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At least 703 records · Page 39

Used fuel disposition in crystalline rocks

The U.S. Department of Energy Office of Nuclear Energy, Office of Fuel Cycle Technology established the Used Fuel Disposition Campaign (UFDC) in fiscal year 2010 (FY10) to conduct the research and development (R&D) activities related to storage, transportation and disposal of used nuclear fuel and high level nuclear waste. The objective of the Crystalline Disposal R&D Work Package is to advance our understanding of long-term disposal of used fuel in crystalline rocks and to develop necessary experimental and computational capabilities to evaluate various disposal concepts in such media.

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Plan for Development and Application of Risk Assessment Approach for Transportation Package Approval of an MNPP for Domestic Highway Shipment

For nuclear reactors, Probabilistic Risk Assessment (PRA) has been conducted since the 1970s [e.g., see WASH-1400 (NRC 1975), NUREG-1150 (NRC 1990), and NUREG-1935 (Chang et al. 2012)]. PRA has also been used to assess a dry cask storage system at a nuclear power plant [see NUREG-1864 (NRC 2007)]. PRA techniques have also been applied to the transportation of spent nuclear fuel, most notably in NUREG/CR-4829 (Fischer et al. 1987), NUREG/CR-6672 (Sprung et al. 2000), and NUREG-2125 (NRC 2014). Additional guidance is provided in International Atomic Energy Agency (IAEA) IAEA-TECDOC-1346 (2003). Transportation PRA was also used in the evaluations of transportation impacts in reports such as the Repository Final Environmental Impact Statement (DOE 2002) and the Repository Final Supplemental Environmental Impact Statement (DOE 2008).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of the B-25 Low Level Radioactive Waste Container in Taiwan - 1900x

Chinshan Nuclear Power Plant will be permanently stopped its operation in the middle of 2019 and will be the first nuclear power plant stepping into the decommissioning phase in Taiwan. According to Nuclear Reactor Facilities Regulation Act, a nuclear facility must be dismantled in 25 years after being permanently ceased operation. In its decommissioning plan, the B -25 low level radioactive waste container was one of the proposed containers used for this decommissioning task. However, the B-25 waste container was not used or developed in Taiwan before. To meet this need, Institute of Nuclear Energy Research developed a new type of B-25 waste container and will apply to the authority for its usage license. This article provided the information of the low level radioactive waste containers have been approved in Taiwan and described the design specifications of the developed B-25 waste container. Besides, the numerical evaluation and test results were also included in this article. Based on the evaluation and test results, the developed B-25 waste container satisfied the requirements of the storage container and industrial package type 2. It is believed that the usage license would be granted soon after applying it to the authority. (authors)

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Criticality Consideration for Geologic Disposal of DOE-Owned Spent Nuclear Fuels - 20255

Decisions must be made relative to packaging spent nuclear fuel (SNF) for storage, transportation, and eventual disposal. This paper evaluates potential criticality-related loading and packaging considerations for Department of Energy (DOE)-owned SNF that could be impacted favorably or unfavorably by disposal uncertainties and proposes approaches to minimize such impacts. Using Al-Based SNF as an example and drawing on analyses performed in support of the Yucca Mountain Repository License Application (LA), this paper explores the role of natural barriers, engineered barriers, and waste form characteristics relative to reducing the potential for postclosure criticality. Several parameters important to postclosure criticality are considered in order to determine which are useful discriminators to support packaging decisions. Using these parameters, a qualitative assessment is performed relative to three options for managing DOE-owned SNFs. (authors)

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THE SUCCESSFUL UTILIZATION OF COMMERCIAL TREATMENT CAPABILITIES TO DISPOSITION HANFORD NO-PATH-FORWARD SUSPECT TRANSURANIC WASTES

The U.S. Department of Energy (DOE) Richland Operations Office (RL) has adopted the 2015 Vision for Cleanup of the Hanford Site. The CH2M HILL Plateau Remediation Company's (CHPRC) Waste and Fuels Management Project (W&FMP) and their partners support this mission by providing centralized waste management services for the Hanford Site waste generating organizations. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 cubic meters (m{sup 3}) of legacy waste was defined as ''no-path-forward waste.'' A significant portion of this waste (7,650 m{sup 3}) comprised wastes with up to 50 grams of special nuclear materials (SNM) in oversized packages recovered during retrieval operations and large glove boxes removed from Hanford's Plutonium Finishing Plant (PFP). Through a collaborative effort between the DOE, CHPRC, and Perma-Fix Environmental Services, Inc. (PESI), pathways for these problematic wastes were developed and are currently being implemented.

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Development of the 3 m{sup 3} Low Level Radioactive Waste Container in Taiwan - 20037

Chinshan Nuclear Power Plant stopped its business operation in July, 2019, and Kuosheng Nuclear Power Plant will also be permanently shut down in 2023. Nuclear Reactor Facilities Regulation Act requires a nuclear facility must be dismantled in 25 years after being permanently ceased operation. According to the approved decommissioning plan of Chinshan Nuclear Power Plant and the proposed one of Kuosheng Nuclear Power Plant, the 3 m{sup 3} low level radioactive waste container is planned to store the low level radioactive wastes with higher radiation produced during the decommissioning. However, 3 m{sup 3} low level radioactive waste containers are not used or developed in Taiwan before. To meet such need, Institute of Nuclear Energy Research develops a new type of 3 m{sup 3} low level radioactive waste container and performs the corresponding analyses and tests based on Guidelines of Applying the Usage License of Low Level Radioactive Waste Containers and Regulations for the Safe Transport of Radioactive Material. In this article, it provides the information of the low level radioactive waste containers that have been approved or is under reviewing in Taiwan, and so do the design specifications of the developed 3 m{sup 3} waste container. Besides, the numerical evaluation and test results of the developed 3 m{sup 3} waste container are also included in this article. Based on the evaluation and test results, the developed 3 m{sup 3} waste container satisfies the requirements of the storage container and industrial package type 2. Institute of Nuclear Energy Research will submit the usage license applications to the authority this year, and the usage licenses would be granted in 2021, expectedly. (authors)

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Waste Management Analyses for Spent Structures at SNS - Paper 106

The Spallation Neutron Source (SNS) in Oak Ridge, Tennessee, is an accelerator driven neutron scattering facility for materials research. Presently, the SNS operates at 1.2 megawatt (MW) proton beam power incident on a mercury target with a proton beam energy of 1 GeV and 60 Hz repetition rate. The SNS consists of an accelerator facility, target facility, and a world-class suite of neutron scattering instruments supporting material, life-science and fundamental physics research. According to the SNS operations plan, some components are replaced when they reach their end-of-life due to radiation induced material damage or burn-up, or because of mechanical failure. During operation, these components, especially those in proximity to the target, are exposed to a radiation environment and build up significant activity during their service lifetime. All these components must be safely removed, placed in a container/package for storage, and ultimately transported offsite for disposal. An accurate estimate of the radionuclide inventory is performed assuming realistic irradiation history and decay time in order to classify a spent component. On the basis of a calculated radionuclide inventory, the spent component is classified, and an appropriate cask/package for transport and storage is suggested. After the choice for the container is made, radiation transport calculations are performed for the facility component placed inside the container, to ensure that the container is compliant with waste management regulations. When necessary, additional shielding is added. For each spent component before it goes out of the SNS site, supporting documentation is produced with a radionuclide inventory prediction, dose rate analyses and, if required, additional shielding. A dose rate analyses for the exposure prediction of personnel during components change out are performed as well. Most of the effort is concentrated on the target facility components such as target, proton beam window (PBW) and core vessel insert (CVI) plugs because of severe radiation conditions. Additionally, analyses are performed for the accelerator facility components, which are removed from the facility. Activation analyses and residual dose rates analyses are performed to support safe handling of SNS spent components at the SNS site and during transport to the waste disposal site. On the basis of the calculated radionuclide inventories, spent components characterized and classified with respect to waste management guidelines, and activity-based determinations are made regarding appropriate packaging for transport. (authors)

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Steam Dryer Segmentation and Packaging at Grand Gulf Nuclear Station - 13577

Entergy recently performed an Extended Power Up-rate (EPU) on their Grand Gulf Nuclear Station, near Port Gibson, Mississippi. To support the EPU, a new Steam Dryer Assembly was installed during the last refueling outage. Due to limited access into the containment, the large Replacement Steam Dryer (RSD) had to be brought into the containment in pieces and then final assembly was completed on the refueling floor before installation into the reactor. Likewise, the highly contaminated Original Steam Dryer (OSD) had to be segmented into manageable sections, loaded into specially designed shielded containers, and rigged out of containment where they will be safely stored until final disposal is accomplished at an acceptable waste repository. Westinghouse Nuclear Services was contracted by Entergy to segment, package and remove the OSD from containment. This work was performed on critical path during the most recent refueling outage. The segmentation was performed underwater to minimize radiation exposure to the workers. Special hydraulic saws were developed for the cutting operations based on Westinghouse designs previously used in Sweden to segment ABB Reactor Internals. The mechanical cutting method was selected because of its proven reliability and the minimal cutting debris that is generated by the process. Maintaining stability of the large OSD sections during cutting was accomplished using a custom built support stand that was installed into the Moisture Separator Pool after the Moisture Separator was installed back in the reactor vessel. The OSD was then moved from the Steam Dryer Pool to the Moisture Separator Pool for segmentation. This scenario resolved the logistical challenge of having two steam dryers and a moisture separator in containment simultaneously. A water filtration/vacuum unit was supplied to maintain water clarity during the cutting and handling operations and to collect the cutting chips. (authors)

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Volatilisation and oxidation of aluminium scraps fed into incineration furnaces

Highlights: Black-Right-Pointing-Pointer Aluminium packaging partitioning in MSW incineration residues is evaluated. Black-Right-Pointing-Pointer The amount of aluminium packaging recoverable from the bottom ashes is evaluated. Black-Right-Pointing-Pointer Aluminium packaging oxidation rate in the residues of MSW incineration is evaluated. Black-Right-Pointing-Pointer 80% of aluminium cans, 51% of trays and 27% of foils can be recovered from bottom ashes. - Abstract: Ferrous and non-ferrous metal scraps are increasingly recovered from municipal solid waste incineration bottom ash and used in the production of secondary steel and aluminium. However, during the incineration process, metal scraps contained in the waste undergo volatilisation and oxidation processes, which determine a loss of their recoverable mass. The present paper evaluates the behaviour of different types of aluminium packaging materials in a full-scale waste to energy plant during standard operation. Their partitioning and oxidation level in the residues of the incineration process are evaluated, together with the amount of potentially recoverable aluminium. About 80% of post-consumer cans, 51% of trays and 27% of foils can be recovered through an advanced treatment of bottom ash combined with a melting process in the saline furnace for the production of secondary aluminium. The residual amount of aluminium concentrates in the fly ash or in the fine fraction of the bottom ash and its recovery is virtually impossible using the current eddy current separation technology. The average oxidation levels of the aluminium in the residues of the incineration process is equal to 9.2% for cans, 17.4% for trays and 58.8% for foils. The differences between the tested packaging materials are related to their thickness, mechanical strength and to the alloy.

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A Critical Review of Challenges Faced by Converting Food Waste to Bioenergy Through Anaerobic Digestion and Hydrothermal Liquefaction

Here, the conventional approaches for handling food waste has been incineration, composting, landfilling, and anaerobic digestion for producing biogas. In light of organic waste bans and newly discovered presence of per- and polyfluorinated substances (PFAS) in food, food packaging materials, and compost, this review provides a critical summary of what has been investigated and reported and what needs to be considered when choosing suitable pathways for food waste. In addition to the fundamental principles inherent to anaerobic digestion and hydrothermal liquefaction, challenges for each process are identified followed by discussion of potential solutions to resolve the bottlenecks.

09 BIOMASS FUELS↗

Standard Operating Procedure (SOP) Repair Of Cracks And Spalls Within The Concrete Facade Covering The Runit Island Waste Containment Structure

The concrete repair shall be performed using a combination of pre-packaged cement-based mortar (Structural Concrete® V/O) supplied by Five Star Products, Inc., and CRACKBOND® products from Adhesive Technology Corp., Pompano Beach, Florida. Five Star Structural Concrete® V/O is a high strength, rapid setting, one component, permanent concrete repair material used primarily for vertical and overhead structural repairs. The material produces a repair which is dimensionally stable, develops an integral bond to existing concrete, and restores structural integrity within hours of placement (Five Star DESIGN-A-SPEC™ Guidelines). The CRACKBOND® family of products are two component, high strength structural epoxy formulations specifically designed to repair cracks in concrete.

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Biological Recycling of Polyethylene Terephthalate: A Mini-Review

Plastics have become an inseparable part of our life in the society due to their merits such as low cost, light weight, good mechanical and thermal properties along with easy process ability. Since the last few decades, the unrestrained use of plastics in various fields like agriculture, packaging, transportation has raised a serious issue about its disposal and pollution. An efficient decomposition of plastics takes about 1000 years. Not only there is an increase in the problem of waste disposal of plastics but also they release CO{sub 2} and dioxins on burning which contributes to cause pollution. There is a growing concern for use of efficient microorganisms meant for biodegradation of non-degradable synthetic polymer because conventionally used methods for plastic disposal are ineffective for plastic waste management. This review paper focuses on one such type of plastic-polyethylene terephthalate (PET) which is widely used in a flexible packaging application, its manufacturing process and the various effects of waste generated by it on the surrounding environment. This review discusses an overview of various existing plastic disposal methods along with their limitations and mechanism of biodegradation of PET and factors affecting biodegradation.

54 ENVIRONMENTAL SCIENCES↗

Characterization Techniques for TRU Waste from the Analytical Laboratory Hot Cells at Idaho National Laboratory - 19610

Since the first reactor went hot in 1942, there has been a never ending desire to improve the nuclear reactor and the fuel that it burns. The Idaho National Laboratory has been a key player in reactor and fuel development. As more research projects begin, the spectrum of waste parameters grows. Over the years, the Analytical laboratory (AL) at the INL Materials and Fuels Complex (MFC) has supported a large range of various research projects. Each project brings its own challenge and analytical technique that, in turn, yields its own unique waste parameters. The AL Hot Cells at INL consist of 6 small cells that can support around 12 different fuel research projects at a time along with infrastructure and environmental samples. Each different project yields different levels of transuranic (TRU) material, RCRA constituents, dose, and contamination levels. The waste parameters from each project dictate a path forward for the waste as Low-Level Waste (LLW), Contact Handled TRU (CH-TRU), Mixed Contact Handled TRU (M-CH-TRU), Remote Handled TRU (RH-TRU), or Mixed Remote Handled TRU (M-RHTRU). With visual examination techniques and process knowledge as an aid, there have been waste handling methods developed that allow debris and sample material to be segregated. Unique characterization mechanisms have been applied for each debris and sample material that are generally more accurate than smear and surface contamination data. For debris found in the Hot Cells, analytical data is gathered and conservatively summarized for each fuel type and research method. This consists of data from Quadrupole Inductively Coupled Plasma Mass Spectrometry (Q-ICP-MS), Gamma Spectroscopy, and some data verification through Thermal Ionization Mass Spectrometry (TIMS) and Multi-Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS). This data gives a range of TRU isotope concentrations. Fuel age is taken into account to determine the dominant TRU activity isotopes. Cesium concentrations are also gathered. Each parameter/fuel type then ties a measured Cesium number to TRU content based on historical analytical data. Debris in the hot cells is segregated by project and packaged in individual 2 gallon metal paint cans. Each can is counted by gamma spectroscopy in a calculated and measured geometry. Using process knowledge to determine the fuel or project type, newly counted cesium numbers can be applied to the historic analytical data and TRU content can be estimated. The TRU content allows waste debris to be segregated between CH-TRU and RH-TRU. Packaging techniques can allow for the minimization of RHTRU waste. Sample material received into the AL must be dissolved into a liquid matrix in order to be analyzed in most cases. When analysis is complete, these liquids go through a consolidation, complexation, neutralization, and solidification process where real time analytical data can be used to calculate TRU content and RCRA constitutes. These calculations allow for no further analytical data to be gathered. The waste matrix is considered stable and acceptable by all Acceptable Knowledge (AK) documents. The unique techniques used in characterizing and handling the Analytical Laboratories Hot Cell waste can be applied to various situations and various laboratories. They allow for more accurate data for waste constituents and also for the minimization of RH-TRU waste. (authors)

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Polymer grafted aramid nanofiber reinforces immiscible waste polypropylene/poly(ethylene terephthalate)

Polypropylene (PP) and poly(ethylene terephthalate) (PET) are plastics commonly used for packaging because of their excellent barrier and mechanical properties. The properties of these plastics are often diminished after mechanical recycling, inevitably causing down-cycling. Furthermore, this problem is exacerbated when different kinds of polymers mix. Aramid nanofibers have the potential to improve the mechanical properties of polymers due to their excellent mechanical properties but their poor dispersion in polymers is a challenge. Grafting polymers onto nanofibers can help address this challenge. In this work, different loading levels (1%, 2%, and 5%) of polymer grafted aramid nanofibers (ANFs) are blended with waste PP/PET (90/10), simulating a PP waste stream containing traces of PET contaminants. Scanning electronic microscopy, rheology, and differential scanning calorimetry results show the affinity of PP functionalized aramid nanofibers (PP_ANF) towards the PP matrix. At 1 wt% of the nanofiber, the size of the PET droplets in the PP matrix of the PP_ANF blend range from 0.2 to 2.0 μm while that of unmodified ANF and PET_ANF blends are in the range of 0.1–6.2 and 0.5–7.4 μm, respectively. In summary, polymer grafted ANFs have the tendency of improving properties of its like polymers due to similarity in the grafting polymer and the polymer matrix.

36 MATERIALS SCIENCE↗

CBFO Approved Filter Vents (Rev. 2)

The purpose of this procedure is to define the process for controlling the list of filter vents that are approved by the U.S. Department of Energy (DOE) Carlsbad Field Office (CBFO) for the packaging, characterization, shipment, and disposal of payload containers at the Waste Isolation Pilot Plant (WIPP). This procedure delineates requirements for the review, preparation, approval, and implementation of revisions to the list of approved filter vents.

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Cf-252 Characterization Documents

Six documents were written by Vance and Associates under contract to the Off-Site Source Recovery Project of Los Alamos National Laboratory. These Six documents provided the basis for characterization of Californium-252 sealed sources and for the packaging and manifesting of this material for disposal at the Waste Isolation Pilot Project. The Six documents are: 1. VA-OSR-10, Development of radionuclide distributions for Cf-252 sealed sources. 2. VA-OSR-11, Uncertainty analysis for Cf-252 sealed sources. 3. VA-OSR-12, To determine the radionuclides in the waste drums containing Cf-252 sealed source waste that are required to be reported under the requirements of the WIPP WAC and the TRAMPAC. 4. VA-OSR-13, Development of the spreadsheet for the radiological calculations for the characterization of Cf-252 sources. 5. VA-OSR-14, Relative importance of neutron-induced fission in Cf-252 sources. 6. VA-OSR-15, Determine upper bound of decay product inventories from a drum of Cf-252 sources. These six documents provide the technical basis for the characterization of Cf-252 sources and will be part of the AK documentation required for submittal to the Central Characterization Project (CCP) of WIPP.

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U.S. Department of Energy/Argonne National Laboratory Training Courses for Packaging, Transportation, and Storage of Radioactive Material - 17269

The U.S. Department of Energy (DOE) Packaging Certification Program (PCP), Office of Packaging and Transportation, Office of Environmental Management, has sponsored a suite of training courses that have been conducted since the early 1990s. Of these training courses, Argonne National Laboratory (Argonne) conducts three courses: (1) Quality Assurance (QA) for Radioactive Material Packaging, (2) American Society for Mechanical Engineers (ASME) Pressure Vessel Code for Nuclear Transport and Storage, and (3) Nuclear and Other Radioactive Materials Transport Security. Argonne has been working with the University of Nevada, Reno, to develop an accredited graduate-level program in nuclear packaging. The Graduate Certificate in Nuclear Packaging (GCNP) program provides an applied knowledge-based curriculum in packaging for nuclear and radioactive materials that complements other graduate programs at the University (for additional information, see https://www.unr.edu/degrees/nuclear-packaging/certificate). (authors)

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Remote-Handled TRU Storage Challenges at Sandia National Laboratories/ New Mexico - 16071

The events at the Waste Isolation Pilot Plant (WIPP) in February 2014 had a significant impact on the packaging, storage, and disposal options at Sandia National Laboratories/New Mexico (Sandia) as with other transuranic generator sites. In 2012, Sandia had identified accountable material for disposal, fresh and irradiated mixed oxide (MOX) and uranium oxide (UO{sub 2}), which had been used in experiments conducted in the 1970's and 1980's. During the 2012/2013 timeframe, discussions with WIPP determined the samples and experimental vessels qualified as remote-handled (RH) transuranic (TRU) waste for disposal at WIPP. Sandia began discussions with WIPP to determine the process to repackage and characterize the material for disposal at WIPP in addition to obtain funding to complete the project. By January 2014, Sandia had contracted with Nuclear Waste Partnerships, LLC (NWP) to provide Central Characterization Program (CCP) certified visual examination (VE) and dose-to-curie (DTC) support to package and certify the accountable material/waste. Work began on February 3, 2014 with the arrival of two VE operators. On February 5, 2014, a salt truck caught fire in the underground and on February 14, 2014, the radiological release occurred at WIPP initiating a suspension of shipments. The length of the suspension was not known at the time. Sandia was only a couple of days into this long-term project, had signed a contract with NWP for support, and was still responsible for meeting earlier negotiated milestones. There were 37 parent containers which included experimental vessels from 3 to 10 feet long and unused material pieces. The original plan was to package each of the 37 parent containers separately, but several issues, including the suspension of shipments, made it imperative that a new path forward be developed. Because space is limited in the Auxiliary Hot Cell Facility (AHCF), the storage configuration for criticality safety and material-at-risk was of the highest importance. Therefore, packaging in the fewest number of containers without exceeding WIPP limits, keeping the accountable items together, and not violating any safety basis requirements was Sandia's goal. (authors)

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