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At least 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Characterization of the SRNL-Washed tank 51 sludge batch 9 qualification sample

Savannah River National Laboratory (SRNL) personnel have been requested to qualify the next sludge batch (Sludge Batch 9 – SB9) for processing at the Defense Waste Processing Facility (DWPF). To accomplish this task, Savannah River Remediation (SRR) sent SRNL a 3-L sample of Tank 51H slurry to be characterized, washed, and then used in a lab-scale demonstration of the DWPF flowsheet (after combining with Tank 40H sludge). SRNL has washed the Tank 51H sample per the Tank Farm washing strategy as of October 20, 2015. A part of the qualification process is extensive radionuclide and chemical characterization of the SRNL-washed Tank 51H slurry. This report documents the chemical characterization of the washed slurry; radiological characterization is in progress and will be documented in a separate report. The analytical results of this characterization are comparable to the Tank Farm projections. Therefore, it is recommended that SRNL use this washed slurry for the ongoing SB9 qualification activities.

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DOE OSTI2016

Liquid secondary waste. Waste form formulation and qualification

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 testing 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.

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DOE OSTI2016

Antifoam Degradation Products in Off Gas and Condensate of Sludge Batch 9 Simulant Nitric-Formic Flowsheet Testing for the Defense Waste Processing Facility

Ten chemical processing cell (CPC) experiments were performed using simulant to evaluate Sludge Batch 9 for sludge-only and coupled processing using the nitric-formic flowsheet in the Defense Waste Processing Facility (DWPF). Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles were performed on eight of the ten. The other two were SRAT cycles only. Samples of the condensate, sludge, and off gas were taken to monitor the chemistry of the CPC experiments. The Savannah River National Laboratory (SRNL) has previously shown antifoam decomposes to form flammable organic products, (hexamethyldisiloxane (HMDSO), trimethylsilanol (TMS), and propanal), that are present in the vapor phase and condensate of the CPC vessels. To minimize antifoam degradation product formation, a new antifoam addition strategy was implemented at SRNL and DWPF to add antifoam undiluted.

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DOE OSTI2016

Results of Hg speciation testing on DWPF SMECT-8, OGCT-1, AND OGCT-2 samples

The Savannah River National Laboratory (SRNL) was tasked with preparing and shipping samples for Hg speciation by Eurofins Frontier Global Sciences, Inc. in Seattle, WA on behalf of the Savannah River Remediation (SRR) Mercury Task Team. The sixteenth shipment of samples was designated to include a Defense Waste Processing Facility (DWPF) Slurry Mix Evaporator Condensate Tank (SMECT) sample from Sludge Receipt and Adjustment Tank (SRAT) Batch 738 processing and two Off-Gas Condensate Tank (OGCT) samples, one following Batch 736 and one following Batch 738. The DWPF sample designations for the three samples analyzed are provided. The Batch 738 ‘End of SME Cycle’ SMECT sample was taken at the conclusion of Slurry Mix Evaporator (SME) operations for this batch and represents the fourth SMECT sample examined from Batch 738. Batch 738 experienced a sludge slurry carryover event, which introduced sludge solids to the SMECT that were particularly evident in the SMECT-5 sample, but less evident in the ‘End of SME Cycle’ SMECT-8 sample.

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DOE OSTI2013

Ion Exchange Column Tests Supporting Technetium Removal Resin Maturation

The primary treatment of the tank waste at the DOE Hanford site will be done in the Waste Treatment and Immobilization Plant, currently under construction. The baseline plan for this facility is to treat the waste, splitting it into High Level Waste (HLW) and Low Activity Waste (LAW). Both waste streams are then separately vitrified as glass and sealed in canisters. The LAW glass will be disposed on site. There are currently no plans to treat the waste to remove technetium, so its disposition path is the LAW glass. Due to the soluble properties of pertechnetate and long half-life of 99 Tc, effective management of 99 Tc is important. Options are being explored to immobilize the supplemental LAW portion of the tank waste, as well as to examine the volatility of 99 Tc during the vitrification process. Removal of 99 Tc, followed by off-site disposal has potential to reduce treatment and disposal costs. A conceptual flow sheets for supplemental LAW treatment and disposal that could benefit from technetium removal will specifically examine removing 99 Tc from the LAW feed stream to supplemental immobilization. SuperLig® 639 is an elutable ion exchange resin. In the tank waste, 99 Tc is predominantly found in the tank supernate as pertechnetate (TcO 4 - ). Perrhenate (ReO 4 - ) has been shown to be a good non-radioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin. This report contains results of experimental ion exchange distribution coefficient and column resin maturation kinetics testing using the resin SuperLig® 639a to selectively remove perrhenate from simulated LAW. This revision includes results from testing to determine effective resin operating temperature range. Loading tests were performed at 45°C, and the computer modeling was updated to include the temperature effects. Equilibrium contact testing indicated that this batch of SuperLig® 639 resin has good performance, with an average perrhenate distribution coefficient of 291 mL/g at a 100:1 phase ratio. This slightly exceeds the computer-modeled equilibrium distribution. The modeling agreed well with the experimental data for perrhenate removal with minor adjustments. Predicted breakthrough performance was on average within about 20% of measured values.

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DOE OSTI2016

Deadtime Calculations of Liquid Scintillator Neutron Detector Using Attenuation Law

Scientists have been working on deadtime problems for radiation detectors since 1940's. In detector systems, a minimum time must separate two different events before they can be measured individually. That minimum time interval is called to be deadtime of detector system. There are two commonly deadtime models: the paralyzing and the non-paralyzing models. Paralyzing model assumes that each event during the deadtime will reset the deadtime and deadtime extension will depend on the count rate. According to the non-paralyzing model, deadtime is fixed after each detected event and all events occurring during deadtime are lost. In this study, deadtime measurements have been obtained for an N-Probe fast neutron (NE-213 Liquid Scintillation) detector manufactured by BTI (Bubble Technology Industries) using a combination of the attenuation law and the assumption of ideal paralyzing and non-paralyzing method. Bubble Technology Industries (BTI) has manufactured a portable neutron scintillation spectrometer (N-Probe) with potential applications at nuclear reactor facilities, waste processing operations, and spent fuel storage areas. The deadtime of the N-Probe detector is not provided by the manufacturer; BTI, and there is no other study published on the N-Probe detector' deadtime. Since the detector will be used in future nuclear fuel investigations, it is important to determine the detector' deadtime for accurate nuclear fuel analysis. The deadtime of the BTI N-Probe was examined using different thicknesses of Plexiglas at the Missouri University of Science and Technology Reactor (MSTR) using attenuation law and average macroscopic cross-section of Plexiglas defined by both paralyzing and non-paralyzing model. (authors)

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DOE OSTI2020

Challenges and Solutions in Operation and Lay-Up of the Modular Caustic- Side Solvent Extraction Unit - 20078

The Modular Caustic-Side Solvent Extraction (CSSX) Unit (MCU) was designed and constructed to provide interim salt waste processing so that the Liquid Waste Disposition Program (LWDP) could continue waste removal and tank closure until start-up of the Salt Waste Processing Facility (SWPF). MCU processes Clarified Salt Solution (CSS), or salt solution (SS) that has undergone actinide removal/filtration, to produce two waste streams: the cesium laden Strip Effluent (SE), which is incorporated into glass at the Defense Waste Processing Facility (DWPF), and the cesium depleted Decontaminated SS (DSS), which is incorporated into grout at the Saltstone Production Facility (SPF). MCU completed start-up in 2008 with an initial operating life of 3 years and design life of 5 years. Prior to reaching the end of the design life, multiple critical components were upgraded and/or replaced to mitigate risks associated with any delays in SWPF start-up. During the remainder of operation (until June 2019), MCU was challenged to maximize processing with minimal additional modifications to repair/improve aging infrastructure, while maintaining low levels of risk to workers and the environment. Notable challenges during the extended operating life of the facility involved both mechanical and operational issues. Two of the most significant issues were the increased frequency of Process Vessel Ventilation (PVV) High Efficiency Particulate Air (HEPA) filter change-outs and the biofouling of the SE coalescer (SEC). The major impacts of both issues were that replacement of these components 1) resulted in high exposure to workers, and 2) required substantial downtime, thus hindering achievement of processing goals. PVV HEPA filter replacements are required based on dose rate and differential pressure (dP) limits. After introducing a higher-curie feed to MCU, the operating time between change outs was reduced by approximately half. SEC replacements are required based on dP limits. After restarting from an extended outage, the operating life of the SEC was significantly reduced by more than half due to biofouling. As a result of extensive troubleshooting and process improvements, the facility was able to recover performance and extend the operating life of the PVV HEPA filters and the SEC media. Through troubleshooting both of these issues, the facility was also able to learn from prior experience and adapt in order to minimize down time and maintain throughput. Although the facility took advantage of short periods of downtime to perform nonintrusive troubleshooting and minor corrective/preventative maintenance activities, longer outages were periodically required for corrective maintenance involving process cell entry. During these outages, remote tools and specialized shield plates were employed to minimize dose to workers. Mock-ups of non-routine maintenance activities were critical to early identification of potential issues and improvements, which shortened outage duration. Consistent with the facility's operational strategy, evolutions needed for MCU layup were identified and sequenced to ensure safe conditions within the facility, meet pre-defined criteria for lay-up configuration, and protect workers, while minimizing the impact to SWPF integration. De-inventory and flushing of process areas was prioritized so that the total residual inventory was reduced as efficiently as possible. This paper discusses lessons learned and best practices from troubleshooting unique issues, performing complex maintenance activities, and implementing a layup strategy that best supports the overall system mission. (authors)

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DOE OSTI2019

Panel Session 51: Idaho Transuranic Waste Challenges, Solutions and Implications for the Future (3/5)

This panel focused on the retrieval, characterization, treatment, certification, and shipping of transuranic (TRU) waste at the Idaho National Laboratory (INL) Site, Idaho, USA. The INL began receiving TRU waste from offsite generators (primarily the Rocky Flats Plant in Colorado, USA) beginning in 1954. Until 1970, the INL Site disposed of barrels, boxes, and other debris in a shallow, unlined landfill. Beginning in 1970, the INL Site began storing TRU waste on an asphalt pad, which was then covered with several feet of soil. This practice continued until the late 1980's at which time a large air support building was erected over the waste. In the mid-to-late 1990's, the air support building was removed and a massive building was constructed over the entire dirt mound. Contractors retrieved the waste from the early 2000's through February 2017 to support an Idaho Settlement Agreement milestone that required the retrieval of 65,000 cubic meters of TRU waste and removal from Idaho. This panel focused on the Idaho Transuranic Waste Challenges, Solutions and Implications for the Future. The Idaho Transuranic Waste Program includes buried and stored waste retrieval and processing of contact- and remote-handled transuranic waste. Over the duration of the program, the Idaho site has conducted targeted retrieval of buried waste from the 1950's and 1960's; designed, permitted, built, operated, and completed the mission of a facility to process 65,000 cubic meters of waste; and shipped more than 57,000 cubic meters of waste to WIPP. Presentation proposed: Transuranic Waste Challenges and Solutions (Hoss Brown, Fluor Idaho, LLC (USA), John McCoy, Fluor Idaho, LLC (USA), James Malmo, US DOE (USA), Doug Pruitt, Department of Energy Idaho Field Office (USA), Kenneth Princen, US DOE (USA))

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DOE OSTI2016

PLAN-TA9-2443(U), Rev. B Remediated Nitrate Salt (RNS) Surrogate Formulation and Testing Standard Procedure

This document identifies scope and some general procedural steps for performing Remediated Nitrate Salt (RNS) Surrogate Formulation and Testing. This Test Plan describes the requirements, responsibilities, and process for preparing and testing a range of chemical surrogates intended to mimic the energetic response of waste created during processing of legacy nitrate salts. The surrogates developed are expected to bound1 the thermal and mechanical sensitivity of such waste, allowing for the development of process parameters required to minimize the risk to worker and public when processing this waste. Such parameters will be based on the worst-case kinetic parameters as derived from APTAC measurements as well as the development of controls to mitigate sensitivities that may exist due to friction, impact, and spark. This Test Plan will define the scope and technical approach for activities that implement Quality Assurance requirements relevant to formulation and testing.

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DOE OSTI2016

The Consolidation of Glass-Ceramic Waste forms by Hot Isostatic Pressing: Sample Optimisation - 16581

Zirconolite based glass-ceramics were developed for the disposition of plutonium residues at the UK's Sellafield site. Glass-ceramics act as double barrier systems, separating Pu into a durable ceramic phase and retaining miscellaneous material in the amorphous phase. Here we discuss the optimisation of processing parameters and formulations of these alternative waste-forms. All samples were HIPed at 1250 deg. C, 100 MPa for 4 hrs. By investigating the effects of different pre-HIP heat treatments on sample quality and throughput, and glass phase compositions on the crystalline phase assemblage, an optimised method and wasteform composition were achieved. (authors)

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DOE OSTI2020

Advanced Mixed Waste Treatment Project Overcomes Challenges to Complete Treatment of Debris Waste - 20616

The Advanced Mixed Waste Treatment Project (AMWTP), operated by Fluor Idaho for the Department of Energy Idaho Operations Office, completed the last of its debris waste treatment in the fall of 2019. Concluding that work-scope was certainly not without its challenges. Fluor Idaho and its employees had to adapt to treating extremely challenging waste types, such as high-fissile gram materials and potentially pyrophoric and reactive wastes coming down the stretch. It also had to complete the sizing of large waste boxes from Mound Laboratories in inner-contamination enclosures designed to capture the extremely fine, 'flighty' plutonium-238 in its high-efficiency particulate air filters. Very few waste treatment facilities in the DOE Complex have completed their missions. In fact, most large-scale waste treatment facilities are still under construction with many months or years remaining before 'hot operations' begin. AMWTP is unique because it functioned as a 'one-stop shop.' Transuranic and low-level waste characterization, treatment, certification, and shipping all occurred there. AMWTP concluded the most challenging part of its mission and will continue to ship material out of the state of Idaho for the next decade. This paper will provide a background of the AMWTP mission, and the adaptations to its operating permit, equipment, or treatment processes to address challenging waste types. AMWTP will leave behind a legacy of being one of the most successful waste treatment facilities in the history of the DOE Complex. The facility was well-conceived, well-constructed, and well-executed; it was done right from start to finish. During the course of its waste treatment mission, the facility was required to adapt to changing conditions by treating extremely challenging waste types. (authors)

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DOE OSTI2015

Creating and sustaining a safety conscious work environment - 15240

Organizational Culture, Safety Culture, and a Safety Conscious Work Environment (SCWE) are psychological constructs that describe behaviors of organizations, especially with respect to behaviors needed to ensure safe, highly reliable work. These attributes of culture are often singled out in surveys, root cause analyses and management expectations as the basis for being safe or as the cause of an incident. Organizations that wish to improve culture often do not have a framework to measure and act upon tasks or projects that will promote a positive safety culture (and thus a positive SCWE). Combining the work of organizational behavior pioneers with the 'blue collar' approaches to continuous improvement offers framework organizations may use to improve safety culture while devoting efforts to improving conditions that are meaningful and tangible to floor level employees. The framework consists of the use of Safety Culture and SCWE surveys and focus groups to complement each other and offer topical areas that need improvement or study in the organization. A Safety Culture Improvement Panel will then direct specific improvement projects based on these topic areas and relate these projects to causal variables that effect incremental changes in culture. By applying principles of process improvement to the improvement of culture, the framework is formed to have employees work 'day-today' issues while effectively addressing the often vague or abstract notion of improving culture. A case study of the use of these principles is given by following the SCWE Assessment conducted by the Transuranic Waste Processing Center (TWPC) and subsequent improvements derived from data found in the assessment. The attributes of a SCWE and an effective safety culture accomplish much more than just having a safe place to work. These characteristics result in a high performance, continuously improving organization. The experience of commercial nuclear power, the aviation industry, and other high performing endeavors bear this out. Achieving an effective SCWE is not easy, and sustaining it is even more difficult, but the results are surely worth the effort. The techniques discussed in this presentation are effective, available, and cost-effective; they are easily adaptable for use in other organizations to create, sustain and improve a SCWE. (authors)

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DOE OSTI2016

Initial Testing of Off-Gas Recycle Condensate Liquid from the WTP Low Activity Waste Vitrification Process - 16432

The Waste Treatment and Immobilization Plant (WTP) under construction at the Hanford site is designed to treat and immobilize the inventory of High-Level Waste stored in the 177 underground storage tanks. The WTP process flow was designed to pre-treat this waste, separating it into High Level Waste (HLW) and Low Activity Waste (LAW) fractions for vitrification. Vitrification of the waste generates a condensate stream from the off-gas processes originating from two off-gas treatment unit operations, the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP). 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 processed through the LAW melter again. If the Pretreatment facility is unavailable, this baseline disposition pathway is not viable. Additionally, chloride, fluoride, and sulfate are partially volatile in the melter, accumulating to high concentrations while recycling and impacting the glass waste loading and facility throughput. The primary radionuclide that volatizes and accumulates is {sup 99}Tc. The long half-life and environmental mobility of {sup 99}Tc complicates disposal options for this secondary aqueous waste stream. This task is investigating {sup 99}Tc removal to examine the potential for diverting this stream to an alternate disposition path and avoiding impacting the glass waste loading. Separation methods for {sup 99}Tc have focused on using inorganic precipitation agents that can remove it by reducing it from soluble pertechnetate ion, with a +7 oxidation state, to insoluble technetium dioxide in the +4 oxidation state. Stannous chloride (SnCl{sub 2}) has been found highly effective at causing the reduction and precipitating {sup 99}Tc for removal, but it also precipitates other components. The primary competitor that precipitates is chromium, which consumes the redox activity of the Sn(II), causing addition of more of the SnCl{sub 2}. Research has examined the optimum quantity of SnCl{sub 2}, mixing time, and the impact of other redox-active species. It is also important to know the stability of the precipitated {sup 99}Tc in process conditions, since it is expected it will eventually be re-oxidized by oxygen and dissolve. This work is expected to determine if a {sup 99}Tc removal step could be used to decontaminate the bulk of this stream, and make it suitable for other disposal paths so that the chloride, fluoride, and sulfate can be diverted away from the glass melter. (authors)

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DOE OSTI2016

Mercury Phase II Study - Mercury Behavior in Salt Processing Flowsheet

Mercury (Hg) in the Savannah River Site Liquid Waste System (LWS) originated from decades of canyon processing where it was used as a catalyst for dissolving the aluminum cladding of reactor fuel. Approximately 60 metric tons of mercury is currently present throughout the LWS. Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. In February 2015, a Mercury Program Team was established at the request of the Department of Energy to develop a comprehensive action plan for long-term management and removal of mercury. Evaluation was focused in two Phases. Phase I activities assessed the Liquid Waste inventory and chemical processing behavior using a system-by-system review methodology, and determined the speciation of the different mercury forms (Hg+, Hg++, elemental Hg, organomercury, and soluble versus insoluble mercury) within the LWS. Phase II activities are building on the Phase I activities, and results of the LWS flowsheet evaluations will be summarized in three reports: Mercury Behavior in the Salt Processing Flowsheet (i.e. this report); Mercury Behavior in the Defense Waste Processing Facility (DWPF) Flowsheet; and Mercury behavior in the Tank Farm Flowsheet (Evaporator Operations). The evaluation of the mercury behavior in the salt processing flowsheet indicates, inter alia, the following: (1) In the assembled Salt Batches 7, 8 and 9 in Tank 21, the total mercury is mostly soluble with methylmercury (MHg) contributing over 50% of the total mercury. Based on the analyses of samples from 2H Evaporator feed and drop tanks (Tanks 38/43), the source of MHg in Salt Batches 7, 8 and 9 can be attributed to the 2H evaporator concentrate used in assembling the salt batches. The 2H Evaporator is used to evaporate DWPF recycle water. (2) Comparison of data between Tank 21/49, Salt Solution Feed Tank (SSFT), Decontaminated Salt Solution Hold Tank (DSSHT), and Tank 50 samples suggests that the total mercury as well as speciated forms in the assembled salt batches in Tanks 21/49 pass through the Actinide Removal Process (ARP) / Modular Caustic Side Solvent Extraction Unit (MCU) process to Tank 50 with no significant change in the mercury chemistry. (3) In Tank 50, Decontaminated Salt Solution (DSS) from ARP/MCU is the major contributor to the total mercury including MHg. (4) Speciation analyses of TCLP leached solutions of the grout samples prepared from Tank 21, as well as Tank 50 samples, show the majority of the mercury released in the solution is MHg.

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DOE OSTI2015

Concept of DT fuel cycle for a fusion neutron source

A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 10{sup 18} neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Sustainable Security through International Source Repatriation - 19659

Ensuring that there is a viable pathway for characterizing, transporting, and disposing of disused radioactive sealed sources (sources) is essential to any end-of-life management strategy. The Off-Site Source Recovery Program, sponsored by the Department of Energy National Nuclear Security Administration's Office of Radiological Security, provides permanent and sustained threat reduction through the secure removal and repatriation of disused sealed radioactive sources of United States origin at international locations. With over two decades of operations, the Off-Site Source Recovery Program has developed source removal processes and techniques to ensure the compliant, effective and timely disposition of the material. For device disassembly support, the program subcontracts vendors qualified and experienced in source extraction from devices. United States origin source repatriations provide for permanent threat reduction and reduce their overall global footprint. (author)

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DOE OSTI2015

Effects Influencing Plutonium-Absorber Interactions and Distributions in Routine and Upset Waste Treatment Plant Operations

This report is the third in a series of analyses written in support of a plan to revise the Hanford Waste Treatment and Immobilization Plant (WTP) Preliminary Criticality Safety Evaluation Report (CSER) that is being implemented at the request of the U.S. Department of Energy (DOE) Criticality Safety Group. A report on the chemical disposition of plutonium in Hanford tank wastes was prepared as Phase 1 of this plan (Delegard and Jones 2015). Phase 2 is the provision of a chemistry report to describe the potential impacts on criticality safety of waste processing operations within the WTP (Freer 2014). In accordance with the request from the Environmental and Nuclear Safety Department of the WTP (Miles and Losey 2012), the Phase 2 report assessed the potential for WTP process conditions within and outside the range of normal control parameters to change the ratio of fissile material to neutron-absorbing material in the waste as it is processed with an eye towards potential implications for criticality safety. The Phase 2 study also considered the implications should WTP processes take place within the credible range of chemistry upset conditions. In the present Phase 3 report, the 28 phenomena described in the Phase 2 report were considered with respect to the disposition of plutonium and various absorber elements. The phenomena identified in the Phase 2 report are evaluated in light of the Phase 1 report and other resources to determine the impacts these phenomena might have to alter the plutonium/absorber dispositions and ratios. The outcomes of the Phase 3 evaluations then can be used to inform subsequent engineering decisions and provide reasonable paths forward to mitigate or overcome real or potential criticality concern in plant operations.

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DOE OSTI2016

The New International Project on Irradiated Graphite Processing Approaches (GRAPA) - 16559

Developing of strategies and methods for the management of spent nuclear (irradiated) graphite is one of the major task in decommissioning internationally. More than 250.000 Mg of i-graphite have been accumulated worldwide. Having completed in 2014 a comprehensive collaborative research programme on the treatment options for the disposal of i-graphite wastes to meet disposal criteria, the IAEA decided to maintain an active group of international specialists in this area as part of the International Decommissioning Network (IDN). To foster further development in R and D and steps towards industrial applications in i-graphite management this group was requested to propose the terms of reference and work programme for a new IAEA project. The basic concept for this future programme has been formulated by the German Forschungszentrum (Research Centre) Juelich, Institute IEK-6. The initiative has expanded to include representation from UK, USA, France, Italy, Russia, Ukraine and Lithuania. Additional Member States of the IAEA are expected to join the project. The concept presented by the IEK-6 was discussed among the experts and extended on an international meeting at IAEA in Vienna in February 2015. In addition to the development of the terms of reference a number of individual tasks have been identified. The project will combine a comprehensive state-of-the-art study with concerted actions as well as cross-cutting sub-projects. This will include a practical demonstration of the removal, treatment and disposal of highly-irradiated graphite from a decommissioned reactor. Prior to official implementation of this project preparatory work has been started on topic areas where weaknesses in the scientific knowledge or the technology have been identified. The scope of the international project ('Irradiated Graphite Processing Approaches, GRAPA') and its expected development as well as the possible impact of GRAPA on the future international i-graphite research activities will be described in this presentation. The project will be performed jointly by both the IAEA International Decommissioning Network and the new IAEA Pretreatment Network (IMM, under development). (authors)

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