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Management of the Cs/Sr Capsule Project at the Hanford Site. Technology Readiness Assessment Report

The Federal Project Director (FPD) for the U.S. Department of Energy (DOE), Richland Operations Office (RL) Waste Management and D&D Division (WMD) requested a Technology Readiness Assessment (TRA) for the Management of the Cesium/Strontium Capsule Storage Project (MCSCP) at the Waste Encapsulation and Storage Facility (WESF) on the Hanford Site in Washington State. The MCSCP CD-1 TRA was performed by a team selected in collaboration between the Office of Environmental Management (EM) Chief Engineer (EM-3.3) and RL, WMD FPD. The TRA Team included subject matter and technical experts having experience in cask storage, process engineering, and system design who were independent of the MCSCP, and the team was led by the Director of Operations and Processes from the EM Chief Engineer's Office (EM-3.32). Movement of the Cs/Sr capsules to dry storage, based on information from the conceptual design, involves (1) capsule packaging, (2) capsule transfer, and (3) capsule storage. The project has developed a conceptual process, described in 30059-R-02, "NAC Conceptual Design Report for the Management of the Cesium and Strontium Capsules Project", which identifies the five major activities in the process to complete the transfer from storage pool to pad-mounted cask storage. The process, shown schematically in Figure 1, is comprised of the following process steps: (1) loading capsules into the UCS; (2) UCS processing; (3) UCS insertion into the TSC Basket; (4) cask transport from WESF to CSA and (5) extended storage at the CSA.

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Pathways for Disposal of Commercially-Generated Tritiated Waste

From a waste disposal standpoint, tritium is a major challenge. Because it behaves like hydrogen, tritium exchanges readily with hydrogen in the ground water and moves easily through the ground. Land disposal sites must control the tritium activity and mobility of incoming wastes to protect human health and the environment. Consequently, disposal of tritiated low-level wastes is highly regulated and disposal options are limited. The United States has had eight operating commercial facilities licensed for low-level radioactive waste disposal, only four of which are currently receiving waste. Each of these is licensed and regulated by its state. Only two of these sites accept waste from states outside of their specified regional compact. For waste streams that cannot be disposed directly at one of the four active commercial low-level waste disposal facilities, processing facilities offer various forms of tritiated low-level waste processing and treatment, and then transport and dispose of the residuals at a disposal facility. These processing facilities may remove and recycle tritium, reduce waste volume, solidify liquid waste, remove hazardous constituents, or perform a number of additional treatments. Waste brokers also offer many low-level and mixed waste management and transportation services. These services can be especially helpful for small-quantity tritiated-waste generators, such as universities, research institutions, medical facilities, and some industries. The information contained in this report covers general capabilities and requirements for the various disposal/processing facilities and brokerage companies, but is not considered exhaustive. Typically, each facility has extensive waste acceptance criteria and will require a generator to thoroughly characterize their wastes. Then a contractual agreement between the waste generator and the disposal/processing/broker entity must be in place before waste is accepted. Costs for tritiated waste transportation, processing and disposal vary based a number of factors. In many cases, wastes with very low radioactivity are priced primarily based on weight or volume. For higher activities, costs are based on both volume and activity, with the activity-based charges usually being much larger than volume-based charges. Other factors affecting cost include location, waste classification and form, other hazards in the waste, etc. Costs may be based on general guidelines used by an individual disposal or processing site, but final costs are established by specific contract with each generator. For this report, seven hypothetical waste streams intended to represent commercially-generated tritiated waste were defined in order to calculate comparative costs. Ballpark costs for disposition of these hypothetical waste streams were calculated. These costs ranged from thousands to millions of dollars. Due to the complexity of the cost-determining factors mentioned above, the costs calculated in this report should be understood to represent very rough cost estimates for the various hypothetical wastes. Actual costs could be higher or could be lower due to quantity discounts or other factors.

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Assessment of Geochemical Environment for the Proposed INL Remote-Handled Low-Level Waste Disposal Facility

Conservative sorption parameters have been estimated for the proposed Idaho National Laboratory Remote-Handled Low-Level Waste Disposal Facility. This analysis considers the influence of soils, concrete, and steel components on water chemistry and the influence of water chemistry on the relative partitioning of radionuclides over the life of the facility. A set of estimated conservative distribution coefficients for the primary media encountered by transported radionuclides has been recommended. These media include the vault system, concrete-sand-gravel mix, alluvium, and sedimentary interbeds. This analysis was prepared to support the performance assessment required by U.S. Department of Energy Order 435.1, 'Radioactive Waste Management.' The estimated distribution coefficients are provided to support release and transport calculations of radionuclides from the waste form through the vadose zone. A range of sorption parameters are provided for each key transport media, with recommended values being conservative. The range of uncertainty has been bounded through an assessment of most-likely-minimum and most-likely-maximum distribution coefficient values. The range allows for adequate assessment of mean facility performance while providing the basis for uncertainty analysis.

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Compliance of NNSS Activities with P322-4 Issues Management and NNSSWAC Requirements

The Institutional Quality and Performance Assurance Division's Quality Support Services Group conducted a surveillance of the compliance of Waste Management Program activities to Issues Management/Corrective Action requirements communicated in P322-4 and the Nevada National Security Site Waste Acceptance criteria (NNSSWAC). All corrective actions reviewed and records examined were found to be compliant with appropriate and relevant procedures. There were no Findings or Opportunities for Improvement noted.

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Regulation of the First British Power Reactor to Enter into an Extended Period of 'Care and Maintenance' - 18028

All eleven of the British first-generation Magnox nuclear power plants have now ceased operating and are in various stages of decommissioning. The Magnox reactors were graphite moderated, using carbon dioxide as the primary coolant. The reactor cores are contained in steel reactor pressure vessels, except for the last two stations which have pre-stressed concrete pressure vessels. All but two of the Magnox reactor stations have been completely de-fueled. The spent fuel removed has been despatched for reprocessing at Sellafield in the north-west of England. On behalf of the UK government the Nuclear Decommissioning Authority (NDA) is responsible for cleaning up the majority of the civil nuclear legacy. The current NDA decommissioning strategy is to prepare the Magnox reactor sites for an extended period of Care and Maintenance (C and M), during which quiescent safe storage will take place (also known as safe enclosure), whereby the cores and associated structures will accrue the benefits of radioactive decay over a period of decades. This will reduce the radiation doses to workers during final dismantling and site clearance. The two Magnox rector units at the Bradwell licensed nuclear site in the south-east of England are the first scheduled to enter this quiescent safe-storage period. This is planned to commence in the second-half of 2018, and is scheduled to last for around 70 years. The Office for Nuclear Regulation (ONR) is the independent regulatory body, responsible for enforcing safety and security requirements at nuclear installations in Great Britain (GB). This paper outlines the national strategic context and how ONR's regulatory regime is applied to nuclear power plant decommissioning. It addresses some of the specific challenges posed at Bradwell during the preparatory works for entry to C and M and how ONR has addressed these. It describes how this will culminate in a regulatory decision on the acceptability of the safety and security of the Bradwell site to enter this new stage of its life cycle. Over the last few years ONR has assessed and inspected various safety aspects of decommissioning, radioactive waste management and other preparations for Bradwell's entry into C and M. Some of the operator's proposals have provided challenges to ONR, particularly where there have been changes to established approaches taken at other sites, as the licensee seeks to minimise its costs. These have included how far to remove redundant structures, systems and components prior to C and M; how to define interim and end-states for fuel cooling ponds and other contaminated areas; and how to treat and store Intermediate Level Waste (ILW), for which there is no geological disposal facility yet available. ONR has begun its final assessment of safety and security justifications supplied by the site licensee, Magnox Ltd (ML). ONR is using its legal powers conferred to it under the site licence conditions to authorise this first-of-a-kind transition into a 70 year period of safe storage. (authors)

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Expanded Use of ISOCS Characterization Techniques for WIPP Certification at the Idaho Clean-up Project - 19623

The Advanced Mixed Waste Treatment Project (AMWTP) has been retrieving, treating, characterizing, and shipping legacy Transuranic (TRU) waste for disposal since 2003 as part of the overall Idaho Cleanup Project (ICP) in support of the US Department of Energy. As the ICP has progressed, over 700 containers have been identified that cannot be successfully assayed using the current suite of non-destructive assay (NDA) instruments. To meet this challenge, Fluor Idaho, LLC, and Mirion Technologies (formerly Canberra Industries) have teamed to deploy the In-Situ Object Counting System (ISOCS) to supplement the current NDA capabilities. ISOCS has long been used in various radioactive waste management applications, but never has the ISOCS ability to perform intrinsic efficiency modeling of discrete waste containers been employed to characterize and certify TRU waste for disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. Since the primary disposal facility for this waste is the WIPP, the characterization processes for determining the quantity of radioactive material entrained in the waste are subject to WIPP Waste Acceptance Criteria (WAC) quality requirements and associated performance criteria. Until now, the Carlsbad Field Office (CBFO) WIPP-certified NDA equipment has utilized pre-defined efficiency calibrations for fixed geometries. However, due to the radioactive material content, activity levels, waste stream matrix, and/or the geometries of various size drums, boxes and other-odd-shaped containers or container integrity issues, it may not be suitable to characterize these containers with any of the existing WIPP-certified NDA equipment located at the AMWTP. Therefore, advanced technologies are needed to complete characterization of these problematic waste containers. The July 5, 2016 release of the WIPP WAC, DOE/WIPP-02-3122, Revision 8.0, Appendix A, 'Radioassay Requirements for Transuranic Waste,' was expanded to include several newer concepts such as American Society for Testing and Materials (ASTM) approved methods for efficiency-based calibrations and the use of mathematical computer modeling techniques for calibration. This change prompted AMWTP to pursue obtaining WIPP certification authority for expanded ISOCS techniques to deal with the remaining challenging waste containers and the development and implementation of Mirion's Figures of Merit (FOM) evaluations. These vital FOM evaluations enhance the ISOCS's modeling capabilities by optimizing the dynamic efficiency calibration for each object and reducing the assay's Total Measurement Uncertainty (TMU) budget. (authors)

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Decommissioning Demonstration - San Onofre Nuclear Generating Station South Yard Facility - 16544

Following the decision to retire Units 2 and 3 at the San Onofre Nuclear Generating Station (SONGS), Southern California Edison (SCE) staffed a multi-discipline team to investigate decommissioning best practices. SCE also tasked this team with reducing the overall radiological footprint and radioactive material inventory in the near term. The SCE decommissioning activities investigation team targeted projects at the facility which could reduce the source term and risk, and investigate management approaches with decommissioning contractors. SCE, through this team, contracted EnergySolutions to provide a solution for processing and disposal of equipment and materials at a storage facility south of the plant. The South Yard Facility included offices, instrument repair facilities, a large machine shop, outage equipment storage, waste management facilities, and decontamination facilities. SCE contracted EnergySolutions to manage, supervise, plan, and provide labor for the disposition of the legacy waste stored at the South Yard to include the Radioactive Equipment/Materials Storage (REMS) Pad and South Yard Facility. Under the Contractor Work Authorization (CWA), SCE authorized EnergySolutions to consolidate, package, transport, and dispose of the Low Level Radioactive Waste (Class A) and Mixed Low Level Radioactive Waste from the aforementioned areas. The scope also included disposition of material that was expected to be free from radiological contamination, but required management as a hazardous waste. The legacy waste consisted of an inventory of containers designated as trash provided by SCE, as well as, all materials stored within the rooms of the SYF. (authors)

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Preliminary Concept of Operations for the Spent Fuel Management System+ - 17134

The Nuclear Fuels Storage and Transportation Planning Project (NFST) within the US Department of Energy's Office of Nuclear Energy is tasked with identifying, planning, and conducting activities for developing interim storage and transportation capabilities in support of an integrated waste management system (IWMS). The IWMS is expected to transport Spent Nuclear Fuel (SNF) from a reactor and deliver it to a repository in a suitable container, with the possibility of interim storage. The system will be composed of multiple subsystems including possible interim storage facilities (ISFs), one or more repositories, facilities to package and/or repackage SNF, and transportation. The project team is analyzing options for an IWMS to identify viable options. To support analysis, the project team has developed a concept of operations framework for SNF related components of the potential integrated system and the interdependencies between those components. A goal of this work is to aid system analysts in developing consistent models across the project. The concept of operations framework will be updated periodically as new developments emerge. At a high level, SNF is expected to be transported from reactor sites to a repository. Initially, SNF is unloaded from reactors and placed in spent fuel pools (SFPs) for wet storage at utility sites. To maintain space in SFPs, SNF is placed in containers at reactor sites for storage and/or transportation only after the SNF has cooled enough to satisfy loading limits. Once transportation requirements are met, the SNF may be transported to an ISF, where it is stored until a repository is developed, or it is transported directly to a repository if one is available.While the high level operation of the system is straightforward, analysts must evaluate numerous alternatives, such as the number of ISFs (if any), ISF design, the stage at which SNF repackaging occurs (if needed), repackaging technology, types of containers used, repository design, component sizing, and timing of events. These alternatives arise due to technological, economic, or policy considerations. This paper provides an overview of the potential alternatives identified in the concept of operations framework at a conceptual level. (authors)

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DASAO: software tool for the management of safeguards, waste and decommissioning

Decommissioning of nuclear facilities is a complex process involving operations such as detailed surveys, decontamination and dismantling of equipment's, demolition of buildings and management of resulting waste and nuclear materials if any. This process takes place in a well-developed legal framework and is controlled and followed-up by stakeholders like the Safety Authority, the Radwaste management Agency and the Safeguards Organism. In the framework of its nuclear waste and decommissioning program and more specifically the decommissioning of the BR3 reactor, SCK-CEN has developed different software tools to secure the waste and material traceability, to support the sound management of the decommissioning project and to facilitate the control and the follow-up by the stakeholders. In the case of Belgium, it concerns the Federal Agency for Nuclear Control, the National Agency for radioactive waste management and fissile material and EURATOM and IAEA. In 2005, Belgonucleaire decided to shutdown her Dessel MOX fuel fabrication plant and the production stopped in 2006. According to the final decommissioning plan ('PDF') approved by NIRAS, the decommissioning works should start in 2008 at the earliest. In 2006, the management of Belgonucleaire identified the need for an integrated database and decided to entrust SCK-CEN with its development, because SCK-CEN relies on previous experience in comparable applications namely already approved by authorities such as NIRAS, FANC and EURATOM. The main objectives of this integrated software tool are: - simplified and updated safeguards; - waste and material traceability; - computerized documentation; - support to project management; - periodic and final reporting to waste and safety authorities. The software called DASAO (Database for Safeguards, Waste and Decommissioning) was successfully commissioned in 2008 and extensively used from 2009 to the satisfaction of Belgonucleaire and the stakeholders. SCK-CEN is now implementing a simplified release of the software for the management of the decommissioning of the Thetis reactor. Its decommissioning will start in March 2013 and will be completed by the end of 2014. (authors)

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Saltstone First Quarter Calendar Year 2018 (1QCY18) Toxicity Characteristic Leaching Procedure (TCLP) Results

The aqueous waste from Tank 50 (salt solution) is sampled quarterly for transfers to the Saltstone Production Facility (SPF). Salt solution is treated at SPF and disposed of in the Saltstone Disposal Facility (SDF). A SDF waste form (saltstone) was prepared in the Savannah River National Laboratory (SRNL) from the Tank 50 Waste Acceptance Criteria (WAC) sample1 and Z-Area premix material for the first quarter of calendar year 2018 (1QCY18). 2,3 Results from this memorandum support Task 2: ‘Grout Leaching Analyses’ of the Task Technical Request (TTR). 2 After a 28 day cure, a sample of the SDF waste form was collected and shipped to a certified laboratory for analysis using the Toxicity Characteristic Leaching Procedure (TCLP). 4 The 1QCY18 saltstone sample met the South Carolina (SC) Code of Regulations for Hazardous Waste Management Regulations (HWMR) 6 1-79.261.24 and 61-79-268.48 requirements for a non-hazardous waste form with respect to Resource Conservation and Recovery Act (RCRA) metals and Underlying Hazardous Constituents (UHCs), and also met the Saltstone WAC.

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Probabilistic Performance Assessment for Deep Borehole Disposal of Cs/Sr Capsules - 17143

The US Department of Energy (DOE) performed an assessment of disposal options for spent nuclear fuel (SNF) and high-level radioactive waste (HLW) that recommended the consideration of deep borehole disposal (DBD) of smaller DOE-managed waste forms, such as cesium (Cs) and strontium (Sr) capsules. To further assess the safety and viability of the DBD concept, post-closure performance assessment (PA) analyses were performed for DBD of Cs/Sr capsules. The post-closure PA included a reference design concept for Cs/Sr capsule disposal, feature, event, and process (FEP) analysis, scenario development, and modeling using the PFLOTRAN code. PA simulations were run for both nominal and disturbed scenarios for long-term radionuclide transport away from the deep disposal borehole. The nominal (expected) post-closure release scenario included FEPs for short-duration (a few hundred years) thermally-induced upward fluid flux through the borehole seals and/or disturbed rock zone (DRZ) followed by longer-term slow diffusive transport. Simulations included a baseline deterministic run and a set of probabilistic realizations (using Latin Hypercube Sampling (LHS) from parameter distributions) to examine the sensitivity and importance of the long-term radionuclide transport to processes and parameters such as: waste package durability; seal porosity and permeability, disturbed rock zone (DRZ) porosity and permeability; and radionuclide sorption in the seal materials, DRZ, and crystalline host rock. The disturbed scenario included a waste package 'stuck' in the crystalline basement above the emplacement zone near a hypothetical borehole-intersecting fracture. Simulations included two deterministic runs to examine the sensitivity to the regional head gradient. The nominal and disturbed scenario PA results suggest that a favorable safety case can be developed for DBD of Cs/Sr capsules and helped to identify key areas of uncertainty, which in turn can inform future DBD research and development (R and D) activities, including a field demonstration project. (authors)

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An analysis of the technical status of high level radioactive waste and spent fuel management systems

The technical status of the old U.S. mailine program for high level radioactive nuclear waste management, and the newly-developing program for disposal of unreprocessed spent fuel was assessed. The method of long term containment for both of these waste forms is considered to be deep geologic isolation in bedded salt. Each major component of both waste management systems is analyzed in terms of its scientific feasibility, technical achievability and engineering achievability. The resulting matrix leads to a systematic identification of major unresolved technical or scientific questions and/or gaps in these programs.

English, T.↗

US/German Collaboration in Salt Repository Research, Design and Operation - 13243

Recent developments in the US and Germany [1-3] have precipitated renewed efforts in salt repository investigations and related studies. Both the German rock salt repository activities and the US waste management programs currently face challenges that may adversely affect their respective current and future state-of-the-art core capabilities in rock salt repository science and technology. The research agenda being pursued by our respective countries leverages collective efforts for the benefit of both programs. The topics addressed by the US/German salt repository collaborations align well with the findings and recommendations summarized in the January 2012 US Blue Ribbon Commission on America's Nuclear Future (BRC) report [4] and are consistent with the aspirations of the key topics of the Strategic Research Agenda of the Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) [5]. Against this background, a revival of joint efforts in salt repository investigations after some years of hibernation has been undertaken to leverage collective efforts in salt repository research, design, operations, and related issues for the benefit of respective programs and to form a basis for providing an attractive, cost-effective insurance against the premature loss of virtually irreplaceable scientific expertise and institutional memory. (authors)

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Thermal and mechanical stabilization process of the organic fraction of the municipal solid waste

Graphical abstract: Display Omitted - Highlights: • A domestic scale prototype for the pre-treatment of OFMSW has been tested. • Two grinding techniques are compared and thermopress is used for the drying stage. • Increasing temperature up to 170 °C reduces energy consumption of the drying stage. • In the range 5–10 bar a reduction of 97% of the initial volume is obtained. • In most cases energy recovery from the dried waste matches energy consumption. - Abstract: In the present study a thermo-mechanical treatment for the disposal of the Organic Fraction of Municipal Solid Waste (OFMSW) at apartment or condominium scale is proposed. The process presents several advantages allowing to perform a significant volume and moisture reduction of the produced waste at domestic scale thus producing a material with an increased storability and improved characteristics (e.g. calorific value) that make it available for further alternative uses. The assessment of the applicability of the proposed waste pretreatment in a new scheme of waste management system requires several research steps involving different competences and application scales. In this context, a preliminary study is needed targeting to the evaluation and minimization of the energy consumption associated to the process. To this aim, in the present paper, two configurations of a domestic appliance prototype have been presented and the effect of some operating variables has been investigated in order to select the proper configuration and the best set of operating conditions capable to minimize the duration and the energy consumption of the process. The performances of the prototype have been also tested on three model mixtures representing a possible daily domestic waste and compared with an existing commercially available appliance. The results obtained show that a daily application of the process is feasible given the short treatment time required and the energy consumption comparable to the one of the common domestic appliances. Finally, the evaluation of the energy recovered in the final product per unit weight of raw material shows that in most cases it is comparable to the energy required from the treatment.

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Preliminary Evaluation of Removing Spent Nuclear Fuel from Shutdown Nuclear Power Plant Sites - 2018 - 19101

This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in multi-assembly canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. The U.S. Department of Energy Office of Integrated Waste Management (DOE-IWM) continues to conduct evaluations of removing spent nuclear fuel (SNF) from shutdown nuclear power plant sites. The 14 shutdown sites included in the evaluation to date have nuclear power reactors that have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The 14 shutdown sites are Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, Vermont Yankee, and Fort Calhoun. The Oyster Creek Nuclear Generating Station ceased operations in September 2018 and will be included in future evaluations. The evaluations of the shutdown sites contain four components: - A characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory; - A description of the on-site infrastructure at the shutdown sites; - An evaluation of the near-site transportation infrastructure and experience at the shutdown sites - An evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. The primary sources for the inventory of SNF were the DOE GC-859 and RW-859 spent nuclear fuel inventory databases. This inventory has been augmented with data obtained from the shutdown sites for SNF discharges that occurred after June 2013. The primary sources for information on the conditions of site and near-site transportation infrastructure and experience included field observations and information collected during visits to the 14 shutdown sites, information provided by managers and staff at the shutdown sites, and Google Earth imagery. Extensive photographs taken during the shutdown site visits have proven to be particularly useful in documenting the current conditions at and near the shutdown sites. Tribal and state representatives, State and Regional Group representatives, and Federal Railroad Administration representatives have participated in 11 of the shutdown site visits. In addition to visiting the shutdown site, site visits often include activities such as meeting with local railroads to discuss the near-site railroad infrastructure and meeting with community engagement panels. Oyster Creek permanently ceased power operations on September 17, 2018 and fuel was removed from the reactor vessel on September 25, 2018. Planning for the Oyster Creek site visit has begun and is conducted in three stages. In the first stage, data and documents that are applicable to Oyster Creek are collected. In the second stage, questions are developed based on the data and documents collected during the first stage. These questions are submitted to the Oyster Creek site approximately one month before the site visit. The third stage involves establishing specific schedules and participants. Participants will likely include tribal and state representatives, the Council of State Governments - Eastern Regional Conference, and the Federal Railroad Administration. State participation will likely include a representative of the State Rail Safety Participation Program, the New Jersey Department of Transportation, and the New Jersey Bureau of Nuclear Engineering. The site visit is tentatively expected to occur in April or May of 2019. (authors)

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Implications of Plutonium isotopic separation on closed fuel cycles and repository design

Advances in laser enrichment may enable relatively low-cost plutonium isotopic separation. This would have large impacts on LWR closed fuel cycles and waste management. If Pu-240 is removed before recycling plutonium as mixed oxide (MOX) fuel, it would dramatically reduce the buildup of higher plutonium isotopes, Americium, and Curium. Pu-240 is a fertile material and thus can be replaced by U-238. Eliminating the higher plutonium isotopes in MOX fuel increases the Doppler feedback, simplifies reactor control, and allows infinite recycle of MOX plutonium in LWRs. Eliminating fertile Pu-240 and Pu-242 reduces the plutonium content in MOX fuel and simplifies fabrication. Reducing production of Pu-241 reduces production of Am-241 - the primary heat generator in spent nuclear fuels after several decades. Reducing heat generating Am-241 would reduce repository cost and waste toxicity. Avoiding Am- 241 avoids its decay product Np-237, a nuclide that partly controls long-term oxidizing repository performance. Most of these benefits also apply to LWR plutonium recycled into fast reactors. There are benefits for plutonium isotopic separation in fast reactor fuel cycles (particularly removal of Pu-242) but the benefits are less. (author)

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Technical specifications for mechanical recycling of agricultural plastic waste

Highlights: • Technical specifications for agricultural plastic wastes (APWs) recycling proposed. • Specifications are the base for best economical and environmental APW valorisation. • Analysis of APW reveals inherent characteristics and constraints of APW streams. • Thorough survey on mechanical recycling processes and industry as it applies to APW. • Specifications for APW recycling tested, adjusted and verified through pilot trials. - Abstract: Technical specifications appropriate for the recycling of agricultural plastic wastes (APWs), widely accepted by the recycling industry were developed. The specifications establish quality standards to be met by the agricultural plastics producers, users and the agricultural plastic waste management chain. They constitute the base for the best economical and environmental valorisation of the APW. The analysis of the APW streams conducted across Europe in the framework of the European project “LabelAgriWaste” revealed the inherent characteristics of the APW streams and the inherent constraints (technical or economical) of the APW. The APW stream properties related to its recycling potential and measured during pilot trials are presented and a subsequent universally accepted simplified and expanded list of APW recycling technical specifications is proposed and justified. The list includes two sets of specifications, applied to two different quality categories of recyclable APW: one for pellet production process (“Quality I”) and another one for plastic profile production process (“Quality II”). Parameters that are taken into consideration in the specifications include the APW physical characteristics, contamination, composition and degradation. The proposed specifications are focused on polyethylene based APW that represents the vast majority of the APW stream. However, the specifications can be adjusted to cover also APW of different materials (e.g. PP or PVC) that are found in very small quantities in protected cultivations in Europe. The adoption of the proposed specifications could transform this waste stream into a labelled commodity traded freely in the market and will constitute the base for the best economical and environmental valorisation of the APW.

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Panel Session 34: US DOE Featured Site: Savannah River Operations Office: 70 Years of Service

The Savannah River Site celebrates its 70. anniversary on November 28, 2020. On this date in 1950, President Harry S. Truman requested the Dupont Company to design, build and operate what was then known as the Savannah River Plant in response to the Soviet Union's detonation of its first atomic weapon, which set the Cold War into motion. During the 1950's, six South Carolina towns were relocated for the construction of SRS and by 1953, the 310 square mile site was complete. Nearly 40,000 workers were employed t build five nuclear reactors and support facilities, two chemical separations plants, heavy water extraction plant, nuclear fuel and target fabrication facility tritium extraction facility and waste management facilities. SRS played a key role in winning the Cold War and for seven decades, SRS has been a leader within the DOE complex. Today, the site supports environmental stewardship and maintains the nation's nuclear deterrent while ensuring the safekeeping and disposal of domestic and international nuclear materials. The site continues to support the nation's nuclear defense as it explores new potential NNSA missions. SRS has a proud 70- year history and looks forward to a future of service as a national asset and strong community partner. The session was kicked off with a video message from Secretary of Energy, Dan Brouillette, who thanked employees past and current for their efforts. The video also provided an overview of the history and future of the Site. This panel provided an overview of the Savannah River Site's 70 years of service (history, challenges, opportunities, and future) presented by the SRS's Senior leadership and the local Aiken, South Carolina Mayor. Panelists with presentations: US DOE Secretary Dan Brouillette's Overview of DOE and SRS - A Legacy of 70 Years of Service (Amy Boyette); 70 Years of Service (Michael Budney); 70 Years of Service (Stuart MacVean); SRS Liquid Waste (Thomas Foster); National Nuclear Security Administration (Nicole Nelson-Jean)

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