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This plan defines the requirements for managing the transportation of transuranic (TRU) and TRU mixed waste from a generator/storage site to the Waste Isolation Pilot Plant (WIPP) or to another generator/storage site (inter-site). This plan was developed in accordance with Department of Energy (DOE) Order 460.2A, Departmental Materials Transportation and Packaging Management.
The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. The DOE standardized canister will be relied upon to confine radionuclides and preclude moderator in order to ensure safe, long-term management and disposition of DOE SNF. This strategy offers an avenue for avoiding the characterization costs (i.e. funds and exposure) and other technical challenges associated with the traditional approach of relying on fuel-specific properties as the basis for demonstrating compliance with the regulator. The use of the standardized canister offers the following benefits: • The standardized canister design is complete. It has been tested under a variety of drop and accident scenarios and was subsequently licensed under 10 CFR 72 as a storage system as part of the FWENC Facility. It is also part of the license submittal for Yucca Mountain. • There is high confidence that the standardized canister design can be stored within commercially available storage casks and licensed as part of a transportation-system package. • The robustness of the standardized canister simplifies accident analyses and does not rely on characterization of the contained waste. This simplifies safety analysis, provides a greater level of confidence with stake holders and reduces cost and risk. • Significant savings result from a common design for the associated packaging, closure, and loadout facilities. • Design efforts, procurement, and fabrication are simplified • Safety and surety of operations are enhanced by standardizing handling activities and associated training, procedures, equipment, and tooling. The standardized canister is the preferred package for these materials and has withstood scrutiny by the Yucca Mountain Repository License Application, the NRC’s Advisory Committee on Nuclear Waste, the Nuclear Waste Technical Review Board, an American Society of Mechanical Engineers peer review, and the Foster Wheeler Environmental Corporation storage license application. This evaluation summarizes and documents the status of the design and rationale for the DOE standardized canister and outlines the remaining needs to allow deployment. In addition, it discusses the factors that have the potential to influence storage, transportation, and disposal of DOE SNF.
Spent ion-exchange resins are considered to be problematic waste that, in many cases, requires special approaches and precautions during their conditioning to meet the acceptance criteria for disposal. In Belgium, for economical reasons, the Volume Reduction Factor is a key criterion. After Tractebel Engineering performed a technical and economical comparison of the industrially available systems, Tihange NPP decided to install a spent ion-exchange resins hot supercompaction unit with Tractebel Engineering in the role of architect-engineer. The treatment and conditioning unit processes the spent ion-exchange resins through the following steps: dewatering of the resins, drying the resins under deep vacuum, discharging the dried resins into compactable drums, super-compacting the drums to generate pellets, grouting the pellets into standard 400 litres waste drums (overpacks) licensed for final disposal in the near-surface repository in Belgium. Several developments were required to adapt the reference process and equipment to PWR spent ion-exchange bead resins and Belgian radioactive waste acceptance criteria. In order to avoid cracks on the compacted drum, and external surface contamination from resin leaks, some improvements were achieved to minimize spring-back as well as the risk of cracking the drum wall. Placing the compactable drum inside a second, slightly larger drum, guarantees clean and reproducible pellets. Currently the commissioning phase is on-going. Numerous process validation tests have been completed. An acceptance file was transmitted to the Belgian Waste Management Authority recently. This will be followed by demonstration tests necessary to obtain their final acceptance of the installation. More than 3 800 drums of mixed powdered and bead resins have been processed by the reference Hot Compaction process, achieving a Volume Reduction Factor (VRF) of 2.5. The equipment has been proven to be a reliable technology with low operation and maintenance costs. Tractebel Engineering has managed the construction of a new application of this process in Belgium at Tihange NPP. Several developments were required to adapt the reference process and equipment to PWR spent ion-exchange bead resins and Belgian radioactive waste acceptance criteria. The chosen method of conditioning (draining, drying and compaction of the spent resins followed by grouting of the pellets in a 400 litres drum) immobilizes the spent resins under the form of a solid, compact, stable, and non dispersible block free of interstitial water. The various series of inactive tests which were conducted at Tihange NPP, helped among others to determine the best design of the compactable drum and lid and to set the value of critical parameters such as vapour temperature at the end of drying, speed, force and duration of compaction. In an environment of very limited space for interim storage and in the absence of an operational final repository site, or in the case of high final disposal costs, the process exhibits the following key advantages: - Achieving a Volume Reduction Factor (VRF) close to 1 (overpack included) for the interim storage instead of increased volumes observed with other currently available processes; - Achieving a water free end product; - Creating a flexible waste product for interim storage (pellet), which can be retrieved and routed into alternative types of package later, if not initially grouted; - Using well proven standard technologies like drying and compaction; - Flexible use of the system components for the supercompaction of other operational solid waste streams when not conducting resins conditioning campaigns. (authors)
The Savannah River Site (SRS) disposes of low-level radioactive waste (LLW) and stabilizes high-level radioactive waste (HLW) tanks in the subsurface environment. Calculations used to establish the radiological limits of these facilities are referred to as Performance Assessments (PAs), Special Analyses (SAs), and Composite Analyses (CAs). The purpose of this document is to revise the existing geochemical data package used for these calculations (Kaplan, 2016). This work builds on earlier compilations of geochemical data, referred to a geochemical data packages (Kaplan, 2007; Kaplan, 2010; Kaplan, 2016; McDowell-Boyer et al., 2000). This work is being conducted as part of the on-going maintenance program of the SRS PA programs that periodically updates calculations and data packages when new information becomes available.
Oak Ridge National Laboratory (ORNL) produces transuranic (TRU) isotopes for the US Department of Energy's (US DoE's) heavy-element research program and other programs for medical and industrial applications. As a result, the research facilities generate remote-handled TRU liquid low-level waste (LLLW). This LLLW is presently collected and mixed with larger volumes of legacy and environmental management waste streams for future treatment. Once the legacy waste is disposed of as part of the US DOE Environmental Management program, the large-scale waste treatment processes will be decommissioned, and new capabilities will be needed to characterize, treat, package, and dispose of the newly generated LLLW at ORNL. Various approaches are being considered for addressing treatment and disposal of the future LLLW streams. A variety of processing methods and final waste forms is under consideration. The impacts of the chemical and radionuclide compositions of the waste stream are being evaluated. (authors)
The cost of sending mass into space with current propulsion technology is very expensive, making every item a crucial element of the space mission. It is essential that all materials be used to their fullest potential. Items like food, packaging, clothing, paper towels, gloves, etc., normally become trash and take up space after use. These waste materials are currently either burned up upon reentry in earth's atmosphere or sent on cargo return vehicles back to earth: a very wasteful method. The purpose of this project was to utilize these materials and create useful products like water and methane gas, which is used for rocket fuel, to further supply a deep space mission. The system used was a thermal degradation reactor with the configuration of a down-draft gasifier. The reactor was loaded with approximately 100g of trash simulant and heated with two external ceramic heaters with separate temperature control in order to create pyrolysis and gasification in one zone and incineration iri a second zone simultaneously. Trash was loaded into the top half of the reactor to undergo pyrolysis while the downdraft gas experienced gasification or incineration to treat tars and maximize the production of carbon dioxide. Minor products included carbon monoxide, methane, and other hydrocarbons. The carbon dioxide produced can be sent to a Sabatier reactor to convert the gas into methane, which can be used as rocket propellant. In order to maximize the carbon dioxide and useful gases produced, and minimize the unwanted tars and leftover ashen material, multiple experiments were performed with altered parameters such as differing temperatures, flow rates, and location of inlet air flow. According to the data received from these experiments, the process will be further scaled up and optimized to ultimately create a system that reduces trash buildup while at the same time providing enough useful gases to potentially fill a methane tank that could fuel a lunar ascent vehicle or other deep space mission.
This work is being performed as part of the DOE NE Spent Fuel and Waste Science and Technology Campaign, Argillite and Crystalline Rock R&D work packages: SF-20AN01030101 and SF-20AN01030201 respectively. This document is submitted in fulfillment of milestone M4SF-20AN010301013 for Argillite R&D and the milestone M4SF-20AN010302013 for Crystalline R&D
When transporting highly radioactive material, it is vital to provide a successful delivery. Without proper containment, radioactive materials can become exposed to the surrounding environment and pose a serious threat. To ensure a safe transition of radioactive material, packaging components must be able to withstand high levels of heat during the event of a fire while maintaining the integrity of the structure to safety contain the radioactive material. Various types of polyurethane and silicone have been selected to observe at high temperatures to analyze outgassing characteristics of the materials.
EnergySolutions and its team partners, NAC International, Exelon Nuclear Partners, Talisman International, TerranearPMC, Booz Allen Hamilton and Sargent and Lundy, have carried out a study to develop concepts for a Consolidated Storage Facility (CSF) for the USA's stocks of commercial Used Nuclear Fuel (UNF), and the packaging and transport provisions required to move the UNF to the CSF. The UNF is currently stored at all 65 operating nuclear reactor sites in the US, and at 10 shutdown sites. The study was funded by the US Department of Energy and followed the recommendations of the Blue Ribbon Commission on America's Nuclear Future (BRC), one of which was that the US should make prompt efforts to develop one or more consolidated storage facilities for commercial UNF. The study showed that viable schemes can be devised to move all UNF and store it at a CSF, but that a range of schemes is required to accommodate the present widely varying UNF storage arrangements. Although most UNF that is currently stored at operating reactor sites is in water-filled pools, a significant amount is now dry stored in concrete casks. At the shutdown sites, the UNF is dry stored at all but two of the ten sites. Various types of UNF dry storage configurations are used at the operating sites and shutdown sites that include vertical storage casks that are also licensed for transportation, vertical casks that are licensed for storage only, and horizontally orientated storage modules. The shutdown sites have limited to nonexistent UNF handling infrastructure and several no longer have railroad connections, complicating UNF handling and transport off the site. However four methods were identified that will satisfactorily retrieve the UNF canisters within the storage casks and transport them to the CSF. The study showed that all of the issues associated with the transportation and storage of UNF from all sites in the US can be accommodated by adopting a staged approach to the construction of the CSF. Stage 1 requires only a cask storage pad and railroad interface to be constructed, and the CSF can then receive the UNF that is in transportable storage casks. Stage 2 adds a canister handling facility, a storage cask fabrication facility and an expanded storage pad, and enables the receipt of all canistered UNF from both operating and shutdown sites. Stage 3 provides a repackaging facility with a water-filled pool that provides flexibility for a range of repackaging scenarios. This includes receiving and repackaging 'bare' UNF into suitable canisters that can be placed into interim storage at the CSF, and enables UNF that is being received, or already in storage onsite, to be repackaged into canisters that are suitable for disposal at a geologic repository. The study used the 'Total System Model' (TSM) to analyze a range of CSF capacities and operating scenarios with differing parameters covering UNF pickup orders, one or more CSF sites, CSF start dates, CSF receipt rates and geologic repository start dates. The TSM was originally developed to model movement of UNF to the Yucca Mountain repository and was modified for this study to enable the CSF to become the 'gateway' to a future geologic repository. The TSM analysis enabled costs to be estimated for each scenario and showed how these are influenced by each of the parameters. This information will provide essential underpinning for a future Conceptual Design preparation. (authors)
This report represents completion of milestone deliverable M2SF-18SNO10309013 "Inventory and Waste Characterization Status Report and OWL Update that reports on FY2018 activities for the work package (WP) SF-18SNO1030901. This report provides the detailed final information for completed FY2018 work activities for WP SF-18SN01030901, and a summary of priorities for FY2019. This status report on FY2018 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the two planned management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the Beta testing and changes implemented from it. There are two processes being planned in FY2018, which will be implemented in FY2019. One process covers methods for interfacing with the DOE SNF DB (DOE 2007) at INL on the numerous entries for DOE managed SNF, and the other process covers the management of updates to, and version control/archiving of, the OWL database. In FY2018, we have pursued three studies to evaluate/redefine waste form characteristics and/or performance models. First characteristic isotopic ratios for various waste forms included in postclosure performance studies are being evaluated to delineate isotope ratio tags that quantitatively identify each particular waste form. This evaluation arose due to questions regarding the relative contributions of radionuclides from disparate waste forms in GDSA results, particularly, radionuclide contributions of DOE-managed SNF vs HLW glass. In our second study we are evaluating the bases of glass waste degradation rate models to the HIP calcine waste form. The HIP calcine may likely be a ceramic matrix material, with multiple ceramic phases with/without a glass phase. The ceramic phases are likely to have different degradation performance from the glass portion. The distribution of radionuclides among those various phases may also be a factor in the radionuclide release rates. Additionally, we have an ongoing investigation of the performance behavior of TRISO particle fuels and are developing a stochastic model for the degradation of those fuels that accounts for simultaneous corrosion of the silicon carbide (SiC) layer and radionuclide diffusion through it. The detailed model of the TRISO particles themselves, will be merged with models of the degradation behavior(s) of the graphite matrix (either prismatic compacts or spherical "pebbles") containing the particles and the hexagonal graphite elements holding the compacts.
The Calcine Disposition Project is responsible for retrieval, treatment by hot-isostatic pressure, packaging, and disposal of highly radioactive calcine stored at the Idaho Nuclear Technology and Engineering Center at the Idaho National Laboratory Site in southeast Idaho. In the 2009 Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement the Department of Energy documented the selection of hot-isostatic pressure as the technology to treat the calcine. The Record of Decision specifies that the treatment results in a volume-reduced, monolithic waste form suitable for transport outside of Idaho by a target date of December 31, 2035. That target date is specified in the 1995 Idaho Settlement Agreement to treat and prepare the calcine for transport out of Idaho in exchange for allowing storage of Navy spent nuclear fuel at the INL Site. The project is completing the design of the calcine-treatment process and facility to comply with Record of Decision, Settlement Agreement, Idaho Department of Environmental Quality, and Department of Energy requirements. A systems engineering approach is being used to define the project mission and requirements, manage risks, and establish the safety basis for decision making in compliance with DOE O 413.3B, 'Program and Project Management for the Acquisition of Capital Assets'. The approach draws heavily on 'design-for-quality' tools to systematically add quality, predict design reliability, and manage variation in the earliest possible stages of design when it is most efficient. Use of these tools provides a standardized basis for interfacing systems to interact across system boundaries and promotes system integration on a facility-wide basis. A mass and energy model was developed to assist in the design of process equipment, determine material-flow parameters, and estimate process emissions. Data generated from failure modes and effects analysis and reliability, availability, maintainability, and inspectability analysis were incorporated into a time and motion model to validate and verify the capability to complete treatment of the calcine within the required schedule. The Calcine Disposition Project systems engineering approach, including use of industry-proven design-for-quality tools and quantitative assessment techniques, has strengthened the project's design capability to meet its intended mission in a safe, cost-effective, and timely manner. Use of these tools has been particularly helpful to the project in early design planning to manage variation; improve requirements and high-consequence risk management; and more effectively apply alternative, interface, failure mode, RAMI, and time and motion analyses at the earliest possible stages of design when their application is most efficient and cost effective. The project is using these tools to design and develop HIP treatment of highly radioactive calcine to produce a volume-reduced, monolithic waste form with immobilization of hazardous and radioactive constituents. (authors)
The potential of lignocellulosic fibers obtained by dry grinding of pinhão coat as fillers in starch filmogenic solutions for packaging applications was evaluated in this work. To improve the incorporation of this waste into the starch solutions different physical and chemical treatments were conducted. Thereafter, morphology, chemical structure, crystallinity and water absorption of the pinhão coat powders were determined. The composites were also characterized regarding their morphology, chemical structure, crystallinity, mechanical properties, water vapor permeability and hydrophilicity. Poor fiber/matrix adhesion and high water absorption of the fibers were evidenced. Consequently, water vapor permeability of composites was increased by incorporating the fibers. Moreover, mechanical properties were improved and the morphological results were used to support the water absorption differences among the powders. Regarding the food packaging applications, starch/pinhão coat composites appeared as promising materials to reach the requirements of respiring food products.
The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation has sponsored a suite of training courses that have been conducted since the early 1990s. Of these training courses, the Savannah River National Laboratory (SRNL) is currently leading the effort for two courses: (1) Management of SARP Preparation, and (2) Nuclear and Other Radioactive Materials Transport Security and Leak Testing. This paper provides an overview of both courses. (authors)
Packages used to transport spent nuclear fuel (SNF) are required by the U.S. Code of Federal Regulations 10 CFR 71.71 to demonstrate satisfactory performance during a drop scenario. While the CFR is meant to ensure safe package function, it does not evaluate survival of the SNF within. The U.S. Department of Energy Spent Fuel and Waste Science and Technology program is working on closing the knowledge gap related to the response of SNF to external mechanical loads, including the hypothetical 30 cm package drop scenario in the CFR. In support of this effort, LS-DYNA finite element simulations were developed by Pacific Northwest National Laboratory (PNNL) to model generic drop scenarios at both the package and fuel assembly level. The models were validated against one-third scale package and full scale fuel assembly drop test data and were exercised to predict fuel cladding strains in a narrow range of model configurations. This work describes a large-scale parametric study conducted by PNNL using the previously developed and validated PWR finite element model, with the addition of a new generic BWR assembly model. The motivation for the parametric study was to characterize the broad range of SNF responses in the 30 cm package drop scenario. This was accomplished by varying the drop orientation, fuel assembly type (17x17 PWR and 10x10 BWR), burnup, cladding temperature, spacer grid buckling load, package mass, impact limiter stiffness, and mechanical gap conditions within the basket. A MATLAB framework was developed to automate LS-DYNA model generation and execution on PNNL institutional computing resources. In total, over 2000 simulations were performed. For each simulation, the SNF response was quantified in terms of permanent grid deformation, fuel rod contact pressure, and strains within the fuel rods, guide tubes, and water rods. The results provide valuable insight into the range of responses that could be reasonably expected from SNF in the hypothetical drop scenario, as well as the sensitivity to each input parameter. The results of this parametric study are a key component of the testing and modeling strategy the Spent Fuel and Waste Science and Technology program is using to close the external loads knowledge gap.
The Disposal Research & Development (Disposal R&D) Campaign of the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), Office of Spent Fuel & High-Level Waste Disposition is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high-level nuclear waste (HLW). A high priority for Disposal R&D is disposal system modeling (Sassani et al. 2023). The Geologic Disposal Safety Assessment (GDSA) work package is charged with developing a disposal system modeling and analysis capability for evaluating generic disposal system performance for nuclear waste in geologic media.
Food packaging is one of the leading sectors for the end use of plastics. Bioplastic is produced from natural renewable resources such as crops, wood pulp, and herbaceous fibers. This article summarizes the rational design of bioplastics from natural resources for food packaging. The bioplastic properties considered include thermal properties, mechanical performance, oxygen/moisture resistance, and biodegradability. Bioplastic degradability and technologies for handling bioplastic waste are discussed. Various aspects of the sustainability of bioplastics (e.g., environmental profile, techno-economic analysis, and societal impact) are investigated. The main challenges of bioplastic application, such as low fracture strain and inferior barrier properties, are discussed. Mitigation approaches to overcome the mechanical properties of bioplastics, such as adding plasticizers, are also discussed. Bioplastic can have properties comparable with fossil-based plastics. In conclusion, bioplastic can be an alternative to conventional plastic in most applications of food packaging, which can reduce the carbon footprint and environmental impact because of its biodegradability.
The U.S. Department of Energy's (DoE's) Used Fuel Disposition Campaign (UFDC), in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry have initiated a group of projects to obtain data to support the enhancement of the technical bases for the extended storage and transportation of high-burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF). Under this overall effort, a U.S. Nuclear Regulatory Commission-licensed storage cask was loaded with Thirty-Two (32) HBU SNF assemblies. The cask was modified to allow radial and axial temperature profiles to be measured using thermocouple lances inserted through the lid. Twenty-five (25) rods from similar assemblies of similar claddings, in-reactor placement, and burnup histories (herein called 'sister rods') have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory (ORNL). To meet UFD's research objectives, testing on ten (10) of these rods is required to be performed at Pacific Northwest National Laboratory (PNNL). Although the twenty-five rods were originally received at ORNL in the 10-160B, the PNNL facility that would need to receive the ten (10) rods did not have an active history with using the 10-160B. This paper first identifies and summarizes various criteria and evaluations used in helping to select which package will be used for the shipment of the ten (10) sister rods to PNNL (i.e., Bases on why the NACLWT was chosen). With the shipment scheduled to be complete, identified potential concerns, issues, lessons learned, and transportation summary documents were studied (post-shipment). The adequacy of the criteria and evaluations used to choose the NAC-LWT were accessed to determine their adequacy. The results show the criteria were adequate. However, it was noted that DOE Order 460.2A, DOE M 460.2-1A, and DOE G 460.2-1 have differences in the terminology and scope, resulting in clarity issues and differences in regulatory interpretation. Since one-of-a kind spent fuel shipments are uncommon, the topic will be significant interest to people who work in radioactive material packaging and transportation. (authors)
The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Spent Fuel and Waste Science & Technology (SFWST) research program is guided by the high-level goal of closing prioritized knowledge gaps related to spent nuclear fuel (SNF) storage and transportation, which are summarized by Saltzstein et al. (2020). One of the high-priority knowledge gaps is the identification and quantification of mechanical loads that are expected to affect SNF during normal conditions of transportation and storage to inform the range of physical SNF test programs. This report uses modeling and analysis methods to estimate the mechanical loads on spent nuclear fuel (SNF) in the general 30 cm package drop scenario. The drop scenario assumes impact limiters are in place in the transportation configuration and the impact surface is perfectly rigid. The goal of this analysis is to consider the universe of potential mechanical loading conditions that can happen to SNF and present the results in a manner that is useful for materials testing, decision making, and regulatory rule making purposes. This study uses validated finite element models and methods to perform a broad parametric study of key variables that can affect the mechanical loads on SNF during a hypothetical package free drop scenario. Physical drop test data from a cask and fuel assembly drop test campaign is the basis for model validation. Additionally, the results of the parametric study are used to inform a damage model, which uses multiple nonlinear regression to estimate the relationships between input variables and output response. The parametric finite element analyses consider thousands of input variable combinations, while the damage model estimates millions of combinations. The breadth of this study provides confidence that the potential range of mechanical loads that SNF might experience during the general 30 cm package drop scenario are characterized well enough to consider this knowledge gap closed. While this report documents the overall peak values calculated in this study, the 95 th percentile values, the histograms, and the observed trends are equally important. This study covered a large range of SNF temperatures, room temperature to 300°C, and burnups, 10 GWd/MTU to 62 GWd/MTU. Each temperature and burnup combination has a different cladding yield strain, so it is more meaningful to summarize the calculated cladding strain response as its factor of safety, which is defined relative to the yield strain. The factor of safety is calculated as the yield strain divided by the peak cladding strain. A factor of safety greater than unity indicates that the cladding remains below yield, whereas a value less than unity is indicative of plastic deformation. In all cases of this study a safety factor over 1.0 was calculated, although in the most limiting case at 300°C the safety factor was only 1.01, which suggests that yielding could occur when additional loads like rod internal pressure are included. When the temperature is restricted to 200°C the limiting safety factor increases to 1.28, which has significant margin to accommodate internal pressure and potential local cladding defects that could cause a local stress concentration. An important trend in the calculated fuel rod mechanical loads is that the 2 nd highest loaded fuel rod in an assembly tends to be significantly lower than the highest loaded rod. The implication is that even if one rod in an assembly experiences a failure the loads would have to be significantly higher to cause two or more rods to fail.