate analysis and lineage-specific analysis determined that the Bacteroidetes and Actinobacteria phyla account for most of the significant differences observed between the layers. The decreased diversity in the FW layer and increased members of families containing known cellulose-degrading microorganisms suggest that the FW layer is an enrichment environment for these organisms. These results suggest that the presence of the cellulosic material significantly influences the bacterial community structure in a stratified soil system. PMID:20305022

Protocol for the E-Area Low Level Waste Facility Disposal Limits Database

DOE Office of Scientific and Technical Information (OSTI.GOV)

Swingle, R

2006-01-31

A database has been developed to contain the disposal limits for the E-Area Low Level Waste Facility (ELLWF). This database originates in the form of an EXCEL{copyright} workbook. The pertinent sheets are translated to PDF format using Adobe ACROBAT{copyright}. The PDF version of the database is accessible from the Solid Waste Division web page on SHRINE. In addition to containing the various disposal unit limits, the database also contains hyperlinks to the original references for all limits. It is anticipated that database will be revised each time there is an addition, deletion or revision of any of the ELLWF radionuclidemore » disposal limits.« less

The mixed low-level waste problem in BE/NWN capsule

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hensley, D.C.

1999-07-01

The Boh Environmental, LLC (BE) and Northwest Nuclear, LLC (NWN) program addresses the problem of diminishing capacity in the United States to store mixed waste. A lack of an alternative program has caused the US Department of Energy (DOE) to indefinitely store all of its mixed waste in Resource Conservation and Recovery Act (RCRA) compliant storage facilities. Unfortunately, this capacity is fast approaching the administrative control limit. The combination of unique BE encapsulation and NWN waste characterization technologies provides an effective solution to DOE's mixed-waste dilemma. The BE ARROW-PAK technique encapsulates mixed low-level waste (MLLW) in extra-high molecular weight, high-densitymore » polyethylene, pipe-grade resin cylinders. ARROW-PAK applications include waste treatment, disposal, transportation (per 49 CFR 173), vault encasement, and interim/long-term storage for 100 to 300 yr. One of the first demonstrations of this treatment/storage technique successfully treated 880 mixed-waste debris drums at the DOE Hanford Site in 1997. NWN, deploying the APNea neutron assay technology, provides the screening and characterization capability necessary to ensure that radioactive waste is correctly categorized as either transuranic (TRU) or LLW. MLLW resulting from D and D activities conducted at the Oak Ridge East Tennessee Technology Park will be placed into ARROW-PAK containers following comprehensive characterization of the waste by NWN. The characterized and encapsulated waste will then be shipped to a commercial disposal facility, where the shipments meet all waste acceptance criteria of the disposal facility including treatment criteria.« less

(Low-level radioactive waste management techniques)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Hoesen, S.D.; Kennerly, J.M.; Williams, L.C.

1988-08-08

The US team consisting of representatives of Oak Ridge National Laboratory (ORNL), Savannah River plant (SRP), Idaho National Engineering Laboratory (INEL), and the Department of Energy, Oak Ridge Operations participated in a training program on French low-level radioactive waste (LLW) management techniques. Training in the rigorous waste characterization, acceptance and certification procedures required in France was provided at Agence Nationale pour les Gestion des Dechets Radioactif (ANDRA) offices in Paris.

Bentonite-Clay Waste Form for the Immobilization of Cesium and Strontium from Fuel Processing Waste Streams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaminski, Michael D.; Mertz, Carol J.

2016-01-01

The physical properties of a surrogate waste form containing cesium, strontium, rubidium, and barium sintered into bentonite clay were evaluated for several simulant feed streams: chlorinated cobalt dicarbollide/polyethylene glycol (CCD-PEG) strip solution, nitrate salt, and chloride salt feeds. We sintered bentonite clay samples with a loading of 30 mass% of cesium, strontium, rubidium, and barium to a density of approximately 3 g/cm 3. Sintering temperatures of up to 1000°C did not result in volatility of cesium. Instead, there was an increase in crystallinity of the waste form upon sintering to 1000ºC for chloride- and nitrate-salt loaded clays. The nitrate saltmore » feed produced various cesium pollucite phases, while the chloride salt feed did not produce these familiar phases. In fact, many of the x-ray diffraction peaks could not be matched to known phases. Assemblages of silicates were formed that incorporated the Sr, Rb, and Ba ions. Gas evolution during sintering to 1000°C was significant (35% weight loss for the CCD-PEG waste-loaded clay), with significant water being evolved at approximately 600°C.« less

Safeguards and retrievability from waste forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Danker, W.

1996-05-01

This report describes issues discussed at a session from the PLutonium Stabilization and Immobilization Workshop related to safeguards and retrievability from waste forms. Throughout the discussion, the group probed the goals of disposition efforts, particularly an understanding of the {open_quotes}spent fuel standard{close_quotes}, since the disposition material form derives from these goals. The group felt strongly that not only the disposition goals but safeguards to meet these goals could affect the material form. Accordingly, the Department was encouraged to explore and apply safeguards as early in the implementation process as possible. It was emphasized that this was particularly true for anymore » planned use of existing facilities. It is much easier to build safeguards approaches into the development of new facilities, than to backfit existing facilities. Accordingly, special safeguards challenges are likely to be encountered, given the cost and schedule advantages offered by use of existing facilities.« less

Mineral assemblage transformation of a metakaolin-based waste form after geopolymer encapsulation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, Benjamin D.; Neeway, James J.; Snyder, Michelle M. V.

2016-05-01

Current plans for nuclear waste vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP) lack the capacity to treat all of the low activity waste (LAW) that is not encapsulated in the vitrified product. Fluidized Bed Steam Reforming (FBSR) is one of the supplemental technologies under consideration to fill this gap. The FBSR process results in a granular product mainly composed of feldspathoid mineral phases that encapsulate the LAW and other contaminants of concern (COCs). In order to better understand the characteristics of the FBSR product, characterization testing has been performed on the granular product as well asmore » the granular product encapsulated in a monolithic geopolymer binder. The non-radioactive simulated tank waste samples created for use in this study are the result of a 2008 Department of Energy sponsored Engineering Scale Technology Demonstration (ESTD) in 2008. These samples were created from waste simulant that was chemically shimmed to resemble actual tank waste, and rhenium has been used as a substitute for technetium. Another set of samples was created by the Savannah River Site Bench-Scale Reformer (BSR) using a chemical shim of Savannah River Site Tank 50 waste in order to simulate a blend of 68 Hanford tank wastes. This paper presents results from coal and moisture removal tests along with XRD, SEM, and BET analyses showing that the major mineral components are predominantly sodium aluminosilicate minerals and that the mineral product is highly porous. Results also show that the materials pass the short-term leach tests: the Toxicity Characteristic Leaching Procedure (TCLP) and Product Consistency Test (PCT).« less

Method for stabilizing low-level mixed wastes at room temperature

DOEpatents

Wagh, A.S.; Singh, D.

1997-07-08

A method to stabilize solid and liquid waste at room temperature is provided comprising combining solid waste with a starter oxide to obtain a powder, contacting the powder with an acid solution to create a slurry, said acid solution containing the liquid waste, shaping the now-mixed slurry into a predetermined form, and allowing the now-formed slurry to set. The invention also provides for a method to encapsulate and stabilize waste containing cesium comprising combining the waste with Zr(OH){sub 4} to create a solid-phase mixture, mixing phosphoric acid with the solid-phase mixture to create a slurry, subjecting the slurry to pressure; and allowing the now pressurized slurry to set. Lastly, the invention provides for a method to stabilize liquid waste, comprising supplying a powder containing magnesium, sodium and phosphate in predetermined proportions, mixing said powder with the liquid waste, such as tritium, and allowing the resulting slurry to set. 4 figs.

Method for stabilizing low-level mixed wastes at room temperature

DOEpatents

Wagh, Arun S.; Singh, Dileep

1997-01-01

A method to stabilize solid and liquid waste at room temperature is provided comprising combining solid waste with a starter oxide to obtain a powder, contacting the powder with an acid solution to create a slurry, said acid solution containing the liquid waste, shaping the now-mixed slurry into a predetermined form, and allowing the now-formed slurry to set. The invention also provides for a method to encapsulate and stabilize waste containing cesium comprising combining the waste with Zr(OH).sub.4 to create a solid-phase mixture, mixing phosphoric acid with the solid-phase mixture to create a slurry, subjecting the slurry to pressure; and allowing the now pressurized slurry to set. Lastly, the invention provides for a method to stabilize liquid waste, comprising supplying a powder containing magnesium, sodium and phosphate in predetermined proportions, mixing said powder with the liquid waste, such as tritium, and allowing the resulting slurry to set.

Environmental analysis burial of offsite low-level waste at SRP

NASA Astrophysics Data System (ADS)

Poe, W. L.; Moyer, R. A.

1980-12-01

The environmental effects of receipt and burial of low level naval waste are assessed. This low level waste was sent to the NRC-licensed burial ground operated by Chem-Nuclear Systems, Inc., at Barnwell, South Carolina. The DOE announced that DOE-generated low level waste would no longer be buried at commercial waste burial sites. The SRP was selected to receive the Naval waste described in this analysis. Receipt and burial of these wastes will have a negligible effect on SRP's environment and increase only sightly the environmental effects of the SRP operations discussed in the EIS on SRP waste management operations. The environmental effects of burial of this waste at Chem-Nuclear Burial Ground or at the SRP Burial Ground are described in this environmental analysis to permit assessment of incremental effects caused by the decision to bury this Naval waste in the SRP Burial Ground rather than in the Barnwell Burial Ground. The radiological effects from burial of this waste in either the SRP or Chem-Nuclear Burial Ground are very small when compared to those from natural background radiation or to the annual population dose commitment from operation of SRP. The environmental effects of burial at SRP to dose commitments normally received by the population surrounding SRP are compared.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, Xiangdong; Einziger, Robert E.

1997-01-01

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-08-12

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Process for immobilizing plutonium into vitreous ceramic waste forms

DOEpatents

Feng, X.; Einziger, R.E.

1997-01-28

Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Data quality objectives for TWRS privatization phase 1: confirm tank T is an appropriate feed source for low-activity waste feed batch X

DOE Office of Scientific and Technical Information (OSTI.GOV)

NGUYEN, D.M.

1999-06-01

The US. Department of Energy, Richland Operations Office (DOE-RL) has initiated Phase 1 of a two-phase privatization strategy for treatment and immobilization of low-activity waste (LAW) currently being managed by the Hanford Tank Waste Remediation System (TWRS) Project. In this strategy, DOE will purchase services from a contractor-owned and operated facility under a fixed price. The Phase 1 TWRS privatization contract requires that the Project Hanford Management Contract (PHMC) contractors, on behalf of DOE, deliver LAW feed in specified quantities and composition to the Privatization Contractor in a timely manner (DOE-RL 1996). Additional requirements are imposed by the interface controlmore » document (ICD-19) for LAW feed (PHMC 1997). In response to these requirements, the Tank Waste Remediation System Operation and Utilization Plan (TWRSO and UP) (Kirkbride et al. 1997) was prepared by the PHMC. The TWRSO and UP, as updated by the Readiness-To-Proceed deliverable (Payne et al. 1998), establishes the baseline operating scenario for the delivery of LAW feed to the Privatization Contractor. The scenario specifies tanks from which LAW will be provided for each feed batch, the operational activities needed to prepare and deliver each batch, and the timing of these activities. The operating scenario was developed based on current knowledge of waste composition and chemistry, waste transfer methods, and operating constraints, such as tank farm logistics and availability of tank space. A project master baseline schedule (PMBS) has been developed to implement the operating scenario. The PMBS also includes activities aimed at reducing programmatic risks. One of the activities, ''Confirm Plans and Requirements,'' was identified to verify the basis used to develop the scenario. Additional data on waste quantity, physical and chemical characteristics, and transfer properties will be needed to support this activity. This document describes the data quality objective (DQO

Method of making nanostructured glass-ceramic waste forms

DOEpatents

Gao, Huizhen; Wang, Yifeng; Rodriguez, Mark A.; Bencoe, Denise N.

2014-07-08

A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

Session 35 - Panel: Remaining US Disposition Issues for Orphan or Small Volume Low Level and Low Level Mixed Waste Streams

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blauvelt, Richard; Small, Ken; Gelles, Christine

2006-07-01

Faced with closure schedules as a driving force, significant progress has been made during the last 2 years on the disposition of DOE mixed waste streams thought previously to be problematic. Generators, the Department of Energy and commercial vendors have combined to develop unique disposition paths for former orphan streams. Recent successes and remaining issues will be discussed. The session will also provide an opportunity for Federal agencies to share lessons learned on low- level and mixed low-level waste challenges and identify opportunities for future collaboration. This panel discussion was organized by PAC member Dick Blauvelt, Navarro Research and Engineeringmore » Inc who served as co-chair along with Dave Eaton from INL. In addition, George Antonucci, Duratek Barnwell and Rich Conley, AFSC were invited members of the audience, prepared to contribute the Barnwell and DOD perspective to the issues as needed. Mr. Small provide information regarding the five year 20K M3 window of opportunity at the Nevada Test Site for DOE contractors to dispose of mixed waste that cannot be received at the Energy Solutions (Envirocare) site in Utah because of activity levels. He provided a summary of the waste acceptance criteria and the process sites must follow to be certified to ship. When the volume limit or time limit is met, the site will undergo a RCRA closure. Ms. Gelles summarized the status of the orphan issues, commercial options and the impact of the EM reorganization on her program. She also announced that there would be a follow-on meeting in 2006 to the very successful St. Louis meeting of last year. It will probably take place in Chicago in July. Details to be announced. Mr. McKenney discussed progress made at the Hanford Reservation regarding disposal of their mixed waste inventory. The news is good for the Hanford site but not good for the rest of the DOE complex since shipment for out of state of both low level and low level mixed waste will continue

Operational Strategies for Low-Level Radioactive Waste Disposal Site in Egypt - 13513

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mohamed, Yasser T.

The ultimate aims of treatment and conditioning is to prepare waste for disposal by ensuring that the waste will meet the waste acceptance criteria of a disposal facility. Hence the purpose of low-level waste disposal is to isolate the waste from both people and the environment. The radioactive particles in low-level waste emit the same types of radiation that everyone receives from nature. Most low-level waste fades away to natural background levels of radioactivity in months or years. Virtually all of it diminishes to natural levels in less than 300 years. In Egypt, The Hot Laboratories and Waste Management Centermore » has been established since 1983, as a waste management facility for LLW and ILW and the disposal site licensed for preoperational in 2005. The site accepts the low level waste generated on site and off site and unwanted radioactive sealed sources with half-life less than 30 years for disposal and all types of sources for interim storage prior to the final disposal. Operational requirements at the low-level (LLRW) disposal site are listed in the National Center for Nuclear Safety and Radiation Control NCNSRC guidelines. Additional procedures are listed in the Low-Level Radioactive Waste Disposal Facility Standards Manual. The following describes the current operations at the LLRW disposal site. (authors)« less

Candidate waste forms for immobilisation of waste chloride salt from pyroprocessing of spent nuclear fuel

NASA Astrophysics Data System (ADS)

Vance, E. R.; Davis, J.; Olufson, K.; Chironi, I.; Karatchevtseva, I.; Farnan, I.

2012-01-01

Sodalite/glass bodies prepared by hot isostatic pressing (HIPing) at ˜850 °C/100 MPa are candidates for immobilising fission product-bearing waste KCl-LiCl pyroprocessing salts. To study the capacity of sodalite to structurally incorporate such pyroprocessing salts, K, Li, Cs, Sr, Ba and La were individually targeted for substitution in a Na site in sodalite (Na vacancies targeted as charge compensators for alkaline and rare earths) and studied by X-ray diffraction and scanning electron microscopy after sintering in the range of 800-1000 °C. K and Li appeared to enter the sodalite, but Cs, Sr and Ba formed aluminosilicate phases and La formed an oxyapatite phase. However these non-sodalite phases have reasonable resistance to water leaching. Pure chlorapatite gives superior leach resistance to sodalite, and alkalis, alkaline and rare earth ions are generally known to enter chlorapatite, but attempts to incorporate simulated waste salt formulations into HIPed chlorapatite-based preparations or to substitute Cs alone into the structure of Ca-based chlorapatite were not successful on the basis of scanning electron microscopy. The materials exhibited severe water leachability, mainly in regard to Cs release. Attempts to substitute Cs into Ba- and Sr-based chlorapatites also did not look encouraging. Consequently the use of apatite alone to retain fission product-bearing waste pyroprocessing salts from electrolytic nuclear fuel reprocessing is problematical, but chlorapatite glass-ceramics may be feasible, albeit with reduced waste loadings. Spodiosite, Ca 2(PO 4)Cl, does not appear to be suitable for incorporation of Cl-bearing waste containing fission products.

Method for calcining radioactive wastes

DOEpatents

Bjorklund, William J.; McElroy, Jack L.; Mendel, John E.

1979-01-01

This invention relates to a method for the preparation of radioactive wastes in a low leachability form by calcining the radioactive waste on a fluidized bed of glass frit, removing the calcined waste to melter to form a homogeneous melt of the glass and the calcined waste, and then solidifying the melt to encapsulate the radioactive calcine in a glass matrix.

Controlling changes - lessons learned from waste management facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Johnson, B.M.; Koplow, A.S.; Stoll, F.E.

This paper discusses lessons learned about change control at the Waste Reduction Operations Complex (WROC) and Waste Experimental Reduction Facility (WERF) of the Idaho National Engineering Laboratory (INEL). WROC and WERF have developed and implemented change control and an as-built drawing process and have identified structures, systems, and components (SSCS) for configuration management. The operations have also formed an Independent Review Committee to minimize costs and resources associated with changing documents. WROC and WERF perform waste management activities at the INEL. WROC activities include storage, treatment, and disposal of hazardous and mixed waste. WERF provides volume reduction of solid low-levelmore » waste through compaction, incineration, and sizing operations. WROC and WERF`s efforts aim to improve change control processes that have worked inefficiently in the past.« less

Thirteenth annual U.S. DOE low-level radioactive waste management conference: Proceedings

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1991-12-31

The 40 papers in this document comprise the proceedings of the Department of Energy`s Thirteenth Annual Low-Level Radioactive Waste Management Conference that was held in Atlanta, Georgia, on November 19--21, 1991. General subjects addressed during the conference included: disposal facility design; greater-than-class C low-level waste; public acceptance considerations; waste certification; site characterization; performance assessment; licensing and documentation; emerging low-level waste technologies; waste minimization; mixed waste; tracking and transportation; storage; and regulatory changes. Papers have been processed separately for inclusion on the data base.

Waste Form and Indrift Colloids-Associated Radionuclide Concentrations: Abstraction and Summary

DOE Office of Scientific and Technical Information (OSTI.GOV)

R. Aguilar

This Model Report describes the analysis and abstractions of the colloids process model for the waste form and engineered barrier system components of the total system performance assessment calculations to be performed with the Total System Performance Assessment-License Application model. Included in this report is a description of (1) the types and concentrations of colloids that could be generated in the waste package from degradation of waste forms and the corrosion of the waste package materials, (2) types and concentrations of colloids produced from the steel components of the repository and their potential role in radionuclide transport, and (3) typesmore » and concentrations of colloids present in natural waters in the vicinity of Yucca Mountain. Additionally, attachment/detachment characteristics and mechanisms of colloids anticipated in the repository are addressed and discussed. The abstraction of the process model is intended to capture the most important characteristics of radionuclide-colloid behavior for use in predicting the potential impact of colloid-facilitated radionuclide transport on repository performance.« less

Stabilization of NaCl-containing cuttings wastes in cement concrete by in situ formed mineral phases.

PubMed

Filippov, Lev; Thomas, Fabien; Filippova, Inna; Yvon, Jacques; Morillon-Jeanmaire, Anne

2009-11-15

Disposal of NaCl-containing cuttings is a major environmental concern due to the high solubility of chlorides. The present work aims at reducing the solubility of chloride by encapsulation in low permeability matrix as well as lowering its solubility by trapping into low-solubility phases. Both the studied materials were cuttings from an oil-based mud in oil drillings containing about 50% of halite, and cuttings in water-based mud from gas drilling containing 90% of halite. A reduction in the amount of dissolved salt from 41 to 19% according to normalized leaching tests was obtained by addition of potassium ortho-phosphate in the mortar formula of oil-based cuttings, while the aluminium dihydrogeno-phosphate is even more efficient for the stabilization of water-based cuttings with a NaCl content of 90%. Addition of ortho-phosphate leads to form a continuous and weakly soluble network in the cement matrix, which reduces the release of salt. The formed mineralogical phases were apatite and hydrocalumite. These phases encapsulate the salt grains within a network, thus lowering its interaction with water or/and trap chloride into low-solubility phases. The tested approaches allow to develop a confinement process of NaCl-containing waste of various compositions that can be applied to wastes, whatever the salt content and the nature of the drilling fluids (water or oil).

Characterization of the solid low level mixed waste inventory for the solid waste thermal treatment activity - III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Place, B.G., Westinghouse Hanford

1996-09-24

The existing thermally treatable, radioactive mixed waste inventory is characterized to support implementation of the commercial, 1214 thermal treatment contract. The existing thermally treatable waste inventory has been identified using a decision matrix developed by Josephson et al. (1996). Similar to earlier waste characterization reports (Place 1993 and 1994), hazardous materials, radionuclides, physical properties, and waste container data are statistically analyzed. In addition, the waste inventory data is analyzed to correlate waste constituent data that are important to the implementation of the commercial thermal treatment contract for obtaining permits and for process design. The specific waste parameters, which were analyzed,more » include the following: ``dose equivalent`` curie content, polychlorinated biphenyl (PCB) content, identification of containers with PA-related mobile radionuclides (14C, 12 79Se, 99Tc, and U isotopes), tritium content, debris and non-debris content, container free liquid content, fissile isotope content, identification of dangerous waste codes, asbestos containers, high mercury containers, beryllium dust containers, lead containers, overall waste quantities, analysis of container types, and an estimate of the waste compositional split based on the thermal treatment contractor`s proposed process. A qualitative description of the thermally treatable mixed waste inventory is also provided.« less

Technical viability and development needs for waste forms and facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pegg, I.; Gould, T.

1996-05-01

The objective of this breakout session was to provide a forum to discuss technical issues relating to plutonium-bearing waste forms and their disposal facilities. Specific topics for discussion included the technical viability and development needs associated with the waste forms and/or disposal facilities. The expected end result of the session was an in-depth (so far as the limited time would allow) discussion of key issues by the session participants. The session chairs expressed allowance for, and encouragement of, alternative points of view, as well as encouragement for discussion of any relevant topics not addressed in the paper presentations. It wasmore » not the intent of this session to recommend or advocate any one technology over another.« less

NDA issues with RFETS vitrified waste forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hurd, J.; Veazey, G.

1998-12-31

A study was conducted at Los Alamos National Laboratory (LANL) for the purpose of determining the feasibility of using a segmented gamma scanner (SGS) to accurately perform non-destructive analysis (NDA) on certain Rocky Flats Environmental Technology Site (RFETS) vitrified waste samples. This study was performed on a full-scale vitrified ash sample prepared at LANL according to a procedure similar to that anticipated to be used at RFETS. This sample was composed of a borosilicate-based glass frit, blended with ash to produce a Pu content of {approximately}1 wt %. The glass frit was taken to a degree of melting necessary tomore » achieve a full encapsulation of the ash material. The NDA study performed on this sample showed that SGSs with either {1/2}- or 2-inch collimation can achieve an accuracy better than 6 % relative to calorimetry and {gamma}-ray isotopics. This accuracy is achievable, after application of appropriate bias corrections, for transmissions of about {1/2} % through the waste form and counting times of less than 30 minutes. These results are valid for ash material and graphite fines with the same degree of plutonium particle size, homogeneity, sample density, and sample geometry as the waste form used to obtain the results in this study. A drum-sized thermal neutron counter (TNC) was also included in the study to provide an alternative in the event the SGS failed to meet the required level of accuracy. The preliminary indications are that this method will also achieve the required accuracy with counting times of {approximately}30 minutes and appropriate application of bias corrections. The bias corrections can be avoided in all cases if the instruments are calibrated on standards matching the items.« less

Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, William J.; Zhang, Yanwen

This is the final report of the NEUP project “Radiation and Thermal Effects on Used Nuclear Fuel and Nuclear Waste Forms.” This project started on July 1, 2012 and was successfully completed on June 30, 2016. This report provides an overview of the main achievements, results and findings through the duration of the project. Additional details can be found in the main body of this report and in the individual Quarterly Reports and associated Deliverables of this project, which have been uploaded in PICS-NE. The objective of this research was to advance understanding and develop validated models on the effectsmore » of self-radiation from beta and alpha decay on the response of used nuclear fuel and nuclear waste forms during high-temperature interim storage and long-term permanent disposition. To achieve this objective, model used-fuel materials and model waste form materials were identified, fabricated, and studied.« less

77 FR 58591 - Report on Waste Burial Charges: Changes in Decommissioning Waste Disposal Costs at Low-Level...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-09-21

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0362] Report on Waste Burial Charges: Changes in Decommissioning Waste Disposal Costs at Low-Level Waste Burial Facilities AGENCY: Nuclear Regulatory Commission... Commission) has issued for public comment a document entitled: NUREG-1307 Revision 15, ``Report on Waste...

Waste heat recovery on multiple low-speed reciprocating engines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mayhew, R.E.

1982-09-01

With rising fuel costs, energy conservation has taken on added significance. Installation of Waste Heat Recovery Units (WHRU) on gas turbines is one method used in the past to reduce gas plant fuel consumption. More recently, waste heat recovery on multiple reciprocating compressor engines has also been identified as having energy conservation potential. This paper reviews the development and implementation of a Waste Heat Recovery Unit (WHRU) for multiple low speed engines at the Katy Gas Plant. WHRU's for these engines should be differentiated from high speed engines and gas turbines in that low speed engines produce low frequency, highmore » amplitude pulsating exhaust. The design of a waste heat system must take this potentially destructive pulsation into account. At Katy, the pulsation forces were measured at high amplitude frequencies and then used to design structural stiffness into the various components of the WHRU to minimize vibration and improve system reliability.« less

Experimental Design for Hanford Low-Activity Waste Glasses with High Waste Loading

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piepel, Gregory F.; Cooley, Scott K.; Vienna, John D.

This report discusses the development of an experimental design for the initial phase of the Hanford low-activity waste (LAW) enhanced glass study. This report is based on a manuscript written for an applied statistics journal. Appendices A, B, and E include additional information relevant to the LAW enhanced glass experimental design that is not included in the journal manuscript. The glass composition experimental region is defined by single-component constraints (SCCs), linear multiple-component constraints (MCCs), and a nonlinear MCC involving 15 LAW glass components. Traditional methods and software for designing constrained mixture experiments with SCCs and linear MCCs are not directlymore » applicable because of the nonlinear MCC. A modification of existing methodology to account for the nonlinear MCC was developed and is described in this report. One of the glass components, SO 3, has a solubility limit in glass that depends on the composition of the balance of the glass. A goal was to design the experiment so that SO 3 would not exceed its predicted solubility limit for any of the experimental glasses. The SO 3 solubility limit had previously been modeled by a partial quadratic mixture model expressed in the relative proportions of the 14 other components. The partial quadratic mixture model was used to construct a nonlinear MCC in terms of all 15 components. In addition, there were SCCs and linear MCCs. This report describes how a layered design was generated to (i) account for the SCCs, linear MCCs, and nonlinear MCC and (ii) meet the goals of the study. A layered design consists of points on an outer layer, and inner layer, and a center point. There were 18 outer-layer glasses chosen using optimal experimental design software to augment 147 existing glass compositions that were within the LAW glass composition experimental region. Then 13 inner-layer glasses were chosen with the software to augment the existing and outer-layer glasses. The

Thermoelectric harvesting of low temperature natural/waste heat

NASA Astrophysics Data System (ADS)

Rowe, David Michael

2012-06-01

Apart from specialized space requirements current development in applications of thermoelectric generation mainly relate to reducing harmful carbon emissions and decreasing costly fuel consumption through the recovery of exhaust heat from fossil fuel powered engines and emissions from industrial utilities. Focus on these applications is to the detriment of the wider exploitations of thermoelectrics with other sources of heat energy, and in particular natural occurring and waste low temperature heat, receiving little, if any, attention. In this presentation thermoelectric generation applications, both potential and real in harvesting low temperature waste/natural heat are reviewed. The use of thermoelectrics to harvest solar energy, ocean thermal energy, geothermal heat and waste heat are discussed and their credibility as future large-scale sources of electrical power assessed.

Technical area status report for waste destruction and stabilization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dalton, J.D.; Harris, T.L.; DeWitt, L.M.

1993-08-01

The Office of Environmental Restoration and Waste Management (EM) was established by the Department of Energy (DOE) to direct and coordinate waste management and site remediation programs/activities throughout the DOE complex. In order to successfully achieve the goal of properly managing waste and the cleanup of the DOE sites, the EM was divided into five organizations: the Office of Planning and Resource Management (EM-10); the Office of Environmental Quality Assurance and Resource Management (EM-20); the Office of Waste Operations (EM-30); the Office of Environmental Restoration (EM-40); and the Office of Technology and Development (EM-50). The mission of the Office ofmore » Technology Development (OTD) is to develop treatment technologies for DOE`s operational and environmental restoration wastes where current treatment technologies are inadequate or not available. The Mixed Waste Integrated Program (MWIP) was created by OTD to assist in the development of treatment technologies for the DOE mixed low-level wastes (MLLW). The MWIP has established five Technical Support Groups (TSGs) whose purpose is to identify, evaluate, and develop treatment technologies within five general technical areas representing waste treatment functions from initial waste handling through generation of final waste forms. These TSGs are: (1) Front-End Waste Handling, (2) Physical/Chemical Treatment, (3) Waste Destruction and Stabilization, (4) Second-Stage Destruction and Offgas Treatment, and (5) Final Waste Forms. This report describes the functions of the Waste Destruction and Stabilization (WDS) group. Specifically, the following items are discussed: DOE waste stream identification; summary of previous efforts; summary of WDS treatment technologies; currently funded WDS activities; and recommendations for future activities.« less

A Strategy to Conduct an Analysis of the Long-Term Performance of Low-Activity Waste Glass in a Shallow Subsurface Disposal System at Hanford

DOE Office of Scientific and Technical Information (OSTI.GOV)

BP McGrail, WL Ebert, DH Bacon, DM Strachan

1998-02-18

Privatized services are being procured to vitrify low-activity tank wastes for eventual disposal in a shallow subsurface facility at the Hanford Site. Over 500,000 metric tons of low-activity waste glass will be generated, which is among the largest volumes of waste within the U.S. Department of Energy (DOE) complex and is one of the largest inventories of long-lived radionuclides planned for disposal in a low-level waste facility. Before immobilized waste can be disposed, DOE must approve a "performance assessment," which is a document that describes the impacts of the disposal facility on public health and environmental resources. Because the releasemore » rate of radionuclides from the glass waste form is a key factor determining these impacts, a sound scientific basis for determining their long-term release rates must be developed if this disposal action is to be accepted by regulatory agencies, stakeholders, and the public. In part, the scientific basis is determined from a sound testing strategy. The foundation of the proposed testing strategy is a well accepted mechanistic model that is being used to calculate the glass corrosion behavior over the geologic time scales required for performance assessment. This model requires that six parameters be determined, and the testing program is defined by an appropriate set of laboratory experiments to determine these parameters, and is combined with a set of field experiments to validate the model as a whole. Three general classes of laboratory tests are proposed in this strategy: 1) characterization, 2) accelerated, and 3) service condition. Characterization tests isolate and provide specific information about processes or parameters in theoretical models. Accelerated tests investigate corrosion behavior that will be important over the regulated service life of a disposal system within a laboratory time frame of a few years or less. Service condition tests verify that the techniques used in accelerated tests do

Waste minimization for commercial radioactive materials users generating low-level radioactive waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fischer, D.K.; Gitt, M.; Williams, G.A.

1991-07-01

The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations.more » This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature.« less

77 FR 25760 - Low-Level Radioactive Waste Management and Volume Reduction

Federal Register 2010, 2011, 2012, 2013, 2014

2012-05-01

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0183] Low-Level Radioactive Waste Management and Volume.... Nuclear Regulatory Commission (NRC or the Commission) is revising its 1981 Policy Statement on Low-Level..., the NRC staff issued SECY-10-0043, ``Blending of Low-Level Radioactive Waste'' (ADAMS Accession No...

Corrosion Behavior and Microstructure Influence of Glass-Ceramic Nuclear Waste Forms

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew Asmussen, R.; Neeway, James J.; Kaspar, Tiffany C.

Glass ceramic waste forms present a potentially viable technology for the long term immobilization and disposal of liquid nuclear wastes. Through control of chemistry during fabrication, such waste forms can have designed secondary crystalline phases within a borosilicate glass matrix. In this work, a glass ceramic containing powellite and oxyapatite secondary phases was tested for its corrosion properties in dilute conditions using single pass flow through testing (SPFT). Three glass ceramic samples were prepared using different cooling rates to produce samples with varying microstructure sizes. In testing at 90 °C in buffered pH 7 and pH 9 solutions, it wasmore » found that increasing pH and decreasing microstructure size (resulting from rapid cooling during fabrication) both led to a reduction in overall corrosion rate. The phases of the glass ceramic were found, using a combination of solutions analysis, SEM and AFM, to corrode preferably in the order of powellite > bulk glass matrix > oxyapatite.« less

Geohydrologic aspects for siting and design of low-level radioactive-waste disposal

USGS Publications Warehouse

Bedinger, M.S.

1989-01-01

The objective for siting and design of low-level radioactive-waste repository sites is to isolate the waste from the biosphere until the waste no longer poses an unacceptable hazard as a result of radioactive decay. Low-level radioactive waste commonly is isolated at shallow depths with various engineered features to stabilize the waste and to reduce its dissolution and transport by ground water. The unsaturated zone generally is preferred for isolating the waste. Low-level radioactive waste may need to be isolated for 300 to 500 years. Maintenance and monitoring of the repository site are required by Federal regulations for only the first 100 years. Therefore, geohydrology of the repository site needs to provide natural isolation of the waste for the hazardous period following maintenance of the site. Engineering design of the repository needs to be compatible with the natural geohydrologic conditions at the site. Studies at existing commercial and Federal waste-disposal sites provide information on the problems encountered and the basis for establishing siting guidelines for improved isolation of radioactive waste, engineering design of repository structures, and surveillance needs to assess the effectiveness of the repositories and to provide early warning of problems that may require remedial action.Climate directly affects the hydrology of a site and probably is the most important single factor that affects the suitability of a site for shallow-land burial of low-level radioactive waste. Humid and subhumid regions are not well suited for shallow isolation of low-level radioactive waste in the unsaturated zone; arid regions with zero to small infiltration from precipitation, great depths to the water table, and long flow paths to natural discharge areas are naturally well suited to isolation of the waste. The unsaturated zone is preferred for isolation of low-level radioactive waste. The guiding rationale is to minimize contact of water with the waste and to

Management of Low-Level Radioactive Waste from Research, Hospitals and Nuclear Medical Centers in Egypt - 13469

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hasan, M.A.; Selim, Y.T.; Lasheen, Y.F.

2013-07-01

The application of radioisotopes and radiation sources in medical diagnosis and therapy is an important issue. Physicians can use radioisotopes to diagnose and treat diseases. Methods of treatment, conditioning and management of low level radioactive wastes from the use of radiation sources and radioisotopes in hospitals and nuclear medicine application, are described. Solid Radioactive waste with low-level activity after accumulation, minimization, segregation and measurement, are burned or compressed in a compactor according to the international standards. Conditioned drums are transported to the interim storage site at the Egyptian Atomic Energy Authority (EAEA) represented in Hot Labs and Waste Management Centermore » (HLWMC) for storage and monitoring. (authors)« less

(Low-level waste disposal facility siting and site characterization)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mezga, L.J.; Ketelle, R.H.; Pin, F.G.

A US team consisting of representatives of Oak Ridge National Laboratory (ORNL), Savannah River Plant (SRP), Savannah river Laboratory (SRL), and the Department of Energy Office of Defense Waste and Byproducts Management participated in the fourth meeting held under the US/French Radioactive Waste Management Agreement between the US Department of Energy and the Commissariat a l'Energie Atomique. This meeting, held at Agence Nationale pour les Gestion des Dechets Radioactifs' (ANDRA's) Headquarters in Paris, was a detailed, technical topical workshop focusing on Low-Level Waste Disposal Facility Siting and Site Characterization.'' The meeting also included a visit to the Centre de lamore » Manche waste management facility operated by ANDRA to discuss and observe the French approach to low-level waste management. The final day of the meeting was spent at the offices of Societe Generale pour les Techniques Nouvelles (SGN) discussing potential areas of future cooperation and exchange. 20 figs.« less

Evaluation of final waste forms and recommendations for baseline alternatives to group and glass

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bleier, A.

1997-09-01

An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT&E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidatemore » alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT&E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the remaining

Trivalent chromium removal from wastewater using low cost activated carbon derived from agricultural waste material and activated carbon fabric cloth.

PubMed

Mohan, Dinesh; Singh, Kunwar P; Singh, Vinod K

2006-07-31

An efficient adsorption process is developed for the decontamination of trivalent chromium from tannery effluents. A low cost activated carbon (ATFAC) was prepared from coconut shell fibers (an agricultural waste), characterized and utilized for Cr(III) removal from water/wastewater. A commercially available activated carbon fabric cloth (ACF) was also studied for comparative evaluation. All the equilibrium and kinetic studies were conducted at different temperatures, particle size, pHs, and adsorbent doses in batch mode. The Langmuir and Freundlich isotherm models were applied. The Langmuir model best fit the equilibrium isotherm data. The maximum adsorption capacities of ATFAC and ACF at 25 degrees C are 12.2 and 39.56 mg/g, respectively. Cr(III) adsorption increased with an increase in temperature (10 degrees C: ATFAC--10.97 mg/g, ACF--36.05 mg/g; 40 degrees C: ATFAC--16.10 mg/g, ACF--40.29 mg/g). The kinetic studies were conducted to delineate the effect of temperature, initial adsorbate concentration, particle size of the adsorbent, and solid to liquid ratio. The adsorption of Cr(III) follows the pseudo-second-order rate kinetics. From kinetic studies various rate and thermodynamic parameters such as effective diffusion coefficient, activation energy and entropy of activation were evaluated. The sorption capacity of activated carbon (ATFAC) and activated carbon fabric cloth is comparable to many other adsorbents/carbons/biosorbents utilized for the removal of trivalent chromium from water/wastewater.

Radioactive Waste Management Complex low-level waste radiological performance assessment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maheras, S.J.; Rood, A.S.; Magnuson, S.O.

This report documents the projected radiological dose impacts associated with the disposal of radioactive low-level waste at the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. This radiological performance assessment was conducted to evaluate compliance with applicable radiological criteria of the US Department of Energy and the US Environmental Protection Agency for protection of the public and the environment. The calculations involved modeling the transport of radionuclides from buried waste, to surface soil and subsurface media, and eventually to members of the public via air, groundwater, and food chain pathways. Projections of doses were made for both offsitemore » receptors and individuals inadvertently intruding onto the site after closure. In addition, uncertainty and sensitivity analyses were performed. The results of the analyses indicate compliance with established radiological criteria and provide reasonable assurance that public health and safety will be protected.« less

Decision Support System For Management Of Low-Level Radioactive Waste Disposal At The Nevada Test Site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shott, G.; Yucel, V.; Desotell, L.

2006-07-01

The long-term safety of U.S. Department of Energy (DOE) low-level radioactive disposal facilities is assessed by conducting a performance assessment -- a systematic analysis that compares estimated risks to the public and the environment with performance objectives contained in DOE Manual 435.1-1, Radioactive Waste Management Manual. Before site operations, facilities design features such as final inventory, waste form characteristics, and closure cover design may be uncertain. Site operators need a modeling tool that can be used throughout the operational life of the disposal site to guide decisions regarding the acceptance of problematic waste streams, new disposal cell design, environmental monitoringmore » program design, and final site closure. In response to these needs the National Nuclear Security Administration Nevada Site Office (NNSA/NSO) has developed a decision support system for the Area 5 Radioactive Waste Management Site in Frenchman Flat on the Nevada Test Site. The core of the system is a probabilistic inventory and performance assessment model implemented in the GoldSim{sup R} simulation platform. The modeling platform supports multiple graphic capabilities that allow clear documentation of the model data sources, conceptual model, mathematical implementation, and results. The combined models have the capability to estimate disposal site inventory, contaminant concentrations in environmental media, and radiological doses to members of the public engaged in various activities at multiple locations. The model allows rapid assessment and documentation of the consequences of waste management decisions using the most current site characterization information, radionuclide inventory, and conceptual model. The model is routinely used to provide annual updates of site performance, evaluate the consequences of disposal of new waste streams, develop waste concentration limits, optimize the design of new disposal cells, and assess the adequacy of

The impact of anaerobic microorganisms activities in ruminant waste and coal

NASA Astrophysics Data System (ADS)

Harlia, Ellin; Hamdani, H.; Winantris, Kurnani, Tb. B. A.; Hidayati, Y. A.; Marlina, E. T.; Rahmah, K. N.; Arief, H.; Ridwan, R.; Joni, I. M.

2018-02-01

Ruminant (dairy cattle, beef cattle and buffalo) waste from intensive farming concentrated in highly populated areas when stacked and accumulated in certain heights and in anaerobic condition, may produce Green House Gases (GHGs) which lead to global warming. This condition is generated through fermentation by microorganism contained in livestock waste and biogenic activities on coal. The GHGs include CH4 (methane), CO2 (carbon dioxide) and N2O (nitrous oxide). The GHG emission should be early monitored to minimize greater problems. In the other hand, methane can be utilized as an environmental friendly energy after stored as biogas on digester. The aim of this research is to detect how much GHGs formed from ruminant waste and biogenic activities on coal, which can be utilized as an alternative energy. This research conducted as an explorative study utilizing dairy cattle feces, beef cattle feces, buffalo feces and three types of coal: lignite, bituminous and sub-bituminous, which is separately added into medium 98-5 made from mixture of agar medium and chemical components in powder and crystal form diluted with distilled water and rumen liquid, with six repetitions. Each sample was stored into 250 mL anaerobic digester, observed weekly for period of 4 weeks, analyzed by Gas Chromatography (GC-A14). The result showed that GHGs: CH4, CO2 and N2O were found in all samples. Anticipation of GHGs formation to avoid air pollution is by utilizing livestock waste and coal in aerobic condition or in anaerobic condition through digester.

Biological intrusion of low-level-waste trench covers

NASA Astrophysics Data System (ADS)

Hakonson, T. E.; Gladney, E. S.

The long-term integrity of low-level waste shallow land burialsites is dependent on the interaction of physical, chemical, and biological factors that modify the waste containment system. The need to consider biological processes as being potentially important in reducing the integrity of waste burial site cover treatment is demonstrated. One approach to limiting biological intrusion through the waste cover is to apply a barrier within the profile to limit root and animal penetration with depth. Experiments in the Los Alamos Experimental Engineered Test Facility were initiated to develop and evaluate biological barriers that are effective in minimizing intrusion into waste trenches. The experiments that are described employ four different candidate barrier materials of geologic origin. Experimental variables that will be evaluated, in addition to barrier type, are barrier depth and sil overburden depth.

Ultrasonic waste activated sludge disintegration for improving anaerobic stabilization.

PubMed

Tiehm, A; Nickel, K; Zellhorn, M; Neis, U

2001-06-01

The pretreatment of waste activated sludge by ultrasonic disintegration was studied in order to improve the anaerobic sludge stabilization. The ultrasound frequency was varied within a range from 41 to 3217 kHz. The impact of different ultrasound intensities and treatment times was examined. Sludge disintegration was most significant at low frequencies. Low-frequency ultrasound creates large cavitation bubbles which upon collapse initiate powerful jet streams exerting strong shear forces in the liquid. The decreasing sludge disintegration efficiency observed at higher frequencies was attributed to smaller cavitation bubbles which do not allow the initiation of such strong shear forces. Short sonication times resulted in sludge floc deagglomeration without the destruction of bacteria cells. Longer sonication brought about the break-up of cell walls, the sludge solids were distintegrated and dissolved organic compounds were released. The anaerobic digestion of waste activated sludge following ultrasonic pretreatment causing microbial cell lysis was significantly improved. There was an increase in the volatile solids degradation as well as an increase in the biogas production. The increase in digestion efficiency was proportional to the degree of sludge disintegration. To a lesser degree the deagglomeration of sludge flocs also augmented the anaerobic volatile solids degradation.

40 CFR 61.154 - Standard for active waste disposal sites.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 40 Protection of Environment 8 2011-07-01 2011-07-01 false Standard for active waste disposal... for Asbestos § 61.154 Standard for active waste disposal sites. Each owner or operator of an active... visible emissions to the outside air from any active waste disposal site where asbestos-containing waste...

40 CFR 61.154 - Standard for active waste disposal sites.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 40 Protection of Environment 9 2013-07-01 2013-07-01 false Standard for active waste disposal... for Asbestos § 61.154 Standard for active waste disposal sites. Each owner or operator of an active... visible emissions to the outside air from any active waste disposal site where asbestos-containing waste...

40 CFR 61.154 - Standard for active waste disposal sites.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 40 Protection of Environment 9 2012-07-01 2012-07-01 false Standard for active waste disposal... for Asbestos § 61.154 Standard for active waste disposal sites. Each owner or operator of an active... visible emissions to the outside air from any active waste disposal site where asbestos-containing waste...

40 CFR 61.154 - Standard for active waste disposal sites.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 40 Protection of Environment 8 2010-07-01 2010-07-01 false Standard for active waste disposal... for Asbestos § 61.154 Standard for active waste disposal sites. Each owner or operator of an active... visible emissions to the outside air from any active waste disposal site where asbestos-containing waste...

40 CFR 61.154 - Standard for active waste disposal sites.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 40 Protection of Environment 9 2014-07-01 2014-07-01 false Standard for active waste disposal... for Asbestos § 61.154 Standard for active waste disposal sites. Each owner or operator of an active... visible emissions to the outside air from any active waste disposal site where asbestos-containing waste...

Phase-Pure and Multiphase Ceramic Waste Forms: Microstructure Evolution and Cesium Immobilization

NASA Astrophysics Data System (ADS)

Tumurugoti, Priyatham

Efforts of this thesis are directed towards developing ceramic waste forms as a potential replacement for the conventional glass waste forms for the safe immobilization and disposal of nuclear wastes from the legacy weapons programs as well as commercial power production. The body of this work consists of two equal parts with first focused on multiphase waste form containing hollandite as major phase and the later, on single-phase hollandites for Cs incorporation. Part I: Multiphase waste forms:. Hollandite-rich multiphase waste form compositions processed by melt-solidification and spark plasma sintering (SPS) were characterized, compared, and validated for nuclear waste incorporation. Phase identification by X-ray diffraction (XRD) and electron back-scattered diffraction (EBSD) confirm hollandite as the major phase present in these samples along with perovskite, pyrochlore and zirconolite. Distribution of select elements observed by wavelength dispersive spectroscopy (WDS) maps indicate that Cs forms a secondary phase during SPS processing, which is considered undesirable. On the other hand Cs partitioned into hollandite phase in melt-processed samples. Further analysis of hollandite structure in melt-processed composition, by selected area electron diffraction (SAED), reveals ordered arrangement of tunnel ions (Ba/Cs) and vacancies, suggesting efficient Cs incorporation into the lattice. Following the microstructural analysis, the crystallization behavior of the multiphase composition during melt-processing was studied. The phase assemblage and evolution of hollandite, zirconolite, pyrochlore, and perovskite type structures during melt processing were studied using thermal analysis, in-situ XRD, and scanning electron microscopy (SEM). Samples prepared by melting followed by annealing and quenching were analyzed to determine and measure the progression of the phase assemblage. Samples were melted at 1500°C and heat-treated at crystallization temperatures of 1285

Low-level radioactive waste technology: a selected, annotated bibliography

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fore, C.S.; Vaughan, N.D.; Hyder, L.K.

1980-10-01

This annotated bibliography of 447 references contains scientific, technical, economic, and regulatory information relevant to low-level radioactive waste technology. The bibliography focuses on environmental transport, disposal site, and waste treatment studies. The publication covers both domestic and foreign literature for the period 1952 to 1979. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Environmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated into the data file to improve the ease and accuracy of locating pertinentmore » references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. In addition, each document referenced in this bibliography has been assigned a relevance number to facilitate sorting the documents according to their pertinence to low-level radioactive waste technology. The documents are rated 1, 2, 3, or 4, with 1 indicating direct applicability to low-level radioactive waste technology and 4 indicating that a considerable amount of interpretation is required for the information presented to be applied. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. Indexes are provide for (1) author(s), (2) keywords, (3) subject category, (4) title, (5) geographic location, (6) measured parameters, (7) measured radionuclides, and (8) publication description.« less

Chemical digestion of low level nuclear solid waste material

DOEpatents

Cooley, Carl R.; Lerch, Ronald E.

1976-01-01

A chemical digestion for treatment of low level combustible nuclear solid waste material is provided and comprises reacting the solid waste material with concentrated sulfuric acid at a temperature within the range of 230.degree.-300.degree.C and simultaneously and/or thereafter contacting the reacting mixture with concentrated nitric acid or nitrogen dioxide. In a special embodiment spent ion exchange resins are converted by this chemical digestion to noncombustible gases and a low volume noncombustible residue.

Glass Waste Forms for Oak Ridge Tank Wastes: Fiscal Year 1998 Report for Task Plan SR-16WT-31, Task B

DOE Office of Scientific and Technical Information (OSTI.GOV)

Andrews, M.K.

1999-05-10

Using ORNL information on the characterization of the tank waste sludges, SRTC performed extensive bench-scale vitrification studies using simulants. Several glass systems were tested to ensure the optimum glass composition (based on the glass liquidus temperature, viscosity and durability) is determined. This optimum composition will balance waste loading, melt temperature, waste form performance and disposal requirements. By optimizing the glass composition, a cost savings can be realized during vitrification of the waste. The preferred glass formulation was selected from the bench-scale studies and recommended to ORNL for further testing with samples of actual OR waste tank sludges.

Closed Fuel Cycle Waste Treatment Strategy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vienna, J. D.; Collins, E. D.; Crum, J. V.

This study is aimed at evaluating the existing waste management approaches for nuclear fuel cycle facilities in comparison to the objectives of implementing an advanced fuel cycle in the U.S. under current legal, regulatory, and logistical constructs. The study begins with the Global Nuclear Energy Partnership (GNEP) Integrated Waste Management Strategy (IWMS) (Gombert et al. 2008) as a general strategy and associated Waste Treatment Baseline Study (WTBS) (Gombert et al. 2007). The tenets of the IWMS are equally valid to the current waste management study. However, the flowsheet details have changed significantly from those considered under GNEP. In addition, significantmore » additional waste management technology development has occurred since the GNEP waste management studies were performed. This study updates the information found in the WTBS, summarizes the results of more recent technology development efforts, and describes waste management approaches as they apply to a representative full recycle reprocessing flowsheet. Many of the waste management technologies discussed also apply to other potential flowsheets that involve reprocessing. These applications are occasionally discussed where the data are more readily available. The report summarizes the waste arising from aqueous reprocessing of a typical light-water reactor (LWR) fuel to separate actinides for use in fabricating metal sodium fast reactor (SFR) fuel and from electrochemical reprocessing of the metal SFR fuel to separate actinides for recycle back into the SFR in the form of metal fuel. The primary streams considered and the recommended waste forms include; Tritium in low-water cement in high integrity containers (HICs); Iodine-129: As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica

Biochemical process of low level radioactive liquid simulation waste containing detergent

NASA Astrophysics Data System (ADS)

Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

2015-12-01

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10-5 Ci/m3. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod's model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour-1.

Biochemical process of low level radioactive liquid simulation waste containing detergent

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kundari, Noor Anis, E-mail: nooranis@batan.go.id; Putra, Sugili; Mukaromah, Umi

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive elementmore » in the waste was thorium with activity of 5.10{sup −5} Ci/m{sup 3}. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of

Ceramic Single Phase High-Level Nuclear Waste Forms: Hollandite, Perovskite, and Pyrochlore

NASA Astrophysics Data System (ADS)

Vetter, M.; Wang, J.

2017-12-01

The lack of viable options for the safe, reliable, and long-term storage of nuclear waste is one of the primary roadblocks of nuclear energy's sustainable future. The method being researched is the incorporation and immobilization of harmful radionuclides (Cs, Sr, Actinides, and Lanthanides) into the structure of glasses and ceramics. Borosilicate glasses are the main waste form that is accepted and used by today's nuclear industry, but they aren't the most efficient in terms of waste loading, and durability is still not fully understood. Synroc-phase ceramics (i.e. hollandite, perovskite, pyrochlore, zirconolite) have many attractive qualities that glass waste forms do not: high waste loading, moderate thermal expansion and conductivity, high chemical durability, and high radiation stability. The only downside to ceramics is that they are more complex to process than glass. New compositions can be discovered by using an Artificial Neural Network (ANN) to have more options to optimize the composition, loading for performance by analyzing the non-linear relationships between ionic radii, electronegativity, channel size, and a mineral's ability to incorporate radionuclides into its structure. Cesium can be incorporated into hollandite's A-site, while pyrochlore and perovskite can incorporate actinides and lanthanides into their A-site. The ANN is used to predict new compositions based on hollandite's channel size, as well as the A-O bond distances of pyrochlore and perovskite, and determine which ions can be incorporated. These new compositions will provide more options for more experiments to potentially improve chemical and thermodynamic properties, as well as increased waste loading capabilities.

Low-level radioactive waste management handbook series: Low-level radioactive waste management in medical and biomedical research institutions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1987-03-01

Development of this handbook began in 1982 at the request of the Radhealth Branch of the California Department of Health Services. California Assembly Bill 1513 directed the DHS to ''evaluate the technical and economic feasibility of (1) reducing the volume, reactivity, and chemical and radioactive hazard of (low-level radioactive) waste and (2) substituting nonradioactive or short-lived radioactive materials for those radionuclides which require long-term isolation from the environment. A contract awarded to the University of California at Irvine-UCI (California Std. Agreement 79902), to develop a document focusing on methods for decreasing low-level radioactive waste (LLW) generation in institutions was amore » result of that directive. In early 1985, the US Department of Energy, through EG and G Idaho, Inc., contracted with UCI to expand, update, and revise the California text for national release.« less

Radioactive demonstration of final mineralized waste forms for Hanford waste treatment plant secondary waste (WTP-SW) by fluidized bed steam reforming (FBSR) using the bench scale reformer platform

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, C.; Burket, P.; Cozzi, A.

2014-08-01

The U.S. Department of Energy’s Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford’s tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in themore » time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as 137Cs, 129I, 99Tc, Cl, F, and SO4 that volatilize at the vitrification temperature of 1150°C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW.« less

M3FT-17OR0301070211 - Preparation of Hot Isostatically Pressed AgZ Waste Form Samples

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jubin, Robert Thomas; Bruffey, Stephanie H.; Jordan, Jacob A.

The production of radioactive iodine-bearing waste forms that exhibit long-term stability and are suitable for permanent geologic disposal has been the subject of substantial research interest. One potential method of iodine waste form production is hot isostatic pressing (HIP). Recent studies at Oak Ridge National Laboratory (ORNL) have investigated the conversion of iodine-loaded silver mordenite (I-AgZ) directly to a waste form by HIP. ORNL has performed HIP with a variety of sample compositions and pressing conditions. The base mineral has varied among AgZ (in pure and engineered forms), silver-exchanged faujasite, and silverexchanged zeolite A. Two iodine loading methods, occlusion andmore » chemisorption, have been explored. Additionally, the effects of variations in temperature and pressure of the process have been examined, with temperature ranges of 525°C–1,100°C and pressure ranges of 100–300 MPa. All of these samples remain available to collaborators upon request. The sample preparation detailed in this document is an extension of that work. In addition to previously prepared samples, this report documents the preparation of additional samples to support stability testing. These samples include chemisorbed I-AgZ and pure AgI. Following sample preparation, each sample was processed by HIP by American Isostatic Presses Inc. and returned to ORNL for storage. ORNL will store the samples until they are requested by collaborators for durability testing. The sample set reported here will support waste form durability testing across the national laboratories and will provide insight into the effects of varied iodine content on iodine retention by the produced waste form and on potential improvements in waste form durability provided by the zeolite matrix.« less

Liquid and Gaseous Waste Operations Department annual operating report CY 1996

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maddox, J.J.; Scott, C.B.

1997-03-01

This annual report summarizes operating activities dealing with the process waste system, the liquid low-level waste system, and the gaseous waste system. It also describes upgrade activities dealing with the process and liquid low-level waste systems, the cathodic protection system, a stack ventilation system, and configuration control. Maintenance activities are described dealing with nonradiological wastewater treatment plant, process waste treatment plant and collection system, liquid low-level waste system, and gaseous waste system. Miscellaneous activities include training, audits/reviews/tours, and environmental restoration support.

Laboratory Scoping Tests Of Decontamination Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor-Pashow, Kathryn M.; Nash, Charles A.; Crawford, Charles L.

2014-01-21

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 again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrificationmore » mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task seeks to examine the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that

Siting process for disposal site of low level radiactive waste in Thailand

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamkate, P.; Sriyotha, P.; Thiengtrongjit, S.

The radioactive waste in Thailand is composed of low level waste from the application of radioisotopes in medical treatment and industry, the operation of the 2 MW TRIGA Mark III Research Reactor and the production of radioisotopes at OAEP. In addition, the high activity of sealed radiation sources i.e. Cs-137 Co-60 and Ra-226 are also accumulated. Since the volume of treated waste has been gradually increased, the general needs for a repository become apparent. The near surface disposal method has been chosen for this aspect. The feasibility study on the underground disposal site has been done since 1982. The sitemore » selection criteria have been established, consisting of the rejection criteria, the technical performance criteria and the economic criteria. About 50 locations have been picked for consideration and 5 candidate sites have been selected and subsequent investigated. After thoroughly investigation, a definite location in Ratchburi Province, about 180 kilometers southwest of Bangkok, has been selected as the most suitable place for the near surface disposal of radioactive waste in Thailand.« less

Engineering assessment of low-level liquid waste disposal caisson locations at the 618-11 Burial Grounds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, S.J.; Fischer, D.D.; Crawford, R.C.

1982-06-01

Rockwell Hanford Operations is currently involved in an extensive effort to perform interim ground surface stabilization activities at retired low-level waste burial grounds located at the Hanford Site, Richland, Washington. The principal objective of these activities is to promote increased occupational and radiological safety at burial grounds. Interim stabilization activities include: (1) load testing (traversing burial ground surfaces with heavy equipment to promote incipient collapse of void spaces within the disposal structure and overburden), (2) barrier placement (placement of a {ge} 0.6 m soil barrier over existing overburden), and (3) revegetation (establishment of shallow rooted vegetation on the barrier tomore » mitigate deep rooted plant growth and to reduce erosion). Low-level waste disposal caissons were used in 300 Area Burial Grounds as internment structures for containerized liquid wastes. These caissons, by virtue of their contents, design and methods of closure, require long-term performance evaluation. As an initial activity to evaluate long-term performance, the accurate location of these structures is required. This topical report summarizes engineering activities used to locate caissons in the subsurface environment at the Burial Ground. Activities were conducted to locate caissons during surface stabilization activities. The surface locations were marked, photographed, and recorded on an as built engineering drawing. The recorded location of these caissons will augment long-term observations of confinement structure and engineered surface barrier performance. In addition, accurate caisson location will minimize occupational risk during monitoring and observation activities periodically conducted at the burial ground.« less

Unirradiated testing of the demonstration-scale ceramic waste form at ANL-West

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goff, K.M.; Simpson, M.F.; Bateman, K.J.

1997-12-01

The ceramic waste form is being developed by Argonne National Laboratory (ANL) as part of the demonstration of the electrometallurgical treatment of spent nuclear fuel for disposal. The alkali, alkaline earth, halide, and rare earth fission products are stabilized in zeolite, which is combined with glass and processed in a hot isostatic press (HIP) to form a ceramic composite. The transuranics, including plutonium, are also stabilized in this high-level waste. Most of the laboratory-scale development work is performed in the Chemical Technology Division of ANL in Illinois. At ANL-West in Idaho, this technology is being demonstrated on an engineering scalemore » before implementation with irradiated materials in a remote environment.« less

Ceramization of low and intermediate level radioactive wastes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fiquet, O.; Berson, X.

1993-12-31

A ceramic conditioning is studied for a large variety of low and intermediate level wastes. These wastes arise from several waste streams coming from all process steps of the fuel cycle. The physical properties of ceramics can advantageously be used for radioactive waste immobilization. Their chemical durability can offer a barrier against external aggression. More over, some minerals have possible host sites in their crystal structure for heavy elements which can confer the best immobilization mechanism. The general route for development studies is described giving compositions and process choices. Investigations have been conducted on clay materials and on the processmore » parameters which condition the final product properties. Two practical examples are described concerning chemical precipitation sludge resulting from liquid waste treatment and chamot used as a fluidized bed in a graphite incinerator. Important process parameters are put in evidence and the possibility of a pilot plant development is briefly mentioned. Results of investigations are promising to define a new route of conditioning.« less

System for chemically digesting low level radioactive, solid waste material

DOEpatents

Cowan, Richard G.; Blasewitz, Albert G.

1982-01-01

An improved method and system for chemically digesting low level radioactive, solid waste material having a high through-put. The solid waste material is added to an annular vessel (10) substantially filled with concentrated sulfuric acid. Concentrated nitric acid or nitrogen dioxide is added to the sulfuric acid within the annular vessel while the sulfuric acid is reacting with the solid waste. The solid waste is mixed within the sulfuric acid so that the solid waste is substantilly fully immersed during the reaction. The off gas from the reaction and the products slurry residue is removed from the vessel during the reaction.

Waste minimization for commercial radioactive materials users generating low-level radioactive waste. Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fischer, D.K.; Gitt, M.; Williams, G.A.

1991-07-01

The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations.more » This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature.« less

Source segregation and food waste prevention activities in high-density households in a deprived urban area

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rispo, A.; Williams, I.D., E-mail: idw@soton.ac.uk; Shaw, P.J.

Highlights: • Study of waste management in economically and socially deprived high-density housing. • Food waste segregation, prevention and recycling activities investigated. • Study involved a waste audit and household survey of 1034 households. • Populations in such areas are “hard-to-reach”. • Exceptional efforts and additional resources are required to improve performance. - Abstract: A waste audit and a household questionnaire survey were conducted in high-density housing estates in one of the most economically and socially deprived areas of England (Haringey, London). Such areas are under-represented in published research. The study examined source segregation, potential participation in a food wastemore » segregation scheme, and food waste prevention activities in five estates (1034 households). The results showed that: contamination of recyclables containers was low; ca. 28% of the mixed residual waste’s weight was recyclable; food waste comprised a small proportion of the waste from these residents, probably because of their relatively disadvantaged economic circumstances; and the recycling profile reflected an intermittent pattern of behaviour. Although the majority of respondents reported that they would participate in a food waste separation scheme, the response rate was low and many responses of “don’t know” were recorded. Municipalities committed to foster improved diversion from landfill need to recognise that there is no “quick and easy fix”, regardless of local or national aspirations. Lasting and sustained behaviour change requires time and the quality of service provision and associated infrastructure play a fundamental role in facilitating residents to participate effectively in waste management activities that maximise capture of source-segregated materials. Populations in deprived areas that reside in high-rise, high-density dwellings are “hard-to-reach” in terms of participation in recycling schemes and

Source segregation and food waste prevention activities in high-density households in a deprived urban area.

PubMed

Rispo, A; Williams, I D; Shaw, P J

2015-10-01

A waste audit and a household questionnaire survey were conducted in high-density housing estates in one of the most economically and socially deprived areas of England (Haringey, London). Such areas are under-represented in published research. The study examined source segregation, potential participation in a food waste segregation scheme, and food waste prevention activities in five estates (1034 households). The results showed that: contamination of recyclables containers was low; ca. 28% of the mixed residual waste's weight was recyclable; food waste comprised a small proportion of the waste from these residents, probably because of their relatively disadvantaged economic circumstances; and the recycling profile reflected an intermittent pattern of behaviour. Although the majority of respondents reported that they would participate in a food waste separation scheme, the response rate was low and many responses of "don't know" were recorded. Municipalities committed to foster improved diversion from landfill need to recognise that there is no "quick and easy fix", regardless of local or national aspirations. Lasting and sustained behaviour change requires time and the quality of service provision and associated infrastructure play a fundamental role in facilitating residents to participate effectively in waste management activities that maximise capture of source-segregated materials. Populations in deprived areas that reside in high-rise, high-density dwellings are "hard-to-reach" in terms of participation in recycling schemes and exceptional efforts and additional resources are usually required to improve performance. Copyright © 2015 Elsevier Ltd. All rights reserved.

Special Analysis: Disposal of ETF Activated Carbon Vessels in Slit Trenches at the E-Area Low-Level Waste Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collard, L.B.

2003-08-25

This Special Analysis (SA) addresses two contaminants of concern, H-3 and I-129, in three Effluent Treatment Facility (ETF) Activated Carbon Vessels awaiting disposal as solid waste. The Unreviewed Disposal Question (UDQ) evaluation listed two options for disposal of this waste, disposal as Components-in-Grout (CIG) or disposal in Slit Trenches with sealed openings to restrict release of H-3 form the vessels. Consumption of the CIG inventory limit and consumption of CIG facility volume are shown for the ETF vessels to allow easy comparison with the consumption of Slit Trench inventory limit and consumption of the Slit Trench facility volume . Themore » inventory projections are based on doubling the inventory of the three ETF vessels in the E-Area to account for the unknown inventory of three ETF vessels in the ETF. When the grout ultimately is assumed to degrade hydraulically, the water movement is not impeded as much as the release is accelerated by the presence of the grout. Under these conditions for the CIG trenches relative to the Slit Trenches, the well concentrations are higher, the inventory limit is lower and for a given inventory the inventory limit consumption is higher.« less

Activated carbon derived from waste coffee grounds for stable methane storage.

PubMed

Kemp, K Christian; Baek, Seung Bin; Lee, Wang-Geun; Meyyappan, M; Kim, Kwang S

2015-09-25

An activated carbon material derived from waste coffee grounds is shown to be an effective and stable medium for methane storage. The sample activated at 900 °C displays a surface area of 1040.3 m(2) g(-1) and a micropore volume of 0.574 cm(3) g(-1) and exhibits a stable CH4 adsorption capacity of ∼4.2 mmol g(-1) at 3.0 MPa and a temperature range of 298 ± 10 K. The same material exhibits an impressive hydrogen storage capacity of 1.75 wt% as well at 77 K and 100 kPa. Here, we also propose a mechanism for the formation of activated carbon from spent coffee grounds. At low temperatures, the material has two distinct types with low and high surface areas; however, activation at elevated temperatures drives off the low surface area carbon, leaving behind the porous high surface area activated carbon.

Activated carbon derived from waste coffee grounds for stable methane storage

NASA Astrophysics Data System (ADS)

Kemp, K. Christian; Baek, Seung Bin; Lee, Wang-Geun; Meyyappan, M.; Kim, Kwang S.

2015-09-01

An activated carbon material derived from waste coffee grounds is shown to be an effective and stable medium for methane storage. The sample activated at 900 °C displays a surface area of 1040.3 m2 g-1 and a micropore volume of 0.574 cm3 g-1 and exhibits a stable CH4 adsorption capacity of ˜4.2 mmol g-1 at 3.0 MPa and a temperature range of 298 ± 10 K. The same material exhibits an impressive hydrogen storage capacity of 1.75 wt% as well at 77 K and 100 kPa. Here, we also propose a mechanism for the formation of activated carbon from spent coffee grounds. At low temperatures, the material has two distinct types with low and high surface areas; however, activation at elevated temperatures drives off the low surface area carbon, leaving behind the porous high surface area activated carbon.

Consolidation process for producing ceramic waste forms

DOEpatents

Hash, Harry C.; Hash, Mark C.

2000-01-01

A process for the consolidation and containment of solid or semisolid hazardous waste, which process comprises closing an end of a circular hollow cylinder, filling the cylinder with the hazardous waste, and then cold working the cylinder to reduce its diameter while simultaneously compacting the waste. The open end of the cylinder can be sealed prior to or after the cold working process. The preferred method of cold working is to draw the sealed cylinder containing the hazardous waste through a plurality of dies to simultaneously reduce the diameter of the tube while compacting the waste. This process provides a quick continuous process for consolidating hazardous waste, including radioactive waste.

Final closure of a low level waste disposal facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Potier, J.M.

1995-12-31

The low-level radioactive waste disposal facility operated by the Agence Nationale pour la Gestion des Dechets Radioactifs near La Hague, France was opened in 1969 and is scheduled for final closure in 1996. The last waste package was received in June 1994. The total volume of disposed waste is approximately 525,000 m{sup 3}. The site closure consists of covering the disposal structures with a multi-layer impervious cap system to prevent rainwater from infiltrating the waste isolation system. A monitoring system has been set up to verify the compliance of infiltration rates with hydraulic performance objectives (less than 10 liters permore » square meter and per year).« less

Transporting Radioactive Waste: An Engineering Activity. Grades 5-12.

ERIC Educational Resources Information Center

HAZWRAP, The Hazardous Waste Remedial Actions Program.

This brochure contains an engineering activity for upper elementary, middle school, and high school students that examines the transportation of radioactive waste. The activity is designed to inform students about the existence of radioactive waste and its transportation to disposal sites. Students experiment with methods to contain the waste and…

Co-composting of green waste and food waste at low C/N ratio

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kumar, Mathava; Ou, Y.-L.; Lin, J.-G., E-mail: jglin@mail.nctu.edu.t

2010-04-15

In this study, co-composting of food waste and green waste at low initial carbon to nitrogen (C/N) ratios was investigated using an in-vessel lab-scale composting reactor. The central composite design (CCD) and response surface method (RSM) were applied to obtain the optimal operating conditions over a range of preselected moisture contents (45-75%) and C/N ratios (13.9-19.6). The results indicate that the optimal moisture content for co-composting of food waste and green waste is 60%, and the substrate at a C/N ratio of 19.6 can be decomposed effectively to reduce 33% of total volatile solids (TVS) in 12 days. The TVSmore » reduction can be modeled by using a second-order equation with a good fit. In addition, the compost passes the standard germination index of white radish seed indicating that it can be used as soil amendment.« less

LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Taylor-Pashow, K.; Nash, C.; McCabe, D.

2014-09-29

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 again. Alternate disposition of this stream would eliminate recycling of problematic components, and would enable de-coupled operation of the LAW melter and the Pretreatment Facilities. Eliminating this stream from recycling within WTP would also decrease the LAW vitrificationmore » mission duration and quantity of glass waste. This LAW Off-Gas Condensate stream contains components that are volatile at melter temperatures and are problematic for the glass waste form. Because this stream recycles within WTP, these components accumulate in the Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Approximately 32% of the sodium in Supplemental LAW comes from glass formers used to make the extra glass to dilute the halides to acceptable concentrations in the LAW glass, and diverting the stream reduces the halides in the recycled Condensate and is a key outcome of this work. Additionally, under possible scenarios where the LAW vitrification facility commences operation prior to the WTP Pretreatment facility, identifying a disposition path becomes vitally important. This task examines the potential treatment of this stream to remove radionuclides and subsequently disposition the decontaminated stream elsewhere, such as the Effluent Treatment Facility (ETF), for example. The treatment process envisioned is very similar to that used for the Actinide Removal Process (ARP) that has been operating for years at the Savannah River Site (SRS), and focuses on using mature radionuclide removal technologies that are also

Feasibility of thermophilic anaerobic processes for treating waste activated sludge under low HRT and intermittent mixing.

PubMed

Leite, Wanderli; Magnus, Bruna Scandolara; Guimarães, Lorena Bittencourt; Gottardo, Marco; Belli Filho, Paulo

2017-10-01

Thermophilic anaerobic digestion (AD) arises as an optimized solution for the waste activated sludge (WAS) management. However, there are few feasibility studies using low solids content typically found in the WAS, and that consider uncommon operational conditions such as intermittent mixing and low hydraulic retention time (HRT). In this investigation, a single-stage pilot reactor was used to treat WAS at low HRT (13, 9, 6 and 5 days) and intermittent mixing (withholding mixing 2 h prior feeding). Thermophilic anaerobic digestion (55 °C) was initiated from a mesophilic digester (35 °C) by the one-step startup strategy. Although instabilities on partial alkalinity (1245-3000 mgCaCO 3 /L), volatile fatty acids (1774-6421 mg/L acetic acid) and biogas production (0.21-0.09 m 3 /m 3 reactor .d) were observed, methanogenesis started to recover in 18 days. The thermophilic treatment of WAS at 13 and 9 days HRT efficiently converted VS into biogas (22 and 21%, respectively) and achieved high biogas yield (0.24 and 0.22 m 3 /kgVS fed , respectively). Intermittent mixing improved the retention of methanogens inside the reactor and reduced the washout effect even at low HRT (<9 days). The negative thermal balance found was influenced by the low solids content in the WAS (2.1% TS) and by the heat losses from the digester walls. The energy balance and economic analyses demonstrated the feasibility of thermophilic AD of WAS in a hypothetical full-scale system, when the heat energy could be recovered from methane in a scenario of higher solids concentration in the substrate (>5% TS). Copyright © 2017 Elsevier Ltd. All rights reserved.

Evaluating the biochemical methane potential (BMP) of low-organic waste at Danish landfills.

PubMed

Mou, Zishen; Scheutz, Charlotte; Kjeldsen, Peter

2014-11-01

The biochemical methane potential (BMP) is an essential parameter when using first order decay (FOD) landfill gas (LFG) generation models to estimate methane (CH4) generation from landfills. Different categories of waste (mixed, shredder and sludge waste) with a low-organic content and temporarily stored combustible waste were sampled from four Danish landfills. The waste was characterized in terms of physical characteristics (TS, VS, TC and TOC) and the BMP was analyzed in batch tests. The experiment was set up in triplicate, including blank and control tests. Waste samples were incubated at 55°C for more than 60 days, with continuous monitoring of the cumulative CH4 generation. Results showed that samples of mixed waste and shredder waste had similar BMP results, which was in the range of 5.4-9.1 kg CH4/ton waste (wet weight) on average. As a calculated consequence, their degradable organic carbon content (DOCC) was in the range of 0.44-0.70% of total weight (wet waste). Numeric values of both parameters were much lower than values of traditional municipal solid waste (MSW), as well as default numeric values in current FOD models. The sludge waste and temporarily stored combustible waste showed BMP values of 51.8-69.6 and 106.6-117.3 kg CH4/ton waste on average, respectively, and DOCC values of 3.84-5.12% and 7.96-8.74% of total weight. The same category of waste from different Danish landfills did not show significant variation. This research studied the BMP of Danish low-organic waste for the first time, which is important and valuable for using current FOD LFG generation models to estimate realistic CH4 emissions from modern landfills receiving low-organic waste. Copyright © 2014 Elsevier Ltd. All rights reserved.

Storage for greater-than-Class C low-level radioactive waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beitel, G.A.

1991-12-31

EG and G Idaho, Inc., at the Idaho National Engineering Laboratory (INEL) is actively pursuing technical storage alternatives for greater-than-Class C low-level radioactive waste (GTCC LLW) until a suitable licensed disposal facility is operating. A recently completed study projects that between 2200 and 6000 m{sup 3} of GTCC LLW will be generated by the year 2035; the base case estimate is 3250 m{sup 3}. The current plan envisions a disposal facility available as early as the year 2010. A long-term dedicated storage facility could be available in 1997. In the meantime, it is anticipated that a limited number of sealedmore » sources that are no longer useful and have GTCC concentrations of radionuclides will require storage. Arrangements are being made to provide this interim storage at an existing DOE waste management facility. All interim stored waste will subsequently be moved to the dedicated storage facility once it is operating. Negotiations are under way to establish a host site for interim storage, which may be operational, at the earliest, by the second quarter of 1993. Two major activities toward developing a long-term dedicated storage facility are ongoing. (a) An engineering study, which explores costs for alternatives to provide environmentally safe storage and satisfy all regulations, is being prepared. Details of some of the findings of that study will be presented. (b) There is also an effort under way to seek the assistance of one or more private companies in providing dedicated storage. Alternatives and options will be discussed.« less

SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Matthew C. Morrison; Kenneth J. Bateman; Michael F. Simpson

2010-11-01

ABSTRACT SCALE UP OF CERAMIC WASTE FORMS FOR THE EBR-II SPENT FUEL TREATMENT PROCESS Matthew C. Morrison, Kenneth J. Bateman, Michael F. Simpson Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 The ceramic waste process is the intended method for disposing of waste salt electrolyte, which contains fission products from the fuel-processing electrorefiners (ER) at the INL. When mixed and processed with other materials, the waste salt can be stored in a durable ceramic waste form (CWF). The development of the CWF has recently progressed from small-scale testing and characterization to full-scale implementation and experimentation using surrogate materialsmore » in lieu of the ER electrolyte. Two full-scale (378 kg and 383 kg) CWF test runs have been successfully completed with final densities of 2.2 g/cm3 and 2.1 g/cm3, respectively. The purpose of the first CWF was to establish material preparation parameters. The emphasis of the second pre-qualification test run was to evaluate a preliminary multi-section CWF container design. Other considerations were to finalize material preparation parameters, measure the material height as it consolidates in the furnace, and identify when cracking occurs during the CWF cooldown process.« less

Waste management for different fusion reactor designs

NASA Astrophysics Data System (ADS)

Rocco, Paolo; Zucchetti, Massimo

2000-12-01

Safety and Environmental Assessment of Fusion Power (SEAFP) waste management studies performed up to 1998 concerned three power tokamak designs. In-vessel structural materials consist of V-alloys or low activation martensitic (LAM) steel; tritium-producing materials are Li 2O, Pb-17Li, Li 4SiO 4 with a Be-multiplier; coolants are helium or water. The strategy chosen reduces permanent radwaste by recycling the in-vessel materials and by clearance of the other structures. Limits of the contact dose rate and specific activity of the waste allowing such options are defined accordingly. SEAFP activities for 1999 enlarge the analysis to three additional reactors with in-vessel structures made with SiC/SiC composites. These materials cannot be recycled due to their form and, according to national regulations of E.C. countries, long-lived activation products hinder near-surface burial (NSB).

Method for distinctive estimation of stored acidity forms in acid mine wastes.

PubMed

Li, Jun; Kawashima, Nobuyuki; Fan, Rong; Schumann, Russell C; Gerson, Andrea R; Smart, Roger St C

2014-10-07

Jarosites and schwertmannite can be formed in the unsaturated oxidation zone of sulfide-containing mine waste rock and tailings together with ferrihydrite and goethite. They are also widely found in process wastes from electrometallurgical smelting and metal bioleaching and within drained coastal lowland soils (acid-sulfate soils). These secondary minerals can temporarily store acidity and metals or remove and immobilize contaminants through adsorption, coprecipitation, or structural incorporation, but release both acidity and toxic metals at pH above about 4. Therefore, they have significant relevance to environmental mineralogy through their role in controlling pollutant concentrations and dynamics in contaminated aqueous environments. Most importantly, they have widely different acid release rates at different pHs and strongly affect drainage water acidity dynamics. A procedure for estimation of the amounts of these different forms of nonsulfide stored acidity in mining wastes is required in order to predict acid release rates at any pH. A four-step extraction procedure to quantify jarosite and schwertmannite separately with various soluble sulfate salts has been developed and validated. Corrections to acid potentials and estimation of acid release rates can be reliably based on this method.

Radioactive waste management complex low-level waste radiological composite analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCarthy, J.M.; Becker, B.H.; Magnuson, S.O.

1998-05-01

The composite analysis estimates the projected cumulative impacts to future members of the public from the disposal of low-level radioactive waste (LLW) at the Idaho National Engineering and Environmental Laboratory (INEEL) Radioactive Waste Management Complex (RWMC) and all other sources of radioactive contamination at the INEEL that could interact with the LLW disposal facility to affect the radiological dose. Based upon the composite analysis evaluation, waste buried in the Subsurface Disposal Area (SDA) at the RWMC is the only source at the INEEL that will significantly interact with the LLW facility. The source term used in the composite analysis consistsmore » of all historical SDA subsurface disposals of radionuclides as well as the authorized LLW subsurface disposal inventory and projected LLW subsurface disposal inventory. Exposure scenarios evaluated in the composite analysis include all the all-pathways and groundwater protection scenarios. The projected dose of 58 mrem/yr exceeds the composite analysis guidance dose constraint of 30 mrem/yr; therefore, an options analysis was conducted to determine the feasibility of reducing the projected annual dose. Three options for creating such a reduction were considered: (1) lowering infiltration of precipitation through the waste by providing a better cover, (2) maintaining control over the RWMC and portions of the INEEL indefinitely, and (3) extending the period of institutional control beyond the 100 years assumed in the composite analysis. Of the three options investigated, maintaining control over the RWMC and a small part of the present INEEL appears to be feasible and cost effective.« less

Modeling the long-term durability of concrete barriers in the context of low-activity waste storage

NASA Astrophysics Data System (ADS)

Protière, Y.; Samson, E.; Henocq, P.

2013-07-01

The paper investigates the long-term durability of concrete barriers in contact with a cementitious wasteform designed to immobilize low-activity nuclear waste. The high-pH pore solution of the wasteform contains high concentration level of sulfate, nitrate, nitrite and alkalis. The multilayer concrete/wasteform system was modeled using a multiionic reactive transport model accounting for coupling between species, dissolution/ precipitation reactions, and feedback effect. One of the primary objectives was to investigate the risk associated with the presence of sulfate in the wasteform on the durability of concrete. Simulation results showed that formation of expansive phases, such as gypsum and ettringite, into the concrete barrier was not extensive. Based on those results, it was not possible to conclude that concrete would be severely damaged, even after 5,000 years. Lab work was performed to provide data to validate the modeling results. Paste samples were immersed in sulfate contact solutions and analyzed to measure the impact of the aggressive environment on the material. The results obtained so far tend to confirm the numerical simulations.

Japanese Children and Plate Waste: Contexts of Low Self-Efficacy

ERIC Educational Resources Information Center

Abe, Keina; Akamatsu, Rie

2015-01-01

Objective: Leaving a portion of meals uneaten, known as plate waste, is a serious problem among children in Japan. Although children's confidence that they can completely finish meals is related to plate waste, the circumstances that influence this confidence are not known. This study examined situations in which low self-efficacy for finishing…

A review and overview of nuclear waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murray, R.L.

1984-12-31

An understanding of the status and issues in the management of radioactive wastes is based on technical information on radioactivity, radiation, biological hazard of radiation exposure, radiation standards, and methods of protection. The fission process gives rise to radioactive fission products and neutron bombardment gives activation products. Radioactive wastes are classified according to source: defense, commercial, industrial, and institutional; and according to physical features: uranium mill tailings, high-level, transuranic, and low-level. The nuclear fuel cycle, which contributes a large fraction of annual radioactive waste, starts with uranium ore, includes nuclear reactor use for electrical power generation, and ends with ultimatemore » disposal of residues. The relation of spent fuel storage and reprocessing is governed by technical, economic, and political considerations. Waste has been successfully solidified in glass and other forms and choices of the containers for the waste form are available. Methods of disposal of high-level waste that have been investigated are transmutation by neutron bombardment, shipment to Antartica, deep-hole insertion, subseabed placement, transfer by rocket to an orbit in space, and disposal in a mined cavity. The latter is the favored method. The choices of host geological media are salt, basalt, tuff, and granite.« less

[Substantiation of a complex of radiation-hygienic approaches to the management of very low-level waste].

PubMed

Korenkov, I P; Lashchenova, T N; Shandala, N K

2015-01-01

In the article there are presented materials on radiation-hygienic approaches to the treatment of very low level radioactive waste (VLLW) and industrial waste containing radionuclides. There is done detailed information on radiation-hygienic principles and criteria for the assurance ofradiation safety in the collection, transportation, storage and processing of VLLW as a category of radioactive waste.. Particular attention is paid to the problem of designing VLLW landfill site choice, system of radiation monitoring in operation and decommissioning of the landfill. There are presented data about the criteria for the release of VLLW buried at the site, from regulatory control. Also there are considered in detail the radiation-hygienic requirements for radiation safety of industrial waste containing radionuclides for which there is assumed unlimited and limited use of solid materials in economic activity, based on the requirements ofthe revised Basic Sanitary Rules for Radiation Safety - 99/2010. There are considered basic requirements for the organization of industrial waste landfill. As an example, there-are presented the hygiene requirements for industrial waste management and results of waste categorization in Northern Federal Enterprise for Radioactive Waste Management.

Working conditions and environmental exposures among electronic waste workers in Ghana.

PubMed

Akormedi, Matthew; Asampong, Emmanuel; Fobil, Julius N

2013-01-01

To investigate and describe informal e-waste recycling and working conditions at Agbogbloshie, Accra, Ghana. We conducted in-depth interviews which were qualitatively analysed from a grounded theory perspective. Workers obtained e-waste from the various residential areas in Accra, then dismantled and burned them in open air to recover copper, aluminum, steel, and other products for sale to customers on-site or at the nearby Agbogbloshie market. The processers worked under unhealthy conditions often surrounded by refuse and human excreta without any form of protective gear and were thus exposed to frequent burns, cuts, and inhalation of highly contaminated fumes. We observed no form of social security/support system for the workers, who formed informal associations to support one another in times of difficulty. e-waste recycling working conditions were very challenging and presented serious hazards to worker health and wellbeing. Formalizing the e-waste processing activities requires developing a framework of sustainable financial and social security for the e-waste workers, including adoption of low-cost, socially acceptable, easy-to-operate, and cleaner technologies that would safeguard the health of the workers and the general public.

NRC`s proposed rulemaking on the documentation and reporting of low-level radioactive waste shipment manifest information

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lahs, W.R.; Haisfield, M.F.

1991-12-31

Since the 1982 promulgation of regulations for the land disposal of low-level radioactive waste (LLW), requirements have been in place to control transfers of LLW intended for disposal at licensed land disposal facilities. These requirements established a manifest tracking system and defined processes to control transfers of LLW intended for disposal at a land disposal facility. Because the regulations did not specify the format for the LLW shipment manifests, it was not unexpected that the two operators of the three currently operating disposal sites should each have developed their own manifest forms. The forms have many similarities and the collectedmore » information, in many cases, is identical; however, these manifests incorporate unique operator preferences and also reflect the needs of the Agreement State regulatory authority in the States where the disposal sites are located. Since Agreement State regulations must be compatible with, but need not always be identical to, those of the Nuclear Regulatory Commission (NRC), the possibility of a proliferation of different manifest forms containing variations in collected information could be envisioned. If these manifests were also to serve a shipping paper purpose, effective integration of the Department of Transportations` (DOT) requirements would also have to be addressed. This wide diversity in uses of manifest information by Federal and State regulatory authorities, other State or Compact entities, and disposal site operators, suggested a single consolidated approach to develop a uniform manifest format with a baseline information content and to define recordkeeping requirements. The NRC, in 1989, had embarked on a rulemaking activity to establish a base set of manifest information needs for regulatory purposes. In response to requests from State and Regional Compact organizations who are attempting to design, develop and operate LLW disposal facilities, and with the general support of Agreement State

Process for treating waste water having low concentrations of metallic contaminants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Looney, Brian B; Millings, Margaret R; Nichols, Ralph L

A process for treating waste water having a low level of metallic contaminants by reducing the toxicity level of metallic contaminants to an acceptable level and subsequently discharging the treated waste water into the environment without removing the treated contaminants.

Measurement of actinides and strontium-90 in high activity waste

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maxwell, S.L. III; Nelson, M.R.

1994-08-01

The reliable measurement of trace radionuclides in high activity waste is important to support waste processing activities at SRS (F and H Area Waste Tanks, Extended Sludge Processing (ESP) and In-Tank precipitation (ITP) processing). Separation techniques are needed to remove high levels of gamma activity and alpha/beta interferences prior to analytical measurement. Using new extraction chromatographic resins from EiChrom Industries, Inc., the SRS Central Laboratory has developed new high speed separation methods that enable measurement of neptunium, thorium, uranium, plutonium, americium and strontium-90 in high activity waste solutions. Small particle size resin and applied vacuum are used to reduce analysismore » times and enhance column performance. Extraction chromatographic resins are easy to use and eliminate the generation of contaminated liquid organic waste.« less

Packaging waste prevention activities: A life cycle assessment of the effects on a regional waste management system.

PubMed

Nessi, Simone; Rigamonti, Lucia; Grosso, Mario

2015-09-01

A life cycle assessment was carried out to evaluate the effects of two packaging waste prevention activities on the overall environmental performance of the integrated municipal waste management system of Lombardia region, Italy. The activities are the use of refined tap water instead of bottled water for household consumption and the substitution of liquid detergents packaged in single-use containers by those distributed 'loose' through self-dispensing systems and refillable containers. A 2020 baseline scenario without waste prevention is compared with different waste prevention scenarios, where the two activities are either separately or contemporaneously implemented, by assuming a complete substitution of the traditional product(s). The results show that, when the prevention activities are carried out effectively, a reduction in total waste generation ranging from 0.14% to 0.66% is achieved, corresponding to a 1-4% reduction of the affected packaging waste fractions (plastics and glass). However, the improvements in the overall environmental performance of the waste management system can be far higher, especially when bottled water is substituted. In this case, a nearly 0.5% reduction of the total waste involves improvements ranging mostly between 5 and 23%. Conversely, for the substitution of single-use packaged liquid detergents (0.14% reduction of the total waste), the achieved improvements do not exceed 3% for nearly all impact categories. © The Author(s) 2015.

76 FR 58543 - Draft Policy Statement on Volume Reduction and Low-Level Radioactive Waste Management

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-21

...-Level Radioactive Waste Management AGENCY: Nuclear Regulatory Commission. ACTION: Reopening of comment... for public comment a draft Policy Statement on Volume Reduction and Low-Level Radioactive Waste Management that updates the 1981 Policy Statement on Low-Level Waste Volume Reduction. The revised Policy...

Reduction of 68Ge activity containing liquid waste from 68Ga PET chemistry in nuclear medicine and radiopharmacy by solidification.

PubMed

de Blois, Erik; Chan, Ho Sze; Roy, Kamalika; Krenning, Eric P; Breeman, Wouter A P

PET with 68 Ga from the TiO 2 - or SnO 2 - based 68 Ge/ 68 Ga generators is of increasing interest for PET imaging in nuclear medicine. In general, radionuclidic purity ( 68 Ge vs. 68 Ga activity) of the eluate of these generators varies between 0.01 and 0.001%. Liquid waste containing low amounts of 68 Ge activity is produced by eluting the 68 Ge/ 68 Ga generators and residues from PET chemistry. Since clearance level of 68 Ge activity in waste may not exceed 10 Bq/g, as stated by European Directive 96/29/EURATOM, our purpose was to reduce 68 Ge activity in solution from >10 kBq/g to <10 Bq/g; which implies the solution can be discarded as regular waste. Most efficient method to reduce the 68 Ge activity is by sorption of TiO 2 or Fe 2 O 3 and subsequent centrifugation. The required 10 Bq per mL level of 68 Ge activity in waste was reached by Fe 2 O 3 logarithmically, whereas with TiO 2 asymptotically. The procedure with Fe 2 O 3 eliminates ≥90% of the 68 Ge activity per treatment. Eventually, to simplify the processing a recirculation system was used to investigate 68 Ge activity sorption on TiO 2 , Fe 2 O 3 or Zeolite. Zeolite was introduced for its high sorption at low pH, therefore 68 Ge activity containing waste could directly be used without further interventions. 68 Ge activity containing liquid waste at different HCl concentrations (0.05-1.0 M HCl), was recirculated at 1 mL/min. With Zeolite in the recirculation system, 68 Ge activity showed highest sorption.

Impeding 99Tc(IV) mobility in novel waste forms

PubMed Central

Lee, Mal-Soon; Um, Wooyong; Wang, Guohui; Kruger, Albert A.; Lukens, Wayne W.; Rousseau, Roger; Glezakou, Vassiliki-Alexandra

2016-01-01

Technetium (99Tc) is an abundant, long-lived radioactive fission product whose mobility in the subsurface is largely governed by its oxidation state. Tc immobilization is crucial for radioactive waste management and environmental remediation. Tc(IV) incorporation in spinels has been proposed as a novel method to increase Tc retention in glass waste forms during vitrification. However, experiments under high-temperature and oxic conditions show reoxidation of Tc(IV) to volatile pertechnetate, Tc(VII). Here we examine this problem with ab initio molecular dynamics simulations and propose that, at elevated temperatures, doping with first row transition metal can significantly enhance Tc retention in magnetite in the order Co>Zn>Ni. Experiments with doped spinels at 700 °C provide quantitative confirmation of the theoretical predictions in the same order. This work highlights the power of modern, state-of-the-art simulations to provide essential insights and generate theory-inspired design criteria of complex materials at elevated temperatures. PMID:27357121

Hanford Site Composite Analysis Technical Approach Description: Waste Form Release.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hardie, S.; Paris, B.; Apted, M.

2017-09-14

The U.S. Department of Energy (DOE) in DOE O 435.1 Chg. 1, Radioactive Waste Management, requires the preparation and maintenance of a composite analysis (CA). The primary purpose of the CA is to provide a reasonable expectation that the primary public dose limit is not likely to be exceeded by multiple source terms that may significantly interact with plumes originating at a low-level waste disposal facility. The CA is used to facilitate planning and land use decisions that help assure disposal facility authorization will not result in long-term compliance problems; or, to determine management alternatives, corrective actions or assessment needs,more » if potential problems are identified.« less

Collection of low-grade waste heat for enhanced energy harvesting

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dede, Ercan M., E-mail: eric.dede@tema.toyota.com; Schmalenberg, Paul; Wang, Chi-Ming

Enhanced energy harvesting through the collection of low-grade waste heat is experimentally demonstrated. A structural optimization technique is exploited in the design of a thermal-composite substrate to guide and gather the heat emanating from multiple sources to a predetermined location. A thermoelectric generator is then applied at the selected focusing region to convert the resulting low-grade waste heat to electrical power. The thermal characteristics of the device are experimentally verified by direct temperature measurements of the system and numerically validated via heat conduction simulations. Electrical performance under natural and forced convection is measured, and in both cases, the device withmore » optimized heat flow control plus energy harvesting demonstrates increased power generation when compared with a baseline waste heat recovery system. Electronics applications include energy scavenging for autonomously powered sensor networks or self-actuated devices.« less

Performance of Spent Mushroom Farming Waste (SMFW) Activated Carbon for Ni (II) Removal

NASA Astrophysics Data System (ADS)

Desa, N. S. Md; Ghani, Z. Ab; Talib, S. Abdul; Tay, C. C.

2016-07-01

The feasibility of a low cost agricultural waste of spent mushroom farming waste (SMFW) activated carbon for Ni(II) removal was investigated. The batch adsorption experiments of adsorbent dosage, pH, contact time, metal concentration, and temperature were determined. The samples were shaken at 125 rpm, filtered and analyzed using ICP-OES. The fifty percent of Ni(II) removal was obtained at 0.63 g of adsorbent dosage, pH 5-6 (unadjusted), 60 min contact time, 50 mg/L Ni(II) concentration and 25 °C temperature. The evaluated SMFW activated carbon showed the highest performance on Ni(II) removal compared to commercial Amberlite IRC86 resin and zeolite NK3. The result indicated that SMFW activated carbon is a high potential cation exchange adsorbent and suitable for adsorption process for metal removal. The obtained results contribute toward application of developed SMFW activated carbon in industrial pilot study.

RESULTS OF THE ENVIRONMENTAL MANAGEMENT (EM) CORPORATE PROJECT TEAM DISPOSING WASTE & REDUCING RISK

DOE Office of Scientific and Technical Information (OSTI.GOV)

SHRADER, T.A.; KNERR, R.

2005-01-31

In 2002, the US Department of Energy's (DOE) Office of Environmental Management (EM) released the Top-To-Bottom Review of cognizant clean-up activities around the DOE Complex. The review contained a number of recommendations for changing the way EM operates in order to reduce environmental risk by significantly accelerating clean-up at the DOE-EM sites. In order to develop and implement these recommendations, a number of corporate project teams were formed to identify, evaluate, and initiate implementation of alternatives for the different aspects of clean-up. In August 2002, a corporate team was formed to review all aspects of the management, treatment, and disposalmore » of low level radioactive waste (LLW), mixed low level radioactive waste (MLLW), transuranic waste (TRU), and hazardous waste (HW). Over the next 21 months, the Corporate Project Team: Disposing Waste, Reducing Risk, developed a number of alternatives for implementing the recommendations of the Top-To-Bottom Review based on information developed during numerous site visits and interviews with complex and industry personnel. With input from over a dozen EM sites at various stages of clean-up, the team identified the barriers to the treatment and disposal of low level waste, mixed low level waste, and transuranic waste. Once identified, preliminary design alternatives were developed and presented to the Acquisition Authority (for this project, the Assistant Secretary for Environmental Management) for review and approval. Once the preliminary design was approved, the team down selected to seven key alternatives which were subsequently fully developed in the Project Execution Plan. The seven most viable alternatives were: (1) creation of an Executive Waste Disposal Board; (2) projectizing the disposal of low level waste and mixed low level waste; (3) creation of a National Consolidation and Acceleration Facility for waste; (4) improvements to the Broad Spectrum contract; (5) improvements to the

Thermal investigation of nuclear waste disposal in space

NASA Technical Reports Server (NTRS)

Wilkinson, C. L.

1981-01-01

A thermal analysis has been conducted to determine the allowable size and response of bare and shielded nuclear waste forms in both low earth orbit and at 0.85 astronomical units. Contingency conditions of re-entry with a 45 deg and 60 deg aeroshell are examined as well as re-entry of a spherical shielded waste form. A variety of shielded schemes were examined and the waste form thermal response for each determined. Two optimum configurations were selected. The thermal response of these two shielded waste configurations to indefinite exposure to ground conditions following controlled and uncontrolled re-entry is determined. In all cases the prime criterion is that waste containment must be maintained.

HANFORD MEDIUM-LOW CURIE WASTE PRETREATMENT ALTERNATIVES PROJECT FRACTIONAL CRYSTALLIZATION PILOT SCALE TESTING FINAL REPORT

DOE Office of Scientific and Technical Information (OSTI.GOV)

HERTING DL

2008-09-16

The Fractional Crystallization Pilot Plant was designed and constructed to demonstrate that fractional crystallization is a viable way to separate the high-level and low-activity radioactive waste streams from retrieved Hanford single-shell tank saltcake. The focus of this report is to review the design, construction, and testing details of the fractional crystallization pilot plant not previously disseminated.

Temporal variation of mass-wasting activity in Mount St. Helens crater, Washington, U. S. A. indicated by seismic activity

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mills, H.H.

1991-11-01

In the crater of Mount St. Helens, formed during the eruption of 18 May 1980, thousands of rockfalls may occur in a single day, and some rock and dirty-snow avalanches have traveled more than 1 km from their source. Because most seismic activity in the crater is produced by mass wasting, the former can be used to monitor the latter. The number and amplitude of seismic events per unit time provide a generalized measure of mass-wasting activity. In this study 1-min averages of seismic amplitudes were used as an index of rockfall activity during summer and early fall. Plots ofmore » this index show the diurnal cycle of rockfall activity and establish that the peak in activity occurs in mid to late afternoon. A correlation coefficient of 0.61 was found between daily maximum temperature and average seismic amplitude, although this value increases to 0.72 if a composite temperature variable that includes the maximum temperature of 1 to 3 preceding days as well as the present day is used. Correlation with precipitation is much weaker.« less

The effects of crystallization and residual glass on the chemical durability of iron phosphate waste forms containing 40 wt% of a high MoO3 Collins-CLT waste

NASA Astrophysics Data System (ADS)

Hsu, Jen-Hsien; Bai, Jincheng; Kim, Cheol-Woon; Brow, Richard K.; Szabo, Joe; Zervos, Adam

2018-03-01

The effects of cooling rate on the chemical durability of iron phosphate waste forms containing up to 40 wt% of a high MoO3 Collins-CLT waste simulant were determined at 90 °C using the product consistency test (PCT). The waste form, designated 40wt%-5, meets appropriate Department of Energy (DOE) standards when rapidly quenched from the melt (as-cast) and after slow cooling following the CCC (canister centerline cooling)-protocol, although the quenched glass is more durable. The analysis of samples from the vapor hydration test (VHT) and the aqueous corrosion test (differential recession test) reveals that rare earth orthophosphate (monazite) and Zr-pyrophosphate crystals that form on cooling are more durable than the residual glass in the 40wt%-5 waste form. The residual glass in the CCC-treated samples has a greater average phosphate chain length and a lower Fe/P ratio, and those contribute to its faster corrosion kinetics.

Final repository for Denmark's low- and intermediate level radioactive waste

NASA Astrophysics Data System (ADS)

Nilsson, B.; Gravesen, P.; Petersen, S. S.; Binderup, M.

2012-12-01

Bertel Nilsson*, Peter Gravesen, Stig A. Schack Petersen, Merete Binderup Geological Survey of Denmark and Greenland (GEUS), Øster Voldgade 10, 1350 Copenhagen, Denmark, * email address bn@geus.dk The Danish Parliament decided in 2003 that the temporal disposal of the low- and intermediate level radioactive waste at the nuclear facilities at Risø should find another location for a final repository. The Danish radioactive waste must be stored on Danish land territory (exclusive Greenland) and must hold the entire existing radioactive waste, consisting of the waste from the decommissioning of the nuclear facilities at Risø, and the radioactive waste produced in Denmark from hospitals, universities and industry. The radioactive waste is estimated to a total amount of up to 10,000 m3. The Geological Survey of Denmark and Greenland, GEUS, is responsible for the geological studies of suitable areas for the repository. The task has been to locate and recognize non-fractured Quaternary and Tertiary clays or Precambrian bedrocks with low permeability which can isolate the radioactive waste from the surroundings the coming more than 300 years. Twenty two potential areas have been located and sequential reduced to the most favorable two to three locations taking into consideration geology, hydrogeology, nature protection and climate change conditions. Further detailed environmental and geology investigations will be undertaken at the two to three potential localities in 2013 to 2015. This study together with a study of safe transport of the radioactive waste and an investigation of appropriate repository concepts in relation to geology and safety analyses will constitute the basis upon which the final decision by the Danish Parliament on repository concept and repository location. The final repository is planned to be established and in operation at the earliest 2020.

Final Project Report CFA-14-6357: A New Paradigm for Understanding Multiphase Ceramic Waste Form Performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brinkman, Kyle; Bordia, Rajendra; Reifsnider, Kenneth

This project fabricated model multiphase ceramic waste forms with processing-controlled microstructures followed by advanced characterization with synchrotron and electron microscopy-based 3D tomography to provide elemental and chemical state-specific information resulting in compositional phase maps of ceramic composites. Details of 3D microstructural features were incorporated into computer-based simulations using durability data for individual constituent phases as inputs in order to predict the performance of multiphase waste forms with varying microstructure and phase connectivity.

Development And Initial Testing Of Off-Gas Recycle Liquid From The WTP Low Activity Waste Vitrification Process - 14333

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCabe, Daniel J.; Wilmarth, William R.; Nash, Charles A.

2014-01-07

The Waste Treatment and Immobilization Plant (WTP) process flow was designed to pre-treat feed from the Hanford tank farms, separate it into a High Level Waste (HLW) and Low Activity Waste (LAW) fraction and vitrify each fraction in separate facilities. Vitrification of the waste generates an aqueous condensate stream from the off-gas processes. This stream originates from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrospray Precipitator (WESP). Currently, the baseline plan for disposition of the stream from the LAW melter is to recycle it to the Pretreatment facility where it gets evaporated and processedmore » into the LAW melter again. If the Pretreatment facility is not available, the baseline disposition pathway is not viable. Additionally, some components in the stream are volatile at melter temperatures, thereby accumulating to high concentrations in the scrubbed stream. It would be highly beneficial to divert this stream to an alternate disposition path to alleviate the close-coupled operation of the LAW vitrification and Pretreatment facilities, and to improve long-term throughput and efficiency of the WTP system. In order to determine an alternate disposition path for the LAW SBS/WESP Recycle stream, a range of options are being studied. A simulant of the LAW Off-Gas Condensate was developed, based on the projected composition of this stream, and comparison with pilot-scale testing. The primary radionuclide that vaporizes and accumulates in the stream is Tc-99, but small amounts of several other radionuclides are also projected to be present in this stream. The processes being investigated for managing this stream includes evaporation and radionuclide removal via precipitation and adsorption. During evaporation, it is of interest to investigate the formation of insoluble solids to avoid scaling and plugging of equipment. Key parameters for radionuclide removal include identifying effective precipitation or ion

Extracellular polymers of ozonized waste activated sludge.

PubMed

Liu, J C; Lee, C H; Lai, J Y; Wang, K C; Hsu, Y C; Chang, B V

2001-01-01

Effect of ozonation on characteristics of waste activated sludge was investigated in the current study. Concentrations of cell-bound extracellular polymers (washed ECPs) did not change much upon ozonation, whereas the sum of cell-bound and soluble extracellular polymers (unwashed ECPs) increased with increasing ozone dose. Washed ECPs in original sludge as divided by molecular weight distribution was 39% < 1,000 Da (low MW), 30% from 1,000 to 10,000 Da (medium MW), and 31% > 10,000 Da (high MW). It was observed that the low-MW fraction decreased, and the high-MW fraction increased in ozonized sludge. The unwashed ECPs were characterized as 44% in low MW, 30% in medium MW, and 26% in high MW. Both low-MW and medium-MW fractions of unwashed ECPs decreased while high-MW fraction increased in ozonized sludge. The dewaterability of ozonized sludge, assessed by capillary suction time (CST) and specific resistance to filtration (SRF), deteriorated with ozone dose. The optimal dose of cationic polyelectrolyte increased with increasing ozone dose. The production rate and the accumulated amount of methane gas of ozonized sludge were also higher.

Integrated software system for low level waste management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Worku, G.

1995-12-31

In the continually changing and uncertain world of low level waste management, many generators in the US are faced with the prospect of having to store their waste on site for the indefinite future. This consequently increases the set of tasks performed by the generators in the areas of packaging, characterizing, classifying, screening (if a set of acceptance criteria applies), and managing the inventory for the duration of onsite storage. When disposal sites become available, it is expected that the work will require re-evaluating the waste packages, including possible re-processing, re-packaging, or re-classifying in preparation for shipment for disposal undermore » the regulatory requirements of the time. In this day and age, when there is wide use of computers and computer literacy is at high levels, an important waste management tool would be an integrated software system that aids waste management personnel in conducting these tasks quickly and accurately. It has become evident that such an integrated radwaste management software system offers great benefits to radwaste generators both in the US and other countries. This paper discusses one such approach to integrated radwaste management utilizing some globally accepted radiological assessment software applications.« less

Low temperature ozone oxidation of solid waste surrogates

NASA Astrophysics Data System (ADS)

Nabity, James A.; Lee, Jeffrey M.

2015-09-01

Solid waste management presents a significant challenge to human spaceflight and especially, long-term missions beyond Earth orbit. A six-month mission will generate over 300 kg of solid wastes per crewmember that must be dealt with to eliminate the need for storage and prevent it from becoming a biological hazard to the crew. There are several methods for the treatment of wastes that include oxidation via ozone, incineration, microbial oxidation or pyrolysis and physical methods such as microwave drying and compaction. In recent years, a low temperature oxidation process using ozonated water has been developed for the chemical conversion of organic wastes to CO2 and H2O. Experiments were conducted to evaluate the rate and effectiveness with which ozone oxidized several different waste materials. Increasing the surface area by chopping or shredding the solids into small pieces more than doubled the rate of oxidation. A greater flow of ozone and agitation of the ozonated water system also increased processing rates. Of the materials investigated, plastics have proven the most difficult to oxidize. The processing of plastics above the glass transition temperatures caused the plastics to clump together which reduced the exposed surface area, while processing at lower temperatures reduced surface reaction kinetics.

Sustainable mechanical biological treatment of solid waste in urbanized areas with low recycling rates.

PubMed

Trulli, Ettore; Ferronato, Navarro; Torretta, Vincenzo; Piscitelli, Massimiliano; Masi, Salvatore; Mancini, Ignazio

2018-01-01

Landfill is still the main technological facility used to treat and dispose municipal solid waste (MSW) worldwide. In developing countries, final dumping is applied without environmental monitoring and soil protection since solid waste is mostly sent to open dump sites while, in Europe, landfilling is considered as the last option since reverse logistic approaches or energy recovery are generally encouraged. However, many regions within the European Union continue to dispose of MSW to landfill, since modern facilities have not been introduced owing to unreliable regulations or financial sustainability. In this paper, final disposal activities and pre-treatment operations in an area in southern Italy are discussed, where final disposal is still the main option for treating MSW and the recycling rate is still low. Mechanical biological treatment (MBT) facilities are examined in order to evaluate the organic stabilization practices applied for MSW and the efficiencies in refuse derived fuel production, organic waste stabilization and mass reduction. Implementing MBT before landfilling the environmental impact and waste mass are reduced, up to 30%, since organic fractions are stabilized resulting an oxygen uptake rate less than 1600 mgO 2  h -1  kg -1 VS , and inorganic materials are exploited. Based on experimental data, this work examines MBT application in contexts where recycling and recovery activities have not been fully developed. The evidence of this study led to state that the introduction of MBT facilities is recommended for developing regions with high putrescible waste production in order to decrease environmental pollution and enhance human healthy. Copyright © 2017 Elsevier Ltd. All rights reserved.

Glass-bonded iodosodalite waste form for immobilization of 129 I

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chong, Saehwa; Peterson, Jacob A.; Riley, Brian J.

Immobilization of radioiodine (e.g., 129I, 131I) is an important need for current and future nuclear fuel cycles. For the current work, iodosodalite [Na8(AlSiO4)6I2] was synthesized hydrothermally from metakaolin, NaI, and NaOH. Following hydrothermal treatment, dried unwashed powders were used to make glass-bonded iodosodalite waste forms by heating pressed pellets at 650, 750, or 850 °C with two different types of sodium borosilicate glass binders, i.e., NBS-4 and SA-800. These heat-treated specimens were characterized with X-ray diffraction, Fourier-transform infrared spectroscopy, scanning electron microscopy, energy dispersive spectroscopy, thermal analysis, porosity and density measurements, neutron activation analysis, and inductively-coupled plasma mass spectrometry. Themore » pellets mixed with 10 mass% of NBS-4 or SA-800 and heat-treated at 750 °C contained relatively high percentage iodine retention (~44-47 % of the maximum iodine loading) with relatively low porosities, while other pellets with higher percentages iodine retention either contained higher porosity or were not completely sintered. ASTM C1308 chemical durability tests of monolithic specimens showed a large initial release of Na, Al, Si, and I on the first day, possibly from water-soluble salt crystals or non-durable amorphous phases. Release rates of Na and Si were higher than for Al and I, probably due to a poorly durable Na-Si-O phase from the glass bonding matrix. The cumulative normalized release of iodine was 12.5 g m-2 for the first 10 1-d exchanges, suggestive of coherent dissolution. The average release rate from 10-24 days during the 7-d exchange intervals was 0.2336 g m-2 d-1.« less

Activation and characterization of waste coffee grounds as bio-sorbent

NASA Astrophysics Data System (ADS)

Mariana; Marwan; Mulana, F.; Yunardi; Ismail, T. A.; Hafdiansyah, M. F.

2018-03-01

As the city well known for its culture of coffee drinkers, modern and traditional coffee shops are found everywhere in Banda Aceh, Indonesia. High number of coffee shops in the city generates large quantities of spent coffee grounds as waste without any effort to convert them as other valuable products. In an attempt to reduce environmental problems caused by used coffee grounds, this research was conducted to utilize waste coffee grounds as an activated carbon bio-sorbent. The specific purpose of this research is to improve the performance of coffee grounds bio-sorbent through chemical and physical activation, and to characterize the produced bio-sorbent. Following physical activation by carbonization, a chemical activation was achieved by soaking the carbonized waste coffee grounds in HCl solvent and carbonization process. The activated bio-sorbent was characterized for its morphological properties using Scanning Electron Microscopy (SEM), its functional groups by Fourier Transform Infra-Red Spectrophotometer (FTIR), and its material characteristics using X-Ray Diffraction (XRD). Characterization of the activated carbon prepared from waste coffee grounds shows that it meets standard quality requirement in accordance with Indonesian National Standard, SNI 06-3730-1995. Activation process has modified the functional groups of the waste coffee grounds. Comparing to natural waste coffee grounds, the resulted bio-sorbent demonstrated a more porous surface morphology following activation process. Consequently, such bio-sorbent is a potential source to be used as an adsorbent for various applications.

Comparison of soil heavy metal pollution caused by e-waste recycling activities and traditional industrial operations.

PubMed

He, Kailing; Sun, Zehang; Hu, Yuanan; Zeng, Xiangying; Yu, Zhiqiang; Cheng, Hefa

2017-04-01

The traditional industrial operations are well recognized as an important source of heavy metal pollution, while that caused by the e-waste recycling activities, which have sprouted in some developing countries, is often overlooked. This study was carried out to compare the status of soil heavy metal pollution caused by the traditional industrial operations and the e-waste recycling activities in the Pearl River Delta, and assess whether greater attention should be paid to control the pollution arising from e-waste recycling activities. Both the total contents and the chemical fractionation of major heavy metals (As, Cr, Cd, Ni, Pb, Cu, and Zn) in 50 surface soil samples collected from the e-waste recycling areas and 20 soil samples from the traditional industrial zones were determined. The results show that the soils in the e-waste recycling areas were mainly polluted by Cu, Zn, As, and Cd, while Cu, Zn, As, Cd, and Pb were the major heavy metals in the soils from the traditional industrial zones. Statistical analyses consistently show that Cu, Cd, Pb, and Zn in the surface soils from both types of sites were contributed mostly by human activities, while As, Cr, and Ni in the soils were dominated by natural background. No clear distinction was found on the pollution characteristic of heavy metals in the surface soils between the e-waste recycling areas and traditional industrial zones. The potential ecological risk posed by heavy metals in the surface soils from both types of sites, which was dominated by that from Cd, ranged from low to moderate. Given the much shorter development history of e-waste recycling and its largely unregulated nature, significant efforts should be made to crack down on illegal e-waste recycling and strengthen pollution control for related activities.

Nevada Test Site Waste Acceptance Criteria

DOE Office of Scientific and Technical Information (OSTI.GOV)

U. S. Department of Energy, National Nuclear Security Administration Nevada Site Office

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) waste acceptance criteria (WAC). The WAC provides the requirements, terms, and conditions under which the Nevada Test Site (NTS) will accept low-level radioactive (LLW) and mixed waste (MW) for disposal. It includes requirements for the generator waste certification program, characterization, traceability, waste form, packaging, and transfer. The criteria apply to radioactive waste received at the NTS Area 3 and Area 5 Radioactive Waste Management Complex (RWMC) for storage or disposal.

Effect of low solids retention time and focused pulsed pre-treatment on anaerobic digestion of waste activated sludge.

PubMed

Lee, Il-Su; Rittmann, Bruce E

2011-02-01

The interacting effects of Focused Pulsed (FP) treatment and solids retention time (SRT) were evaluated in laboratory-scale digesters operated at SRTs of 2-20 days. Anaerobic digestion and methanogenesis of waste activated sludge (WAS) were stable for SRT ≥ 5 days, but the effluent soluble organic compounds increased significantly for SRT=2 days due to a combination of faster hydrolysis kinetics and washout of methanogens. FP treatment increased the CH(4) production rate and TCOD removal efficiency by up to 33% and 18%, respectively, at a SRT of 20 days. These effects were the result of an increase in the hydrolysis rate, since the concentrations of soluble components remained low for SRT ≥ 5 days. Alternately, FP pre-treatment of WAS allowed the same conversion of TCOD to CH(4) with a smaller SRT and digester size: e.g., 40% size savings with a CH(4) conversion of 0.23 g CH(4)-COD/g COD(in). Copyright © 2010 Elsevier Ltd. All rights reserved.

Waste heat recovery on multiple low-speed reciprocating engines

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mayhew, R.E.

1984-09-01

With rising fuel costs, energy conservation has taken on added significance. Installation of waste heat recovery units (WHRU's) on gas turbines is one method used in the past to reduce gas plant fuel consumption. More recently, waste heat recovery on multiple reciprocating compressor engines also has been identified as having energy conservation potential. This paper reviews the development and implementation of a WHRU for multiple low-speed engines at the Katy (TX) gas plant. WHRU's for these engines should be differentiated from high-speed engines and gas turbines in that low-speed engines produce low-frequency, high-amplitude pulsating exhaust. The design of a WHRUmore » system must take this potentially destructive pulsation into account. At Katy, the pulsation forces were measured at high-amplitude frequencies and then used to design a pulsation filter and structural stiffness into the various components of the WHRU to minimize vibration and improve system reliability.« less

Impeding 99Tc(IV) mobility in novel waste forms