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At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Long-term stewardship - surveillance and maintenance at Hanford - 15155

In FY 2015, Hanford's Long-Term Stewardship (LTS) Program is tasked to complete the required entries into five cocooned plutonium production reactors (105-C, 105-D, 105-F, 105-H and 105-N), which sit adjacent to the Columbia River. The first reactor entered in October 2014, the cocooned plutonium production reactor, 105-F, signaled the LTS Program's entry into the Surveillance and Maintenance (S and M) era. Reactor entries at Hanford are conducted at 5-year intervals to assess the condition of the structures and evaluate potential deterioration of the reactor core, shield walls and roof. The cocooned structures are designed to protect the reactor core for up to 75 years while radioactive decay continues, ultimately making the structures safe for demolition and removal. The 105-F entry demonstrates that the LTS program at Hanford is shifting from a program focused on transitioning cleaned up land and waste sites to a program focused primarily on S and M activities. The entry process is carefully planned, choreographed and monitored to ensure safety of the workers and minimize the time the reactor structure is opened. The process includes controlled opening of the reactor, air quality monitoring, structural safety inspections, and radiological clearance. Once the structure is cleared for entry, assessments of the structural integrity and radiologic conditions are completed. Evaluation of the interior temperature and flood monitoring equipment is conducted. Facility visits by regulators and DOE personnel close out the process prior to welding the door closed until the next assessment, five years later. This paper will document execution of the entry work package, identify challenges overcome, and provide video documentation as reference material for future workers and the public. The current LTS program has responsibility for approximately 100,000 acres, over 200 waste sites and six cocooned reactors. Just over 5,600 documents have been identified, indexed and tagged for storage in the LTS document library. The program has successfully completed seven consecutive transitions since July 2011, with six more transitions planned in the next 18 months to support the U.S. Department of Energy Richland Operations Office's (DOE-RL) near-term cleanup objectives to significantly reduce the footprint of active cleanup operations for the River Corridor. (authors)

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DOE OSTI2013

APPLICATIONS OF CURRENT TECHNOLOGY FOR CONTINUOUS MONITORING OF SPENT FUEL

Advancements in technology have opened many opportunities to improve upon the current infrastructure surrounding the nuclear fuel cycle. Embedded devices, very small sensors, and wireless technology can be applied to Security, Safety, and Nonproliferation of Spent Nuclear Fuel. Security, separate of current video monitoring systems, can be improved by integrating current wireless technology with a variety of sensors including motion detection, altimeter, accelerometer, and a tagging system. By continually monitoring these sensors, thresholds can be set to sense deviations from nominal values. Then alarms or notifications can be activated as needed. Safety can be improved in several ways. First, human exposure to ionizing radiation can be reduced by using a wireless sensor package on each spent fuel cask to monitor radiation, temperature, humidity, etc. Since the sensor data is monitored remotely operator stay-time is decreased and distance from the spent fuel increased, so the overall radiation exposure is reduced as compared to visual inspections. The second improvement is the ability to monitor continuously rather than periodically. If changes occur to the material, alarm thresholds could be set and notifications made to provide advanced notice of negative data trends. These sensor packages could also record data to be used for scientific evaluation and studies to improve transportation and storage safety. Nonproliferation can be improved for spent fuel transportation and storage by designing an integrated tag that uses current infrastructure for reporting and in an event; tracking can be accomplished using the Iridium satellite system. This technology is similar to GPS but with higher signal strength and penetration power, but lower accuracy. A sensor package can integrate all or some of the above depending on the transportation and storage requirements and regulations. A sensor package can be developed using off the shelf technology and applying it to each specific need. There are products on the market for smart meters, industrial lighting control and home automation that can be applied to the Back End Fuel Cycle. With a little integration and innovation a cost effective solution is achievable.

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DOE OSTI2025

Mixed polyamide and polyester upcycling via chemical autoxidation and engineered Pseudomonas putida

Polyamides, such as nylons, are often used in multi-component materials, like textiles and packaging, and are accompanied with unique recycling challenges. Recently, autoxidation and bioconversion has emerged as a tandem approach for the conversion of mixed plastics waste to single products, however the fate of polyamides in these processes is unknown. Here, we optimized the autoxidation of nylon-6 and nylon-6,6 depolymerization, achieving >92 mol% nitrogen recovery from both substrates, predominantly as acetamide, and 20–27 mol% carbon recovery (not including acetamide). Experiments with 13 C-labeled acetic acid demonstrated that the carbon in acetamide was solvent derived. Autoxidation of mixed nylon-6 and poly(ethylene terephthalate) (PET) post-consumer fibers resulted in similar carbon and nitrogen recoveries from nylon, while PET was depolymerized to terephthalic acid (TPA) at >65 C-mol% recovery. Next, we engineered Pseudomonas putida KT2440 to utilize acetamide as the sole carbon and nitrogen source for growth through the constitutive expression of genes encoding amidase enzymes, including a native amidase (PP_0613) shown to be active on C 2 –C 4 amides. Heterologous chromosomal expression of amiE, encoding the amidase from P. aeruginosa, was found to be superior to PP_0613 constitutive expression in genome integrated strains. Prior engineering to enable TPA conversion to β-ketoadipate pathway intermediate protocatechuate was leveraged and combined with deletion of pcaD to produce muconolactone as a product. Finally, a stacked strain engineered for conversion of acetamide, TPA, and DCAs was evaluated on the reaction product from autoxidation of mixed post-consumer nylon and PET fibers without any supplemental nitrogen, achieving quantitative yields in the presence of supplemental carbon.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Examination of shipping packages 9975-01641, 9975-01692, 9975-03373, 9975-02101 AND 9975-02713

SRNL has assisted in the examination of five 9975 shipping packages following storage of nuclear material in K-Area Complex (KAC). Two packages (9975-01641 and -01692) with water intrusion resulting from a roof leak were selected for detailed examination after internal fiberboard degradation (mold) was observed. 9975-01692 contained regions of saturated fiberboard and significant mold, while the second package was less degraded. A third package (9975-03373) was removed from storage for routine surveillance activities, and set aside for further examination after a musty odor was noted inside. No additional degradation was noted in 9975-03373, but the lower assembly could not be removed from the drum for detailed examination. Two additional packages (9975-02101 and -02713) identified for further examination were among a larger group selected for surveillance as part of a specific focus on high-wattage packages. These two packages displayed several non-conforming conditions, including the following: (1) the axial gap criterion was exceeded, (2) a significant concentration of moisture was found in the bottom fiberboard layers, with active mold in this area, (3) condensation and/or water stains were observed on internal components (drum, lid, air shield), and (4) both drums contained localized corrosion along the bottom lip. It is recommended that a new screening check be implemented for packages that are removed from storage, as well as high wattage packages remaining in storage. An initial survey for corrosion along the drum bottom lip of high wattage packages could identify potential degraded packages for future surveillance focus. In addition, after packages have been removed from storage (and unloaded), the drum bottom lip and underside should be inspected for corrosion. The presence of corrosion could signal the need to remove the lower fiberboard assembly for further inspection of the fiberboard and drum prior to recertification of the package.

42 ENGINEERING↗
NASA NTRSDate not supplied

Design Considerations for Aerocapture Delivery of Uranus Orbiter and Probe

This paper presents an aerocapture system design for delivering the Uranus Orbiter and Probe (UOP) subsystems to their science orbit using the Mars Science Laboratory architecture. A packaging and integration scheme are presented for UOP subsystems, with careful consideration of large components required for outer planet science missions (RTG power generation, RF communications). The implication of packaging multiple RTG’s inside an aeroshell during a multiple year cruise phase will be explained, with possible thermal design solutions to mitigate RTG waste heat analyzed. Mass properties and margins for the system are discussed and an overall mass risk assessment is presented for the entry system and total flight system at launch. Alternative vehicle configurations are explored to include multiple probes and enable science objectives during transit to Uranus. Aerocapture mass savings when compared to the baseline UOP design are presented.

Andrew J. Gomez-Delrio↗
NASA NTRSDate not supplied

Design Considerations for Aerocapture Delivery of Uranus Orbiter and Probe

This paper presents an aerocapture system design for delivering the Uranus Orbiter and Probe (UOP) subsystems to their science orbit using the Mars Science Laboratory architecture. A packaging and integration scheme are presented for UOP subsystems, with careful consideration of large components required for outer planet science missions (RTG power generation, RF communications). The implication of packaging multiple RTG’s inside an aeroshell during a multiple year cruise phase will be explained, with possible thermal design solutions to mitigate RTG waste heat analyzed. Mass properties and margins for the system are discussed and an overall mass risk assessment is presented for the entry system and total flight system at launch. Alternative vehicle configurations are explored to include multiple probes and enable science objectives during transit to Uranus. Aerocapture mass savings when compared to the baseline UOP design are presented.

Aerocapture↗
DOE OSTI2018

Analyses to support waste disposition of SNS Inner Reflector Plug

The Spallation Neutron Source Inner Reflector Plug (IRP) is scheduled for replacement. It reached end of life-time because of moderator neutron poison and decoupler burn-out, which is used for shaping neutron pulses. The replacement of IRP is a complex task due to its location in the area receiving high irradiation, being under significant amount of shielding, and size, which requires removal in segments. In order to support the disposal of spent IRP, waste characterization analyses are performed. These analyses include an accurate estimate of the radionuclide inventory and the residual dose rate. Using this data, the package class is determined for the and appropriate containers for temporary storage and subsequent transport are assigned. Analyses are performed in four steps: a radiation transport analysis of the radiation environment for producing the flux and isotope production information using MCNPX, a multi-cell transmutation analysis based n CINDER90, the generation of decay gamma source terms, and the gamma transport analyses of the decay gammas arriving at residual dose rates again using MCNPX. . Due to the complexity of the geometry of the lower segment, transport analyses for the residual dose rate require elaborate work for the source terms definition.

Popova, Irina I.↗
DOE OSTI2023

Low Activity Waste Tuning Feed Material Testing VSL-18R4350-1 (Final Report)

High level waste (HLW) from the Hanford tank farms will be retrieved and transferred to the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The waste will be pretreated to separate the soluble chemical salts (mostly sodium nitrate/nitrite and sodium hydroxide) from the bulk of the radioactive materials. This decontaminated salt solution is the low activity waste (LAW) stream which constitutes most of the total waste volume. The remaining HLW contains most of the radioactive materials but a fraction of the volume. Each of these waste streams is directed to an independent processing facility where the waste is mixed with glass forming chemicals (mostly silica and borax or boric acid) and fed into the melters for stabilization by conversion into glass. The molten glass is poured into stainless steel containers to produce packages for disposal: local shallow burial for the LAW containers and a future geologic repository for the HLW containers. The LAW facility melters produce significant quantities of NOx-laden off-gas that require abatement in accordance with 10 CFR 830 and air emission requirements. The NOx emissions also pose a safety risk. The LAW facility commissioning sequence requires that the melters be operated and process control loops tuned prior to introducing waste or waste feed materials that produce NOx. Therefore, a temporary or permanent system architecture is needed that provides a feed supply to the melters for the purpose of melter testing and off-gas tuning that does not result in the production of significantly hazardous off-gas products.

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DOE OSTI2020

NASNF Processing and Packaging Technical Study

The ABD program will require development of a strategy for the dispositioning the Non-Aluminum Spent Nuclear Fuel (NASNF), some of which will be very challenging to handle and dissolve as described in this report. Not all the NASNF is suitable for processing through the electrolytic dissolver. The strategy may involve dissolution of the majority of the NASNF inventory in the electrolytic dissolver combined with some other disposition alternative(s) for the fraction that is not suitable for electrolytic dissolution. This report highlights the risks and challenges associated with the baseline approach of using the electrolytic dissolver to process all of the NASNF. In addition, the characteristics of the NASNF and how those characteristics relate to the risks associated with electrolytically dissolving the specific fuel types.

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DOE OSTI2020

Combined Effort to De-Inventory the New Brunswick Laboratory of Plutonium Certified Reference Materials - 20504

To de-inventory plutonium Certified Reference Materials standards (CRMs) from the New Brunswick National Laboratory (NBL), the Argonne National Laboratory (ANL) and Savannah River National Laboratory (SRNL) formed a unique team. This teaming arrangement leveraged the different capabilities and expertise of both laboratories, helping to ensure the effort would be performed both safely and cost-effectively. With the NBL then physically housed on the ANL site, ANL served as the host-site, providing the necessary facilities and support services (e.g., supplying everything from hoods, loading equipment, and scales to Health Physics supports). From a packaging perspective, ANL's responsibilities focused mainly on characterizing and preparing the contents for shipment, followed by performing the Shipper of Record activities. With SRNL being the Package Design Authority of the 9978, the Type B packaging used for this effort, SRNL's scope included developing a Safety Analysis Report in Packaging (SARP) Letter Amendment to add the CRMs as authorized contents to the 9978. Recognizing competency, ANL prodigiously tasked SRNL to perform the actual loading of the CRMs into the 9978s, while physically on the ANL site. This paper details five different aspects of this effort. They are: 1. Characterization/validation of the CRM contents. 2. Development of the 9978 SARP Letter Amendment for the content change. 3. Development of the nitrogen inerting method to address SARP identified potential flammability concerns. 4. Renewal of the SRNL 9978 loading procedure. 5. Performance of the actual work execution activities. In 2017, the first nine packages were successfully loaded and shipped from NBL, while in 2019, the last ten were successfully loaded and planned for shipment. Without ANL leveraging SRNL's strengths, the de-inventorying of NBL of its plutonium CRMs would have likely taken many years longer. (authors)

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DOE OSTI2015

Assessment of Residual Stresses in 3013 Inner and Outer Containers and Teardrop Samples

This report is an assessment performed by LANL that examines packaging for plutonium-bearing materials and the resilience of its design. This report discusses residual stresses in the 3013 outer, the SRS/Hanford and RFETS/LLNL inner containers, and teardrop samples used in studies to assess the potential for SCC in 3013 containers. Residual tensile stresses in the heat affected zones of the closure welds are of particular concern.

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DOE OSTI2021

Draft Experimental Plan for Commercial SNF Degradation in Repository Environments with a Focus on Fracture Matrix Degradation

This report documents work performed supporting US Department of Energy (DOE) Nuclear Energy Spent Fuel and Waste Disposition, Spent Fuel and Waste Science and Technology, under work breakdown structure element 1.08.01.03.04, “Inventory and Waste Form Characteristics and Performance.” In particular, this report fulfills the M3 milestone, M3SF-21OR010309072 "Draft Experimental Plan for Commercial SNF Degradation in Repository Environments” as described in work package SF-21OR01030907 "SNF Degradation Testing-ORNL".This experimental plan is a collaboration between ORNL, SNL, ANL, and PNNL. It is intended to support source term model development and spent fuel degradation testing in a range of generic repository conditions. Potential testing and experiments to support validation of the Fuel Matrix Dissolution Model (FMDM) and the overall development/validation of repository source term modeling are discussed.

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DOE OSTI2016

Low Activity Waste Pretreatment System (LAWPS) Conceptual Design Approach - 16568

The Low Activity Waste Pretreatment System (LAWPS) will be capable of providing the Waste Treatment and Immobilization Project's (WTP) Low Activity Waste Facility (LAW) the feed material needed to operate both of the LAW melters at full capacity. LAWPS will receive supernatant from the Hanford AP tank farm, filter solids from the waste, remove cesium from the waste, and then stage the feed in storage tanks for supplying the LAW facility. The LAWPS project team is currently executing the preliminary design phase of the project leading to Critical Decision 2 (CD-2). CD-2 is the approval of the performance baseline. The Critical Decision 1 (CD-1) package was approved by the Department of Energy Deputy Under Secretary in May 2015. This paper will focus on creating and gaining approval of the CD-1 package and initiating execution of the preliminary design work. The intent of the paper will be to highlight the activities that facilitated achievement of the goals and present key activities that have been performed to increase the likelihood of preliminary design success. Initiating the immobilization of Hanford tank waste will be a tremendous accomplishment. LAWPS is one of the key projects required to achieve this accomplishment. (authors)

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DOE OSTI2015

Requalification of historical intermediate level waste (ILW) for the repository KONRAD on basis of non destructive assay (NDA) and re-evaluated nuclide vectors - 15289

The storage facility for ILW of the HDB contains a large number of historical waste products in 200-l drum geometry, which were produced in the 1970's to the 1990's. In the next years more ILW will be treated and stored at HDB, exceeding the storage capacity of the ILW store. All waste products produced at HDB are treated according to the acceptance criteria of the repository KONRAD and are planned to final storage in this repository. In order to qualify the waste materials for final storage in KONRAD and to remove parts of the inventory from the storage chambers this project was started. The main issues of the project, which will be accompanied by the regulator, are product control measures like gas chromatographic analyses of remotely sampled gas syringes and non destructive assay techniques like?-spectrometry and neutron-analytics. The product control is followed by declaration on the basis of re-evaluated nuclide vectors and a repackaging of the drums in licensed containers. The packaging concept includes a 400-l drum as qualified over-pack, because several of the historical drums cannot be qualified as IP-II or type A package. The annual space between the two drums is filled with quartz sand for stability and additional shielding. The drum set is afterwards inserted into a qualified UBA-container provided by GNS. In the case that the product control measures and the reassessment of the declarations result in a violation of the activity limits of the relevant acceptance criteria for KONRAD, the respective drum has to be physically re-qualified after correspondence with the surveillance authorities. This program will be applied to a fraction of approx. 400 canisters, which were selected on the basis of several quality aspects like dose rate below 100 mSv/h, nuclide inventory below 100 A2 according to paragraph 201 of TS-R-1 and well-known waste composition. In the future more waste products will be treated in the same way, if approved containers exist for the inventory. (authors)

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DOE OSTI2017

Thermal Analysis of a Pipe Over-pack Container with a Combustible Waste Matrix for Los Alamos National Laboratory - 17307

Los Alamos National Laboratory (LANL) Technical Area 55 (TA 55) utilizes several different container types to store and transfer special nuclear material and waste for numerous programs. The Nuclear Process Infrastructure - Infrastructure Operations Group (NPI-2) analyzes and certifies containers to meet the specifications for distinct processes and storage. Engineering analysis performed on these containers include but are not limited to: leak testing, filter efficiency testing, pressure drop test, drop testing, water penetration testing, polymer (o-ring) hardness testing, and thermal steady state testing. Of all of these tests, thermal steady-state testing is most crucial to ensure critical components for specific containers do not fail under an internal heat payload. This testing provides data for setting heat payload limits for different containers. The Pipe Over-pack Container (POC) is a vented carbon steel container with a removable lid designed for storing or transferring nuclear waste. Additionally, TA 55 has been tasked to manage and store all of its Transuranic (TRU) waste for the foreseeable future. Being able to place more material into a single container conserves the physical storage space available at TA 55. The Pipe Over pack Handling and Operations Manual requires each user to ensure that the maximum temperature of each component in the POC is not exceeded. Conducting a thermal steady-state test with a combustible waste matrix on the POC can benefit stakeholders on increasing the material limits of the container. The results of these measurements will establish new wattage limits to meet LANL's transportation requirements. The current LANL Transportation Safety Document (TSD) limits the amount of heat source plutonium to 10 grams per POC, corresponding to approximately 5 Watts of heat. The result of this test was provided to the Packaging and Transportation (OS-PT) group to evaluate new wattage limits that can satisfy the needs of TA 55, while maintaining a safe packaging configuration. (authors)

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DOE OSTI2024

Advanced Reactors Spent Fuel & Waste Science and Technology Program

Based on the higher interest in Advanced Reactor (AR) deployment (e.g., ARDP ) for potential new fuel cycles, the Spent Fuel & Waste Science and Technology (SFWST) Program has begun to evaluate the possible implications of long-term management and final disposition of potential Advance Reactor spent nuclear fuels (SNF) that would be generated in potential advanced reactors. Safely managing and dispositioning the potential future AR SNF, and any other associated radioactive wastes, is the primary focus of this initial preliminary assessment of those. This paper summarizes three primary tasks the Spent Fuel & Waste Science and Technology (SFWST) Program is executing (or collaborating on) related to the back end of the nuclear fuel cycle (BENFC) for potential future advanced reactors: 1. Advanced Reactors Spent Fuel and Waste Streams: Strategies for the BENFC This set of activities define a high-level strategy for how to systematically approach, identify, and close research and development (R&D) gaps/potential issues associated with long-term management and final disposition of AR SNF and other possible AR waste streams. This task involves summarizing advanced reactor concepts, their likely SNF and other waste forms, and identifying previous experience with similar materials, for example from DOE-managed SNF, with closely related characteristics to the potential future AR SNF. Technical R&D gaps between the breadth of detailed understanding for safe storage, transportation and disposal of the existing light water reactor SNF fuel cycle (e.g., see NASEM, 2022) and potential future fuel cycles based on advanced reactors would then be identified. 2. Characterization and Packaging Options of Advanced Reactor SNF These activities evaluate characteristics and packaging options for advanced reactor spent fuel forms. The fuel forms are categorized into three types: (1) tri-structural isotropic (TRISO), (2) metallic, and (3) fuel salt. Emphasis is given to TRISO and metallic SNF and additional waste streams from such AR as driven by the near-term anticipated operation of the Xe-100 and the Natrium reactors as advanced-reactor demonstrations1. Preliminary information for the spent-fuel salt discharged from molten-salt reactors (MSRs) is examined to provide a baseline for future efforts. All calculations and assumptions in this work are based on publicly available information. The following characteristics are calculated or estimated for use in the preliminary assessments: SNF volume and mass, radiation/activity levels through time, thermal conditions through time, potential radionuclide source terms, chemical interactions and evolutions, disposal inventories, and waste-form lifetime. Using those characteristics, calculations to determine the applicability of existing canister designs were performed. These evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, as-loaded radiation, thermal, and criticality analyses to identify constraints for storage, transportation, and disposal. 3. Back-End Management of Advanced Reactors (BEMAR) The DOE NE-8 organization has defined an Integrated Project Team to evaluate the Back End Management of Advanced Reactors (BEMAR) (includes DOE staff from a range of organizations (e.g., NE-81, NE-82, OCED) and national laboratory technical staff within the DOE NE-81 and NE-82 programs). This BEMAR group works directly with advance reactors developers to assess for the DOE the technical feasibility of storage, transportation, and disposal of AR SNF based on the characteristics provided by the developers to DOE (much of which is proprietary). The BEMAR is also tasked to develop rough-order-of-magnitude cost estimates to compare the waste management system for individual advanced reactors to existing light-water reactor management practices. To accomplish this, the BEMAR group is implementing a Systems Engineering approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of the Mobile Systems for Conditioning of Disused Sealed Radioactive Sources in Serbia - 20105

Sealed radioactive sources (SRS) are being used worldwide in the field of medicine, agriculture, industry and research. They can be found in mobile as well as stationary devices. SRS contains radioactive material that is (a) permanently sealed in a capsule or (b) closely bounded and in a solid form. The capsule or material of an SRS should be strong enough to maintain leak tightness under the conditions of use and purpose for which the source was designed, also in case of accidents. In this case only emitted radiation is utilized. Firstly, the hazard from external radiation has to be considered, but the possibility of contamination due to fracture of the capsule should not be disregarded. The radioactive sources are composed of the radiating isotope contained in the filling medium, the single or double isotope holder that partially or totally surrounds the filling medium, the outer cover that contains the parts mentioned above and the capsule closed airtightly by welding or using some other method. The capsule must be tested for leakage periodically. If the SRS is no longer needed (e.g. replaced by a different technique) or it becomes useless for the intended application (e.g. the activity becomes too weak, the equipment containing the source works poorly or becomes obsolete, the source is damaged or leaking) it is considered disused. Disused sealed radioactive sources (DSRS) are typically conditioned and disposed if a facility is available. If the disposal option is not available, conditioned DSRS should be stored under proper conditions. In some cases, the radionuclide(s) in DSRS can be recovered/recycled or the DSRS can be repurposed for other applications. Conditioning of DSRS ensures containment of the radioactive material, provides confinement for leaking sources, provides sufficient radiation shielding, reduces storage/disposal volume by allowing consolidation of multiple sources into a single storage/disposal container, facilitates transport operations and contributes to safety and security as well. Typically, conditioning technologies are deployed either as permanently installed stationary systems in centralized or mobile on-site waste processing facilities, or in a mobile configuration. Centralized stationary facilities provide a single processing location for multiple users that requires transport of the waste to the facility. On the other hand, mobile systems may be provided for the selection and application of the optimum technology for a specific waste stream (such as DSRS) by bringing the process to the point where the waste is generated. In addition, mobile systems could offer additional flexibility by sharing equipment among multiple waste generating sites for processing campaigns that vary in duration, from very short periods to several years. The term 'mobile processing system' refers to any radioactive waste processing system or component which is designed to be transportable and which is not considered permanently installed. Two mobile system for conditioning of disused sealed radioactive sources are developed in the Public Company Nuclear Facilities of Serbia. Development of these mobile systems was supported by SRB9005 national project via Technical Cooperation of the IAEA. The first mobile system, built inside the 20 feet ISO container, will be used for conditioning of DSRS category 3 to 5. The second mobile system, built inside the 7 m long vehicle (Iveco Daily Van), will be used for dismantling of ionizing smoke detectors mostly with Am-241 sources. Designs of the mobile systems were defined in cooperation with two companies from Belgium (Belgoprocess and Leniko) and a Croatian company Ekoteh as well as with the support of the IAEA experts. The generic safety assessment and operational procedures for the mobile systems are developed. Based on safety assessment the acceptance criteria and operational limits and conditions are established. Operational procedures include: (a) equipment and material requirements, (b) assembling procedure of the mobile unit, (c) procedure for acceptance of devices for dismantling and conditioning, (d) dismantling procedure for devices to recover the DSRS, (e) characterization of DSRS, (f) encapsulation procedure of DSRS, (g) disassembling procedure of the mobile unit, and (h) keeping records, identification and traceability. In addition, radiation safety, health safety, security and emergency preparedness plans are prepared. The generic safety assessment and operational procedures could be updated with site specific requirements, DSRS inventory, and different needs for future customers. Developed mobile systems could be used in all situations when it is feasible to perform conditioning of DSRS on the spot in the county and worldwide. Development of these mobile units was just the first step to create the Reference Center for Radioactive Waste Treatment and Disused Radioactive Sources Conditioning for Small Facilities which can become a regional training center in the future, and/or as a tool for comprehensive national search and secure programmes. In the next phases development of e-learning platforms and blended learning packages as well as application for the IAEA Qualified Technical Centre (QTC) for the management of DSRS is foreseen. (authors)

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