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DOE OSTI2011

A framework for understanding waste management studies in construction

During the past decades, construction and demolition (C and D) waste issues have received increasing attention from both practitioners and researchers around the world. A plethora of research relating to C and D waste management (WM) has been published in scholarly journals. However, a comprehensive understanding of the C and D WM research is somehow absent in spite of its proliferation. The aim of this paper is to develop a framework that helps readers understand the C and D WM research as archived in selected journals. Papers under the topic of C and D WM are retrieved based on a set of rigorous procedures. The information of these papers is then analyzed with the assistance of the Qualitative Social Research (QSR) software package NVivo. A framework for understanding C and D WM research is created based on the analytic results. By following the framework, a bibliometric analysis of research in C and D WM is presented, followed by an in-depth literature analysis. It is found that C and D generation, reduction, and recycling are the three major topics in the discipline of C and D WM. Future research is recommended to (a) investigate C and D waste issues in wider scopes including design, maintenance and demolition, (b) develop a unified measurement for waste generation so that WM performance can be compared across various economies, and (c) enhance effectiveness of WM approaches (e.g. waste charging scheme) based on new WM concepts (e.g. Extended Producer Responsibility). In addition to the above research findings, the approach for producing the research framework can be useful references for other studies which attempt to understand the research of a given discipline.

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DOE OSTI2013

Successful Deployment of System for the Storage and Retrieval of Spent/Used Nuclear Fuel from Hanford K-West Fuel Storage Basin-13051

In 2012, a system was deployed to remove, transport, and interim store chemically reactive and highly radioactive sludge material from the Hanford Site's 105-K West Fuel Storage Basin that will be managed as spent/used nuclear fuel. The Knockout Pot (KOP) sludge in the 105-K West Basin was a legacy issue resulting from the spent nuclear fuel (SNF) washing process applied to 2200 metric tons of highly degraded fuel elements following long-term underwater storage. The washing process removed uranium metal and other non-uranium constituents that could pass through a screen with 0.25-inch openings; larger pieces are, by definition, SNF or fuel scrap. When originally retrieved, KOP sludge contained pieces of degraded uranium fuel ranging from 600 microns (μm) to 6350 μm mixed with inert material such as aluminum hydroxide, aluminum wire, and graphite in the same size range. In 2011, a system was developed, tested, successfully deployed and operated to pre-treat KOP sludge as part of 105-K West Basin cleanup. The pretreatment process successfully removed the vast majority of inert material from the KOP sludge stream and reduced the remaining volume of material by approximately 65 percent, down to approximately 50 liters of material requiring management as used fuel. The removal of inert material resulted in significant waste minimization and project cost savings because of the reduced number of transportation/storage containers and improvement in worker safety. The improvement in worker safety is a result of shorter operating times and reduced number of remote handled shipments to the site fuel storage facility. Additionally in 2011, technology development, final design, and cold testing was completed on the system to be used in processing and packaging the remaining KOP material for removal from the basin in much the same manner spent fuel was removed. This system was deployed and successfully operated from June through September 2012, to remove and package the last of the SNF fragments from the 105-K West Basin, ending the long and complex history of the KOP sludge materials. The planning and execution of this project demonstrated how the graded application of DOE O 413.3B, Program and Project Management for the Acquisition of Capital Assets[1], DOE-STD-1189-2008, Integration of Safety into the Design Process[2], and DOE G 413.3-4, U.S. Department of Energy Technology Readiness Assessment Guide[3], can positively affect the outcome of project implementation in the DOE Complex. Provided herein are relevant information, ideas, and tools for use by other projects facing similar issues in managing high risk waste streams in a complex regulatory environment. Positive aspects are also provided of appropriate time spent in detailed planning, design, and testing in non-hazardous environments to reduce project risks in both cost and safety performance, as well as improving confidence in meeting project goals through predictable and reliable performance. (authors)

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DOE OSTI2013

Np(V) Sorption and Diffusion on Montmorillonite Clay Progress Report M4FT-13LL08060327

This progress report (Milestone Number M4FT-13LL08060327) summarizes research conducted at LLNL within Work Package Number FT-13LL080603, Disposal Research – Generic Engineered Barrier System Evaluations. The research is focused on quantifying Np(V) sorption to and diffusion through montmorillonite clay which is the primary mineralogic component of proposed bentonite backfill material in certain nuclear waste repository scenarios.

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DOE OSTI2015

Feasibility studies for the design of disposal vaults for multi-purpose containers (MPCs): construction, operation, ventilation and management of heat - 15175

The Nuclear Decommissioning Authority (NDA) subsidiary, Radioactive Waste Management Ltd (RWM) is responsible for the safe implementation of a Geological Disposal Facility (GDF) for radioactive waste in the UK. As part of their generic programme to underpin the planned designs for a GDF, work is ongoing to explore a range of possible geological disposal concepts. This work developed GDF concept designs for Multi-Purpose Containers (MPC) disposal in the generic host rock types considered: higher strength (crystalline), lower strength sedimentary (sedimentary / clay formations) and evaporite rocks. This paper is specific to concepts for the disposal of MPC for Pressurised Water Reactor (PWR) spent nuclear fuel (SF). The key outcomes of the feasibility studies to improve the understanding of the MPC disposal concept and the specific challenges to construct, operate, ventilate, manage heat and backfill associated with this concept are summarised in this paper. The management of the high loading of radiogenic heat (from 12 PWR SF assemblies in MPCs compared to 4 in 'standard' disposal containers) during an extended open phase (delayed backfilling of the disposal area) of up to 300 years was investigated. The concept designs developed demonstrate it is feasible to construct, operate and ventilate disposal vaults to accommodate open 'storage' periods required for transitional thermal decay of high heat generating waste in higher strength and lower strength sedimentary host rock types considered by RWM. This work focused on developing two contrasting disposal concepts which are applicable to both higher strength and lower strength rock environments; both of which illustrate how vault temperature, stability and groundwater inflows can be controlled to delay backfilling and allow MPC package thermal output to decrease. MPC disposal in evaporite rock does not require a cooling period due to its high thermal conductivity and favourable performance at higher temperatures; therefore the concept developed for this rock includes immediate backfilling of disposal tunnels. A clear understanding of the interactions and limitations associated with heat management during the open period was developed by investigation of a 'Heating Ventilation and Air-Conditioning System' (HVAC). It was concluded that removal of heat from the disposal tunnels / vaults is feasible, but introduces significant operational complexity. (authors)

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DOE OSTI2011

Disposal Systems Evaluation Framework (DSEF) Version 1.0 - Progress Report

The Disposal Systems Evaluation Framework (DSEF) is being developed at Lawrence Livermore National Laboratory to formalize the development and documentation of repository conceptual design options for each waste form and environment combination. This report summarizes current status and plans for the remainder of FY11 and for FY12. This progress report defines the architecture and interface parameters of the DSEF Excel workbook, which contains worksheets that link to each other to provide input and document output from external codes such that concise comparisons between fuel cycles, disposal environments, repository designs and engineered barrier system materials can be performed. Collaborations between other Used Fuel Disposition Campaign work packages and US Department of Energy / Nuclear Energy campaigns are clearly identified. File naming and configuration management is recommended to allow automated abstraction of data from multiple DSEF runs.

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DOE OSTI2023

SFWST Disposal Research R&D 5-Year Plan (FY2023 Update)

This FY2023 report is the second update to the Disposal Research (DR) Research and Development (R&D) 5-year plan for the Spent Fuel and Waste Science and Technology (SFWST) Campaign DR R&D activities. In the planning for FY2020 in the U.S. Department of Energy (DOE) NE-81 SFWST Campaign, the DOE requested development of a high-level summary plan for activities in the DR R&D program for the next five (5)-year period, with periodic updates to this summary plan. The DR R&D 5-year plan was provided to the DOE based initially on the FY2020 priorities and program structure (initial 2020 version of this 5-year plan) and provides a strategic summary guide to the work within the DR R&D technical areas (Control Accounts, CA), focusing on the highest priority technical thrusts. This 5-year plan is a living document (planned to be updated periodically) that provides review of SFWST R&D accomplishments (as seen on the 2021 revision of this 5-year plan), describes changes to technical R&D prioritization based on (a) progress in each technical area (including external technical understanding) with specific accomplishments and (b) any changes in SFWST Campaign objectives and/or funding levels (i.e., Program Direction). Updates to this 5-year plan include the DR R&D adjustments to high-priority knowledge gaps to be investigated in the near-term, as well as the updated longer-term DR R&D directions for the program activities. This plan fulfills the Milestone M2SF23SN010304083 in DR Work Package (WP) SF-23SN01030408 (GDSA - Framework Development – SNL).

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DOE OSTI2010

Separation of packaging plastics by froth flotation in a continuous pilot plant

The objective of the research was to apply froth flotation to separate post-consumer PET (Polyethylene Terephthalate) from other packaging plastics with similar density, in a continuously operated pilot plant. A representative sample composed of 85% PET, 2.5% PVC (Polyvinyl Chloride) and 11.9% PS (Polystyrene) was subjected to a combination of alkaline treatment and surfactant adsorption followed by froth flotation. A mineral processing pilot plant, owned by a Portuguese mining company, was adapted for this purpose. The experimentation showed that it is possible to produce an almost pure concentrate of PET, containing 83% of the PET in feed, in a single bank of mechanical flotation cells. The concentrate grade attained was 97.2% PET, 1.1% PVC and 1.1% PS. By simulation it was shown that the Portuguese recycling industry specifications can be attained if one cleaning and one scavenger stages are added to the circuit.

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DOE OSTI2024

Long-term Moisture Adsorption in Packaged Plutonium Oxide

In 2018, the Department of Energy National Nuclear Security Administration (DOE-NNSA) began implementing dilute and dispose to remove 34 metric tons (MT) of surplus weapons grade plutonium from the US stockpile. Under this plan, surplus plutonium material is converted into plutonium oxide (PuO 2 ) before being stored in metal containers and sent to the DOE Waste Isolation Pilot Plant (WIPP). The dilute and dispose project was implemented as a more cost-effective method for abiding by the Plutonium Management and Disposition Agreement (PMDA) between the USA and Russia, as compared to producing mixed oxide fuel (MOX). The PMDA was originally signed in 2000 and amended in 2010.2 The disassembly of pits and conversion to PuO 2 as part of dilute and dispose is carried out through the Advanced Recovery and Integrated Extraction System (ARIES) developed at Los Alamos National Laboratory (LANL). The dilute portion of dilute and dispose is carried out at Savannah River Site (SRS). With this program in place, it is necessary to ensure the safe, long-term storage of the PuO 2 generated during this process until final disposal at WIPP is accomplished.

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DOE OSTI2017

Nuclear Material Management Database Development at the Nevada National Security Site

The Nuclear Material Management & Storage (NMM&S) program at the Nevada National Security Site (NNSS) have the responsibility for the life-cycle management of accountable nuclear materials (NM) in accordance with Department of Energy (DOE) Order 410.2, “Management of Nuclear Materials” and for ensuring workers are protected from exposure to nuclear materials stored outside of an authorized shipping package under DOE Manual 441.1-1. To ensure these responsibilities were addressed in an effective, concise, and secure manner the development of NMM&S databases was undertaken. Using the existing database server structure and data acquired from the official Material Control & Accountability (MC&A) accounting program development of two separate applications, one for O 410.2 inventory requirements (NMIA) and one for M441, were begun in September 2015. The databases were deployed onto the production server in January 2017. The inventory database allows for entering NM characterization data, production of O 410.2 reports, and the ability to retain history files. The M441 database screens NM to determine if it is subject to the manual's requirements, ability to pull in unclassified RadCon materials, and provide reports as required by NMM&S, NM owners, and headquarters programs. These two databases provide NMM&S the ability to better ensure safe and secure packaging, storage, stabilization, consolidation, or disposition of NM and achieve life-cycle management goals.

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DOE OSTI2017

Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 17374

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license, granted by the competent authority in this field, which has been the Federal Office of Radiation Protection (BfS) until July 2016 and currently is the Federal Office for the Safety of Nuclear Waste Management (BfE). A first license was granted in 1983 for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) at the interim storage facility in Gorleben. This license was reviewed in 1995. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since that time at each site of an operational NPP storage facilities were licensed and erected. Thus additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow) there are now 12 storage facilities onsite the NPPs. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. The German 'Programme for the safe and responsible management of spent fuel and radioactive waste' expects to take the disposal facility into operation in the year 2050. Thus necessary preparations for interim storage till that time have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and a consulting of the German parliament (Bundestag) is mandatory if inevitable events would require an extension of currently licensed interim storage period. In the year 2016, the regulatory body in Germany was given a new structure. The BfS was formerly competent authority in the field of radiation protection since 1989. It had competencies for ionizing and non-ionizing radiation protection (e.g. mobile phone radiation, high voltage power lines and the effects of electric fields), licensing storage of spent nuclear fuel and as the operator of facilities for final disposal. Now there is a new structure with two Federal Offices (BfS and BfE) and a federal company for final disposal (BGE). The BfE is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF are now granted by this office. The search for a repository site will be coordinated there as well. The site selection, construction and operation of the repository will be in the responsibility of BGE. The parts of the former BfS dealing with recovery of the waste from Asse (a former salt mine used for disposal research) and the construction of the repository Konrad (for low and medium level radioactive waste) will be transferred into that company. It is planned to incorporate the German company for construction and management of waste disposal facilities (DBE) which is owned by the German utilities as well. The BfS will focus on the scientific foundations and further advancement of radiation protection. (authors)

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DOE OSTI2024

GDSA Repository Systems Analysis Investigations in FY 2024

The Disposal Research and Development (R&D) Program of the US Department of Energy (DOE) office of Nuclear Energy (NE-8) Spent Fuel and Waste Science and Technology (SFWST) Campaign is to conduct R&D on disposal of spent nuclear fuel (SNF) and high-level waste (HLW). The goal of the Geologic Disposal Safety Assessment (GDSA) within this project is to develop a disposal system modeling and analysis capability that supports the integrated modeling of coupled processes controlling disposal system performance of deep geologic repositories, including uncertainty. This report describes specific activities in the Fiscal Year (FY) 2024 associated with the GDSA Repository Systems Analysis (RSA) work package. The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and repository system performance models in crystalline, argillite, salt, and unsaturated alluvium potential host-rock environments, and to simulate and analyze these generic repository concepts and models using GDSA Framework toolkit, and other tools as needed.

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NASA NTRS2023

Contractors Road Heavy Equipment Area (SWMU 055) 2022 Annual Groundwater Monitoring Report

This document presents a summary of activities completed at the Contractors Road Heavy Equipment (CRHE) Area, located at Kennedy Space Center (KSC), Florida, from June through December 2022. The activities conducted at the CRHE Area include: - Annual groundwater sampling activities associated with sitewide plume monitoring and Underground Injection Control (UIC) monitoring in the former bioremediation Interim Measures (IM) Hot Spot 1 (HS1) area; - Direct push technology (DPT) groundwater investigation; - Semiannual vapor intrusion (VI) evaluation activities. This facility is designated Solid Waste Management Unit (SWMU) Number 055 (SWMU 055) under KSC’s Resource Conservation and Recovery Act (RCRA) Corrective Action program. HydroGeoLogic, Inc. (HGL) prepared this report for the National Aeronautics and Space NASA) under contract number 80KSC019F0096/80KSC019D0012, Technical Directive-03. An Advance Data Package (ADP) presentation of the elements of this report received Team (KSCRT) agreement at the April 5, 2023, KSCRT meeting. The chlorinated volatile organic compound plume appears to be stable vertically and horizontally. Additional DPT groundwater sampling will be completed to support a remedial alternatives evaluation for potential means to expedite groundwater cleanup. There are per- and polyfluorinated alkyl substances in site groundwater, and assessment is planned in the future. Sub-slab soil gas monitoring results do not indicate any concerns with indoor vapor intrusion. Based on the 2022 annual groundwater sampling results, it is recommended that annual sampling continue to alternate with the wet and dry season, with the next sampling event planned for December 2023.

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DOE OSTI2012

Thermal Removal Of Tritium From Concrete And Soil To Reduce Groundwater Impacts

Legacy heavy-water moderator operations at the Savannah River Site (SRS) have resulted in the contamination of equipment pads, building slabs, and surrounding soil with tritium. At the time of discovery the tritium had impacted the shallow (< 3-m) groundwater at the facility. While tritium was present in the groundwater, characterization efforts determined that a significant source remained in a concrete slab at the surface and within the associated vadose zone soils. To prevent continued long-term impacts to the shallow groundwater a CERCLA non-time critical removal action for these source materials was conducted to reduce the leaching of tritium from the vadose zone soils and concrete slabs. In order to minimize transportation and disposal costs, an on-site thermal treatment process was designed, tested, and implemented. The on-site treatment consisted of thermal detritiation of the concrete rubble and soil. During this process concrete rubble was heated to a temperature of 815 deg C (1,500 deg F) resulting in the dehydration and removal of water bound tritium. During heating, tritium contaminated soil was used to provide thermal insulation during which it's temperature exceeded 100 deg C (212 deg F), causing drying and removal of tritium. The thermal treatment process volatiles the water bound tritium and releases it to the atmosphere. The released tritium was considered insignificant based upon Clean Air Act Compliance Package (CAP88) analysis and did not exceed exposure thresholds. A treatability study evaluated the effectiveness of this thermal configuration and viability as a decontamination method for tritium in concrete and soil materials. Post treatment sampling confirmed the effectiveness at reducing tritium to acceptable waste site specific levels. With American Recovery and Reinvestment Act (ARRA) funding three additional treatment cells were assembled utilizing commercial heating equipment and common construction materials. This provided a total of four units to batch treat concrete rubble and soil. Post treatment sampling verified that the activity in the treated soil and concrete met the treatment standards for each medium which allowed the treated concrete rubble and soil to be disposed of on site as backfill. During testing and operations a total of 1,261-m{sup 3} (1,650-yd{sup 3}) of contaminated concrete and soils were treated with an actual incurred cost of $3,980,000. This represents a unit treatment cost of $3,156/m{sup 3} ($2,412/yd{sup 3}). In 2011 the project was recognized with an e-Star Sustainability Award by DOE's Office of Environmental Management.

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DOE OSTI2013

Thermal Removal of Tritium from Concrete and Soil to Reduce Groundwater Impacts - 13197

Legacy heavy-water moderator operations at the Savannah River Site (SRS) have resulted in the contamination of equipment pads, building slabs, and surrounding soil with tritium. At the time of discovery the tritium had impacted the shallow (< 3-m) groundwater at the facility. While tritium was present in the groundwater, characterization efforts determined that a significant source remained in a concrete slab at the surface and within the associated vadose zone soils. To prevent continued long-term impacts to the shallow groundwater a CERCLA non-time critical removal action for these source materials was conducted to reduce the leaching of tritium from the vadose zone soils and concrete slabs. In order to minimize transportation and disposal costs, an on-site thermal treatment process was designed, tested, and implemented. The on-site treatment consisted of thermal detritiation of the concrete rubble and soil. During this process concrete rubble was heated to a temperature of 815 deg. C (1,500 deg. F) resulting in the dehydration and removal of water bound tritium. During heating, tritium contaminated soil was used to provide thermal insulation during which it's temperature exceeded 100 deg. C (212 deg. F), causing drying and removal of tritium. The thermal treatment process volatiles the water bound tritium and releases it to the atmosphere. The released tritium was considered insignificant based upon Clean Air Act Compliance Package (CAP88) analysis and did not exceed exposure thresholds. A treatability study evaluated the effectiveness of this thermal configuration and viability as a decontamination method for tritium in concrete and soil materials. Post treatment sampling confirmed the effectiveness at reducing tritium to acceptable waste site specific levels. With American Recovery and Reinvestment Act (ARRA) funding three additional treatment cells were assembled utilizing commercial heating equipment and common construction materials. This provided a total of four units to batch treat concrete rubble and soil. Post treatment sampling verified that the activity in the treated soil and concrete met the treatment standards for each medium which allowed the treated concrete rubble and soil to be disposed of on-site as backfill. During testing and operations a total of 1,261-m{sup 3} (1,650-yd{sup 3}) of contaminated concrete and soils were treated with an actual incurred cost of $3,980,000. This represents a unit treatment cost of $3,156/m{sup 3} ($2,412/yd{sup 3}). In 2011 the project was recognized with an e-Star Sustainability Award by DOE's Office of Environmental Management. (authors)

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DOE OSTI2010

CH Packaging Operations Manual (Rev. 10.1)

This document provides the user with instructions for assembling a payload. All the steps in Subsections 1.2, Preparing 55-Gallon Drum Payload Assembly; 1.3, Preparing "Short" 85-Gallon Drum Payload Assembly (TRUPACT-II and HalfPACT); 1.4, Preparing "Tall" 85-Gallon Drum Payload Assembly (HalfPACT only); 1.5, Preparing 100-Gallon Drum Payload Assembly; 1.6, Preparing Shielded Container Payload Assembly; 1.7, Preparing SWB Payload Assembly; and 1.8, Preparing TDOP Payload Assembly, must be completed, but may be performed in any order as long as radiological control steps are not bypassed. Transport trailer operations, package loading and unloading from transport trailers, hoisting and rigging activities such as ACGLF operations, equipment checkout and shutdown, and component inspection activities must be performed, but may be performed in any order and in parallel with other activities as long as radiological control steps are not bypassed. Steps involving OCA/ICV lid removal/installation and payload removal/loading may be performed in parallel if there are multiple operators working on the same packaging. Steps involving removal/installation of OCV/ICV upper and lower main O-rings must be performed in sequence, except as noted.

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DOE OSTI2017

X-ray tomography of feed-to-glass transition of simulated borosilicate waste glasses

The feed composition of a high level nuclear waste (HLW) glass melter affects the overall melting rate by influencing the chemical, thermophysical, and morphological properties of a relatively insulating cold cap layer over the molten phase where the primary feed vitrification reactions occur. Data from X ray computed tomography imaging of melting pellets comprised of a simulated high-aluminum HLW feed heated at a rate of 10°C/min reveal the distribution and morphology of bubbles, collectively known as primary foam, within this layer for various SiO 2 /(Li 2 CO 3 +H 3 BO 3 +Na 2 CO 3 ) mass fractions at temperatures between 600°C and 1040°C. To track melting dynamics, cross-sections obtained through the central profile of the pellet were digitally segmented into primary foam and a condensed phase. Pellet dimensions were extracted using Photoshop CS6 tools while the DREAM.3D software package was used to calculate pellet profile area, average and maximum bubble areas, and two-dimensional void fraction. The measured linear increase in the pellet area expansion rates – and therefore the increase in batch gas evolution rates – with SiO 2 /(Li 2 CO 3 +H 3 BO 3 +Na 2 CO 3 ) mass fraction despite an exponential increase in viscosity of the final waste glass at 1050°C and a lower total amount of gas-evolving species suggest that the retention of primary foam with large average bubble size at higher temperatures results from faster reaction kinetics rather than increased viscosity. However, viscosity does affect the initial foam collapse temperature by supporting the growth of larger bubbles. Because the maximum bubble size is limited by the pellet dimensions, larger scale studies are needed to understand primary foam morphology at high temperatures. This temperature-dependent morphological data can be used in future investigations to synthetically generate cold cap structures for use in models of heat transfer within a HLW glass melter.

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DOE OSTI2019

Evaluation of Aluminum-clad Spent Nuclear Fuel during Drying and Dry Storage

The DOE SNF Working Group ASNF Subgroup and the U.S. Nuclear Waste Technical Review Board have identified knowledge gaps regarding the management of aluminum-clad spent nuclear fuel (ASNF).[1, 2] The objectives of this research are to observe, sample and analyze the nature of aluminum surface corrosion and corrosion layer chemistry relevant to the ASNF performance over the duration of the dry storage segment of the fuel life cycle. The primary goal is to build an understanding of ASNF behavior, from reactor service and wet storage through drying and dry storage, with parallel testing of surrogate materials. The premise is that residual water affiliated with the aluminum oxide from corrosion during reactor service and wet storage, continues to influence ASNF performance during dry storage, and that even after successful drying the surface conditions of ASNF have yet to be satisfactorily characterized. The initial scope collected and analyzed specimens from irradiated non-fuel components of aluminum-clad Advanced Test Reactor (ATR) fuel elements that are routinely removed at the end of reactor service life. Techniques have been developed for remote material sampling from ATR elements from vented dry storage at the Irradiated Fuel Storage Facility (IFSF). These techniques involve 1) nipping off a small corner of the aluminum near where the end box had been removed and 2) scraping the exterior (non-fueled) plate surface to collect an oxide sample. This second effort targets ATR elements expected to have seen the longest residence in storage, both in the CPP-603 pool (with poor water quality) and in dry storage at IFSF, within the fuel handling scheduled to consolidate ATR at IFSF and receive the remaining inventory at CPP-666. Follow on work has been proposed to nip corners and collect scrapings immediately following ATR fuel drying, before placement in dry storage. Sample analysis will use multiple techniques to understand the chemical composition and morphology of the surface oxides. A simultaneous effort is progressing to develop surrogate materials to enable rigorous controlled experiments. Consistently produced, these surrogate materials can then be used to demonstrate limitations inherent to remote sampling and to validate performance models under relevant constraints. The combined efforts will substantiate ASNF performance and fuel management decisions regarding drying and packaging over a more complete range of life cycle conditions. [1] DOE Spent Nuclear Fuel Working Group, Aluminum-Clad Spent Nuclear Fuel: Technical Considerations and Challenges for Extended (>50 Years) Dry Storage, DOE/ID RPT#: 1575, prepared by Aluminum-Clad SNF Sub Working Group, June 2017. [2] U.S. Nuclear Waste Technical Review Board, Management and Disposal of U.S. Department of Energy Spent Nuclear Fuel, A Report to the United States Congress and the Secretary of Energy, December 2017, p. 143.

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DOE OSTI2023

Quantifying the environmental benefits of a solvent-based separation process for multilayer plastic films

Food packaging often appears in the form of multilayer (ML) plastic films, which leverage the functional properties of different polymers to achieve specific food protection goals (e.g., oxygen, water, and temperature barriers). These properties are essential to enable long shelf lives, reduce refrigeration usage, mitigate food waste, and increase food accessibility. However, ML film production processes generate large amounts of plastic waste that cannot be mechanically recycled. Recently, we have proposed a process, called solvent-targeted recovery and precipitation (STRAP), that enables the separation and recycling of the constituent polymers of ML films. This technology uses a series of solvent washes that selectively dissolve and precipitate target polymers. Quantifying the environmental benefits of STRAP over virgin resin production is essential for the commercial deployment of this technology. Further, this work uses life cycle assessment (LCA) methods to evaluate these impacts in terms of carbon footprint, energy use, water use, and toxicity. We analyze three STRAP process variants that use antisolvent and temperature-driven precipitation to treat different ML films. Our analysis reveals that the STRAP-A and STRAP-B process variants can provide environmental benefits over virgin film production. Furthermore, it gives valuable insight into the critical components of ML films (specific polymers) and of the STRAP processes (equipment) that are responsible for the highest impacts. Ultimately, we believe that the proposed analysis framework can lead to the design of more environmentally-friendly ML films and recycling processes.

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