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At least 883 records · Page 49
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Use of Separator with Regular Granular Filling in Extraction Processes - 12209

The separator with a regular granular filling has been developed to be used for the operations of the removal of carbon-black impurities from the aqueous flow; the intra-cycle regeneration of back-extractants; and the concentrating back-extraction of plutonium in the extraction SNF reprocessing technology. The process conditions for those operations have been experimentally tested using this separator. Thus, the separator for operations of concentrating reextraction of plutonium and regeneration recycling extractant allows to: - derive plutonium from the organic stream into an extremely concentrated form, reducing the amount of liquid radioactive waste (raffinate, decantate) plutonium branches to a minimum; - controlling the processes of concentration of uranium and plutonium, get uranium-plutonium product with required concentration and the amount of metal with required ratio; - eliminate contamination of the emulsion flows; - simplify the purification of uranium from the extract plutonium; - test results showed that when working on a solution 'hydrazine-DTPA' in the separator it is achieved the extraction of plutonium re-extract - 82% TC - at 56-65%. With the transition to the reextraction by solution 'U (IV)-hydrazine', of plutonium into re-extract increases to 95%, technetium - up to 83%; - regenerate the extractant recycling minimizing the amount of liquid radioactive waste generated as a technology used as well as alternative 'salt-free' systems; - it is found that the regeneration of the extractant solution of soda and EDA in the separator at a load of 3.2 m{sup 3} / (m{sup 2}.h) it can be achieved almost complete phase separation. Residual carryover of 0,005-0,006% emulsion is on the verge of analytical detection. - It is shown that in the depth of separation phase a separation of the extractant regeneration of quality superior is more than 5-7 times to the extraction. (authors)

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

DECOVALEX-2019 (Task E Final Report)

The DECOVALEX Project is an on-going international research collaboration, established in 1992, to advance the understanding and modeling of coupled Thermal (T), Hydrological (H), Mechanical (M) and Chemical (C) processes in geological in geological systems. DECOVALEX was initially motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. Later it was realized that these processes also play a critical role in other subsurface engineering activities, such as subsurface CO 2 storage, enhanced geothermal systems, and unconventional oil and gas production through hydraulic fracturing. Research teams from many countries (e.g., Canada, China, Czech Republic, Finland, France, Germany, Japan, Republic of Korea, Spain, Sweden, Switzerland, Taiwan, United Kingdom, and the United States) various institutions have participated in the DECOVALEX Project over the years, providing a wide range of perspectives and solutions to these complex problems. These institutions represent radioactive waste management organizations, national research institutes, regulatory agencies, universities, as well as industry and consulting groups. This document is the final report of Task E which was proposed and coordinated by Andra, the National Radioactive Waste Management Agency in France, presenting the technical definitions of the problems studied, approaches applied, achievements made and outstanding issues for future research. The purpose of Task E of the DECOVALEX-2019 project is to investigate upscaling THM modelling from small-scale experiments (some cubic meters) to full-scale experiments (some ten cubic meters) and finally to the scale of the waste repository (cubic kilometers). To achieve this aim, the data of two in-situ heating experiments performed by Andra (the French National Radioactive Waste Management Agency) in the Meuse/Haute-Marne Underground Research Laboratory (MHM URL) have formed the basis for the understanding of the THM behavior of the COx at different scales. The first experiment provided the reference values of the THM parameters by means of a calibration exercise and they were used for a blind prediction and an interpretative analysis of the second one.

58 GEOSCIENCES↗
DOE OSTI2018

Development of Novel Foaming Solutions for High Level Waste (HLW) Processing

Foaming occurs during treatment of high level waste (HLW) at Savannah River Site’s Defense Waste Processing Facility (DWPF) in the Chemical Process Cell (CPC) due to high gas generation from process steam and chemical off gas products. The presence of amphiphilic particles in the waste slurry stabilizes the foam. Efficient processing of HLW requires foam control, as foamovers lead to lower productivity and potential radioactive contamination of condensate streams. DWPF currently employs Antifoam 747, a surfactant produced by Momentive Performance Materials, as an antifoaming agent during waste treatment. While it controls foam, processing issues have arisen from its use. During DWPF chemical processing, antifoam must be effective up to 103°C and between a pH range of 3-13. Antifoam 747 is efficient at pH 6-8 but degrades outside this optimal pH range. Savannah River National Laboratory (SRNL) has identified multiple flammable antifoam degradation products during laboratory scale experiments.

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

Radioactive Waste Management and Nuclear Facility Decommissioning Progress in Iraq - 13216

Management of Iraq's radioactive wastes and decommissioning of Iraq's former nuclear facilities are the responsibility of Iraq's Ministry of Science and Technology (MoST). The majority of Iraq's former nuclear facilities are in the Al-Tuwaitha Nuclear Research Center located a few kilometers from the edge of Baghdad. These facilities include bombed and partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. Within these facilities are large numbers of silos, approximately 30 process or waste storage tanks and thousands of drums of uncharacterised radioactive waste. There are also former nuclear facilities/sites that are outside of Al-Tuwaitha and these include the former uranium processing and waste storage facility at Jesira, the dump site near Adaya, the former centrifuge facility at Rashdiya and the former enrichment plant at Tarmiya. In 2005, Iraq lacked the infrastructure needed to decommission its nuclear facilities and manage its radioactive wastes. The lack of infrastructure included: (1) the lack of an organization responsible for decommissioning and radioactive waste management, (2) the lack of a storage facility for radioactive wastes, (3) the lack of professionals with experience in decommissioning and modern waste management practices, (4) the lack of laws and regulations governing decommissioning or radioactive waste management, (5) ongoing security concerns, and (6) limited availability of electricity and internet. Since its creation eight years ago, the MoST has worked with the international community and developed an organizational structure, trained staff, and made great progress in managing radioactive wastes and decommissioning Iraq's former nuclear facilities. This progress has been made, despite the very difficult implementing conditions in Iraq. Within MoST, the Radioactive Waste Treatment and Management Directorate (RWTMD) is responsible for waste management and the Iraqi Decommissioning Directorate (IDD) is responsible for decommissioning activities. The IDD and the RWTMD work together on decommissioning projects. The IDD has developed plans and has completed decommissioning of the GeoPilot Facility in Baghdad and the Active Metallurgical Testing Laboratory (LAMA) in Al-Tuwaitha. Given this experience, the IDD has initiated work on more dangerous facilities. Plans are being developed to characterize, decontaminate and decommission the Tamuz II Research Reactor. The Tammuz Reactor was destroyed by an Israeli air-strike in 1981 and the Tammuz II Reactor was destroyed during the First Gulf War in 1991. In addition to being responsible for managing the decommissioning wastes, the RWTMD is responsible for more than 950 disused sealed radioactive sources, contaminated debris from the first Gulf War and (approximately 900 tons) of naturally-occurring radioactive materials wastes from oil production in Iraq. The RWTMD has trained staff, rehabilitated the Building 39 Radioactive Waste Storage building, rehabilitated portions of the French-built Radioactive Waste Treatment Station, organized and secured thousands of drums of radioactive waste organized and secured the stores of disused sealed radioactive sources. Currently, the IDD and the RWTMD are finalizing plans for the decommissioning of the Tammuz II Research Reactor. (authors)

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

Radioactive Waste Management in a Treatability and Optimisation Perspective - 17527

In order to meet recycling criteria or to fulfil Waste Acceptance Criteria (WAC) for disposal there are many different management concepts for radioactive waste that can be applied. Which concept or set of concepts to use to optimize waste management depends on several factors, among them, the regulatory framework for recycling, the availability of a disposal facility and the acceptance criteria for disposal. The required pre-treatment and other pre-disposal activities should be carefully analyzed for each of the disposal options, as they are significant contributors to the total cost for managing the waste. Waste Acceptance Criteria are living documents that should be updated when regulatory framework changes as well as when new treatment methods and new packaging and barrier systems develop. Each waste treatment facility, has its own specific limitations and therefore its own WAC/CFA (conditions for acceptance), which depends both on the plant design and national legislation. Many waste owners, countries or licensees, have their preferred, and in some cases well established, waste routes for certain waste streams. In some countries, the full chain from waste treatment through to disposal of radioactive waste is in place, whilst other countries still are in the process of establishing their repository program. The worst case scenario for any licensee or waste owner is waste which is not possible to qualify for disposal. This can be due to poor or degraded waste containers, unspecified content in waste packages, or wastes that are not suitable for disposal in their current form. Such deficiencies result in expensive actions related to waste retrieval, new characterisation and/or costly investigations. The EDF Group has a fleet of waste treatment facilities and has, for decades, provided waste treatment services to domestic and international customers. This paper will discuss important waste owner considerations and EDF experiences related to the optimization of the waste management process. This will be from both a treatment and an end-state perspective. (authors)

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

NPP SCCs Based Deregulation Process Application Case Design with Internal Block Diagram of SysML - 19237

- It will be possible to group the radioactive wastes from the perspectives on Systems, Structures and Components (SSCs). This paper covers systematic approaches for deregulating grouped decommissioning wastes. SSCs decomposition structure, mass flow of decommissioning waste, sequence and storage time estimation method for deregulation, and interrelationship between them are presented. The expected benefit is the ability possibly to apply deregulation process to SSCs because SSCs breakdown structure is connected with mass flow of decommissioning waste, deregulation option, and storage time. This allows for the application of deregulation strategies and processes on SSCs basis and can enhance the rationale and traceability of the waste grouping. The sequence for deregulation reflects the logic of ensuring that radioactive nuclides meet self-disposal criteria and storing or disposing radioactive wastes if not met. (authors)

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

Near-Tank Treatment of Hanford Tank Waste: Pilot-Scale Testing - 12107

In order to reduce the number of high level waste canisters that will have to be produced by the Hanford Tank Waste Treatment and Immobilization Plant, supplemental waste treatment technologies are being investigated. One such technology is the Near-Tank Treatment System, which uses continuous sludge leaching, filtration, and ion exchange to process the waste in a simple, compact system. The Near Tank Treatment System is ideally suited for handling Hanford tanks with large amounts of boehmite, a difficult aluminum phase to dissolve. A pilot-scale Near Tank Treatment System was constructed and tested with a boehmite and iron oxyhydroxide waste simulant to evaluate the robustness and effectiveness of the system. The data from the pilot-scale tests were also used to assess scale-up from previously performed bench-scale tests. It should be noted that any work involving the use of simulated HLW has inherent limitations. For this work, every attempt was made to ensure that the simulant mimicked the actual waste performance as closely as possible. However, there are always limitations in the ability to The NTTS Demonstration Test established that all the subsystems could be operated concurrently to process the waste simulant. The NTTS had a stable performance during the Demonstration Test that required very little external manipulation once a steady state was achieved. The CSL demonstrated effectively 90% dissolution of the hard to leach boehmite phase. The cross flow filtration system was able to deliver decontaminated salt solution to the NTCR system. The NTCR system exceeded the design basis by providing effluent below the detection limit beyond the design basis required time. The NTTS system was successfully demonstrated and has shown that leaching and decontamination of the subsequent permeate can be readily achieved using relatively small footprint equipment in an at tank application. (authors)

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

Analytical Results from Salt Batch 10 Routine DSSHT and SEHT Monthly Samples

Strip Effluent (SE) Hold Tank (SEHT) and Decontaminated Salt Solution (DSS) Hold Tank (DSSHT) samples from several “microbatches” of the Integrated Salt Disposition Project (ISDP) “macrobatch” Salt Batch 10 have been analyzed for 238 Pu, 90 Sr, 137 Cs, cations (via Inductively Coupled Plasma Emission Spectroscopy - ICPES), and in some cases anions (via Ion Chromatography Anions - IC-A). The analytical results from the current microbatch samples are similar to those from previous macrobatches. In the Modular Caustic-Side Solvent Extraction Unit (MCU), the Cs removal continues to exceed process requirements. The bulk chemistry of the DSSHT and SEHT samples do not show any signs of unusual behavior except for slightly elevated amounts of sodium in the SE samples. This summary represents the final Salt Batch processed through MCU prior to commencement of process and transfer line tie-ins to support the startup of the Salt Waste Processing Facility (SWPF).

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

Radiation Stability of Benzyl Tributyl Ammonium Chloride towards Technetium-99 Extraction - 13016

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning processes, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, the UREX process has been developed in the United States to co-extract hexavalent uranium (U) and hepta-valent technetium (Tc) by tri-n-butyl phosphate (TBP). Tc-99 is recognized to be one of the most abundant, long-lived radio-toxic isotopes in UNF (half-life, t{sub 1/2} = 2.13 x 10{sup 5} years), and as such, is targeted in UNF separation strategies for isolation and encapsulation in solid waste-forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste-form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flowsheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste-forms for ultimate disposal. In addition, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste-forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macro-compounds such as crown-ethers, aza-crown ethers, quaternary ammonium salts, and resorcin-arenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO{sub 4}{sup -} by benzyl tributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligand's matrix conditions and concentration, as well as varying the organic phase composition (i.e. diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using an external Co-60 source. Post-irradiation solvent extraction measurements will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

International Collaboration Activities on Engineered Barrier Systems

The Used Fuel Disposition Campaign (UFDC) within the DOE Fuel Cycle Technologies (FCT) program has been engaging in international collaborations between repository R&D programs for high-level waste (HLW) disposal to leverage on gathered knowledge and laboratory/field data of near- and far-field processes from experiments at underground research laboratories (URL). Heater test experiments at URLs provide a unique opportunity to mimetically study the thermal effects of heat-generating nuclear waste in subsurface repository environments. Various configurations of these experiments have been carried out at various URLs according to the disposal design concepts of the hosting country repository program. The FEBEX (Full-scale Engineered Barrier Experiment in Crystalline Host Rock) project is a large-scale heater test experiment originated by the Spanish radioactive waste management agency (Empresa Nacional de Residuos Radiactivos S.A. – ENRESA) at the Grimsel Test Site (GTS) URL in Switzerland. The project was subsequently managed by CIEMAT. FEBEX-DP is a concerted effort of various international partners working on the evaluation of sensor data and characterization of samples obtained during the course of this field test and subsequent dismantling. The main purpose of these field-scale experiments is to evaluate feasibility for creation of an engineered barrier system (EBS) with a horizontal configuration according to the Spanish concept of deep geological disposal of high-level radioactive waste in crystalline rock. Another key aspect of this project is to improve the knowledge of coupled processes such as thermal-hydro-mechanical (THM) and thermal-hydro-chemical (THC) operating in the near-field environment. The focus of these is on model development and validation of predictions through model implementation in computational tools to simulate coupled THM and THC processes.

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

Adsorption of Iodine and Tritium from UNF Reprocessing Off-Gas Streams

The need to control emissions of the gaseous radionuclides to the air during nuclear fuel recycling is widely recognized. Gaseous radionuclides can partition to off-gas streams emanating from the various individual processes used in the recycle plant, such as from the mechanical head end (shear), tritium pretreatment, fuel dissolver, vessel vents, cell ventilation, and conversion operations. In traditional recycle plants, the bulk of the four primary volatile fission products (i.e., {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I) is released in the head-end processes, for example, the dissolution process. The usual off-gas treatment sequence of operations targets removal of tritium (as {sup 3}HHO), iodine, carbon (as {sup 14}CO{sub 2}), and the noble gases, in that order. Water and iodine are captured first because the other species are relatively unimpeded by the sorbents used for water and iodine, so they flow downstream to additional treatment processes. Water and iodine do react with one another, and many techniques used to collect one of these species also collect some fraction of the other. It is quite desirable to sequester the iodine and water separately because the tritiated water may be disposed of as a low level waste. If not cleanly separated, contamination of the low-level waste by iodine may elevate its classification to high-level waste and result in greater costs for disposal. The proposed sorption sequence was tested in the laboratory. Use of a second column of AgZ permits the detection of breakthrough from the primary AgZ column while protecting the downstream 3AMS column from collecting iodine. (This arrangement is equivalent to enabling intra-column sampling of the primary column.) Neutron activation analysis of the 3AMS revealed that iodine was below, or at, the level of detection. The results successfully demonstrated that iodine can be sequestered in the AgZ while the water is collected free of iodine in the 3AMS column. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Waste Determination Equivalency - 12172

The Savannah River Site (SRS) is a Department of Energy (DOE) facility encompassing approximately 800 square kilometers near Aiken, South Carolina which began operations in the 1950's with the mission to produce nuclear materials. The SRS contains fifty-one tanks (2 stabilized, 49 yet to be closed) distributed between two liquid radioactive waste storage facilities at SRS containing carbon steel underground tanks with storage capacities ranging from 2,800,000 to 4,900,000 liters. Treatment of the liquid waste from these tanks is essential both to closing older tanks and to maintaining space needed to treat the waste that is eventually vitrified or disposed of onsite. Section 3116 of the Ronald W. Reagan National Defense Authorization Act of Fiscal Year 2005 (NDAA) provides the Secretary of Energy, in consultation with the Nuclear Regulatory Commission (NRC), a methodology to determine that certain waste resulting from prior reprocessing of spent nuclear fuel are not high-level radioactive waste if it can be demonstrated that the waste meets the criteria set forth in Section 3116(a) of the NDAA. The Secretary of Energy, in consultation with the NRC, signed a determination in January 2006, pursuant to Section 3116(a) of the NDAA, for salt waste disposal at the SRS Saltstone Disposal Facility. This determination is based, in part, on the Basis for Section 3116 Determination for Salt Waste Disposal at the Savannah River Site and supporting references, a document that describes the planned methods of liquid waste treatment and the resulting waste streams. The document provides descriptions of the proposed methods for processing salt waste, dividing them into 'Interim Salt Processing' and later processing through the Salt Waste Processing Facility (SWPF). Interim Salt Processing is separated into Deliquification, Dissolution, and Adjustment (DDA) and Actinide Removal Process/Caustic Side Solvent Extraction Unit (ARP/MCU). The Waste Determination was signed by the Secretary of Energy in January of 2006 based on proposed processing techniques with the expectation that it could be revised as new processing capabilities became viable. Once signed, however, it became evident that any changes would require lengthy review and another determination signed by the Secretary of Energy. With the maturation of additional salt removal technologies and the extension of the SWPF start-up date, it becomes necessary to define 'equivalency' to the processes laid out in the original determination. For the purposes of SRS, any waste not processed through Interim Salt Processing must be processed through SWPF or an equivalent process, and therefore a clear statement of the requirements for a process to be equivalent to SWPF becomes necessary. (authors)

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

Sustainable Management of Flowback Water during Hydraulic Fracturing of Marcellus Shale for Natural Gas Production

This study evaluated the feasibility of using abandoned mine drainage (AMD) as make- up water for the reuse of produced water for hydraulic fracturing. There is an abundance of AMD sources near permitted gas wells as documented in this study that can not only serve as makeup water and reduce the demand on high quality water resources but can also as a source of chemicals to treat produced water prior to reuse. The assessment of AMD availability for this purpose based on proximity and relevant regulations was accompanied by bench- and pilot-scale studies to determine optimal treatment to achieve desired water quality for use in hydraulic fracturing. Sulfate ions that are often present in AMD at elevated levels will react with Ba²⁺ and Sr²⁺ in produced water to form insoluble sulfate compounds. Both membrane microfiltration and gravity separation were evaluated for the removal of solids formed as a result of mixing these two impaired waters. Laboratory studies revealed that neither AMD nor barite formed in solution had significant impact on membrane filtration but that some produced waters contained submicron particles that can cause severe fouling of microfiltration membrane. Coagulation/flocculation was found to be an effective process for the removal of suspended solids and both bench- and pilot-scale studies revealed that optimal process conditions can consistently achieve the turbidity of the finished water below 5 NTU. Adjusting the blending ratio of AMD and produced water can achieve the desired effluent sulfate concentration that can be accurately predicted by chemical thermodynamics. Co-treatment of produced water and AMD will result in elevated levels of naturally occurring radioactive materials (NORM) in the solid waste generated in this process due to radium co-precipitation with barium sulfate. Laboratory studies revealed that the mobility of barite that may form in the subsurface due to the presence of sulfate in the fracturing fluid can be controlled by the addition of appropriate antiscalants.

03 NATURAL GAS↗
DOE OSTI2011

Final Report Summary of DM1200 Operation at VSL

The principal objective of this report was to summarize the testing experience on the DuraMelter 1200 (DMI200), which is the High Level Waste (HLW) Pilot Melter located at the Vitreous State Laboratory (VSL). Further objectives were to provide descriptions of the history of all modifications and maintenance, methods of operation, problems and unit failures, and melter emissions and performance while processing a variety of simulated HL W and low activity waste (LAW) feeds for the Hanford Waste Treatment and Immobilization Plant (WTP) and employing a variety of operating methods. All of these objectives were met. The River Protection Project - Hanford Waste Treatment and Immobilization Plant (RPP-WTP) Project has undertaken a 'tiered' approach to vitrification development testing involving computer-based glass formulation, glass property-composition models, crucible melts, and continuous melter tests of increasing, more realistic scales. Melter systems ranging from 0.02 to 1.2 m{sup 2} installed at the Vitreous State Laboratory (VSL) have been used for this purpose, which, in combination with the 3.3 m{sup 2} low activity waste (LAW) Pilot Melter at Duratek, Inc., span more than two orders of magnitude in melt surface area. In this way, less-costly small-scale tests can be used to define the most appropriate tests to be conducted at the larger scales in order to extract maximum benefit from the large-scale tests. For high level waste (HLW) vitrification development, a key component in this approach is the one-third scale DuraMelter 1200 (DM 1200), which is the HLW Pilot Melter that has been installed at VSL with an integrated prototypical off-gas treatment system. That system replaced the DM1000 system that was used for HLW throughput testing during Part B1. Both melters have similar melt surface areas (1.2 m{sup 2}) but the DM1200 is prototypical of the present RPP-WTP HLW melter design whereas the DM1000 was not. In particular, the DM1200 provides for testing on a vitrification system with the specific train of unit operations that has been selected for both HLW and LAW RPP-WTP off-gas treatment.

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

Deep Vadose Zone-Applied Field Research Initiative Fiscal Year 2011 Annual Report

This annual report describes the background of the Deep Vadose Zone-Applied Field Research Initiative, and some of the programmatic approaches and transformational technologies in groundwater and deep vadose zone remediation developed during fiscal year 2011. The Department of Energy (DOE) Office of Technology Innovation and Development's (OTID) mission is to transform science into viable solutions for environmental cleanup. In 2010, OTID developed the Impact Plan, Science and Technology to Reduce the Life Cycle Cost of Closure to outline the benefits of research and development of the lifecycle cost of cleanup across the DOE complex. This plan outlines OTID's ability to reduce by $50 billion, the $200 billion life-cycle cost in waste processing, groundwater and soil, nuclear materials, and deactivation and decommissioning. The projected life-cycle costs and return on investment are based on actual savings realized from technology innovation, development, and insertion into remedial strategies and schedules at the Fernald, Mound, and Ashtabula sites. To achieve our goals, OTID developed Applied Field Research Initiatives to facilitate and accelerate collaborative development and implementation of new tools and approaches that reduce risk, cost and time for site closure. The primary mission of the Deep Vadose Zone-Applied Field Research Initiative (DVZ-AFRI) is to protect our nation's water resources, keeping them clean and safe for future generations. The DVZ-AFRI was established for the DOE to develop effective, science-based solutions for remediating, characterizing, monitoring, and predicting the behavior and fate of deep vadose zone contamination. Subsurface contaminants include radionuclides, metals, organics, and liquid waste that originated from various sources, including legacy waste from the nation's nuclear weapons complexes. The DVZ-AFRI project team is translating strategy into action by working to solve these complex challenges in a collaborative environment that leverages technology and scientific expertise from DOE, Pacific Northwest National Laboratory, CH2M HILL Plateau Remediation Company, and the broad scientific research community. As project manager for the DVZ-AFRI, I have had the privilege this past year to team with creative, talented members of the scientific community nationwide to develop effective long-term solutions to address deep vadose zone contamination. This report highlights how the DVZ-AFRI project team is delivering results by achieving significant programmatic accomplishments, and developing and field-testing transformational technologies to address the nation's most pressing groundwater and vadose zone contamination problems.

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

Goethite Bench-scale and Large-scale Preparation Tests

The Hanford Waste Treatment and Immobilization Plant (WTP) is the keystone for cleanup of high-level radioactive waste from our nation's nuclear defense program. The WTP will process high-level waste from the Hanford tanks and produce immobilized high-level waste glass for disposal at a national repository, low activity waste (LAW) glass, and liquid effluent from the vitrification off-gas scrubbers. The liquid effluent will be stabilized into a secondary waste form (e.g. grout-like material) and disposed on the Hanford site in the Integrated Disposal Facility (IDF) along with the low-activity waste glass. The major long-term environmental impact at Hanford results from technetium that volatilizes from the WTP melters and finally resides in the secondary waste. Laboratory studies have indicated that pertechnetate ({sup 99}TcO{sub 4}{sup -}) can be reduced and captured into a solid solution of {alpha}-FeOOH, goethite (Um 2010). Goethite is a stable mineral and can significantly retard the release of technetium to the environment from the IDF. The laboratory studies were conducted using reaction times of many days, which is typical of environmental subsurface reactions that were the genesis of this new process. This study was the first step in considering adaptation of the slow laboratory steps to a larger-scale and faster process that could be conducted either within the WTP or within the effluent treatment facility (ETF). Two levels of scale-up tests were conducted (25x and 400x). The largest scale-up produced slurries of Fe-rich precipitates that contained rhenium as a nonradioactive surrogate for {sup 99}Tc. The slurries were used in melter tests at Vitreous State Laboratory (VSL) to determine whether captured rhenium was less volatile in the vitrification process than rhenium in an unmodified feed. A critical step in the technetium immobilization process is to chemically reduce Tc(VII) in the pertechnetate (TcO{sub 4}{sup -}) to Tc(Iv)by reaction with the ferrous ion, Fe{sup 2+}-Fe{sup 2+} is oxidized to Fe{sup 3+} - in the presence of goethite seed particles. Rhenium does not mimic that process; it is not a strong enough reducing agent to duplicate the TcO{sub 4}{sup -}/Fe{sup 2+} redox reactions. Laboratory tests conducted in parallel with these scaled tests identified modifications to the liquid chemistry necessary to reduce ReO{sub 4}{sup -} and capture rhenium in the solids at levels similar to those achieved by Um (2010) for inclusion of Tc into goethite. By implementing these changes, Re was incorporated into Fe-rich solids for testing at VSL. The changes also changed the phase of iron that was in the slurry product: rather than forming goethite ({alpha}-FeOOH), the process produced magnetite (Fe{sub 3}O{sub 4}). Magnetite was considered by Pacific Northwest National Laboratory (PNNL) and VSL to probably be a better product to improve Re retention in the melter because it decomposes at a higher temperature than goethite (1538 C vs. 136 C). The feasibility tests at VSL were conducted using Re-rich magnetite. The tests did not indicate an improved retention of Re in the glass during vitrification, but they did indicate an improved melting rate (+60%), which could have significant impact on HLW processing. It is still to be shown whether the Re is a solid solution in the magnetite as {sup 99}Tc was determined to be in goethite.

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

Current Status of Decommissioning Plans and Chemical Decontamination of Metal Scraps in Chinshan NPP - 19304

The decommissioning of nuclear facilities has become a significantly attractive and worldwide topic due to the end of lifetime of many nuclear power plants within next coming years. In Taiwan, Chinshan nuclear power plant owned by the Taiwan Power Company will be the first candidate to be decommissioned according to the Nuclear Reactor Facilities Regulation Act. The Chinshan nuclear power plant located in New Taipei City has two units of 636 MWe boiling water reactors (BWRs) for the generation of 9 billion kWh of electricity per year. The date of cessation of operation of Unit No. 1 and Unit No. 2 will be December in 2018 and July in 2019, respectively. Based on the Regulations for the Review and Approval of Applications for a Decommissioning Permit of Nuclear Reactor Facilities, Taiwan Power Company have submitted a decommissioning plan of Chinshan nuclear power plant to the authority, Atomic Energy Council (AEC). After the review process, the decommissioning plan of Chinshan nuclear power plant has been permitted by AEC in 2017. Decontamination is generally utilized during the period of decommissioning of nuclear facility in order to minimize the radiation exposure of personnel and the cost of management of radioactive wastes. The successful decontamination also probably provides the benefits of lowering the level of radioactive wastes, facilitating potential releases, permitting the reuse of metals and facilitating the radiation principle of As Low As Reasonably Achievable (ALARA) management. Decontamination aims at the removal of radioactive materials from the contaminated surfaces by several technologies such as washing, heating, chemical reaction and mechanical process. The selection of proper decontamination technologies are depended on the requirement of the target materials and the purpose of the decontamination. Chemical decontamination has been proven to be an effective in reducing the radioactivity for metallic materials by using chemical reagents in contact with the contaminated surfaces. During the operation of Chinshan nuclear power plant, chemical decontamination was performed for Unit No.1 and No.2 in 2000 to decontaminate the Reactor Recirculation System (RRS) and Reactor Water Clean-up Systems (RWCU) by using LOMI-NP-LOMI process. Based on the previous experience of chemical decontamination, the dismantled metallic piping is provided as sample by Chinshan nuclear power plant for the evaluation of the feasibility of chemical decontamination. The provided metallic sample is composed of carbon steel with dose rate around 0.2 mSv/hr. In this report, a multistep decontamination process using strongly oxidizing solution and organic acid solution is performed to remove the contaminated oxide layer from the metal surface, and the resulting contact dose rate and decontamination factor (DF) are also shown to discuss the performance in relation to the process control factors. In this paper, the recent progress in decommissioning programs and preliminary results of the decontamination experiment are described to provide usable information regarding the future decommissioning for Chinshan nuclear power plant. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

OWL Change Control Process

The Online Waste Library (OWL) provides a consolidated source of information on Department of Energy-managed radioactive waste likely to require deep geologic disposal. With the release of OWL Version 1.0 in fiscal year 2019 (FY2019), much of the FY2020 work involved developing the OWL change control process and the OWL release process. These two processes (in draft form) were put into use for OWL Version 2.0, which was released in early FY2021. With the knowledge gained, the OWL team refined and documented the two processes in two separate reports. This report focuses on the change control process and discusses the following: (1) definitions and system components; (2) roles and responsibilities; (3) origin of changes; (4) the change control process including the Change List, Task List, activity categories, implementation examples, and checking and review; and (5) the role of the re lease process in ensuring changes in the Change List are incorporated into a public release.

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