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Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents AMP-PAN and CST R91290-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions, but CST should be investigated at a higher pH. Limited removal of transuranic species and almost no removal of uranium was observed in this testing. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in the range of 0.2 to 5 mg Cs per gram of absorbent appears to be achievable.

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Regulatory framework for NORM residues in Belgium

The Belgian radiation protection authority (Federal Agency for Nuclear Control - FANC) has published in March 2013 a decree regulating the acceptance of NORM residues by nonradioactive waste treatment facilities. This regulation is based on the concept of 'work activities involving natural radiation sources' in the sense of article 40 of the 96/29/EURATOM directive. The disposal or processing facilities which accept NORM residues with an activity concentration above a generic exemption level will be considered as 'work activities' and submitted to declaration according to the Belgian radiation protection regulations. On basis of this declaration, specific acceptance criteria for the different types of processing/ disposal of the residues (disposal on landfill, recycling into building materials, etc.) are imposed. FANC has drafted guidelines for these acceptance criteria. A methodological guide for the operators of the concerned facilities was also published. Moreover, sites where significant quantities of NORM residues are or have been disposed, are subjected to an environmental monitoring in the framework of the national program of radiological surveillance of FANC. FANC also introduced in its regulations the concept of anthropogenic radon-prone areas: e.g. former phosphogypsum stacks have been defined as anthropogenic radon-prone areas, which allows some form of regulatory control of these sites. (authors)

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Tank 40 Final SB7b Chemical Characterization Results

A sample of Sludge Batch 7b (SB7b) was taken from Tank 40 in order to obtain radionuclide inventory analyses necessary for compliance with the Waste Acceptance Product Specifications (WAPS) i . The SB7b WAPS sample was also analyzed for chemical composition including noble metals and fissile constituents, and these results are reported here. These analyses along with the WAPS radionuclide analyses will help define the composition of the sludge in Tank 40 that is currently being fed to the Defense Waste Processing Facility (DWPF) as SB7b.

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Conducting Gamma Radiation Surveys on Municipal Roadways in Support of the Port Hope Area Initiative - 16602

The Port Hope Area Initiative (PHAI) is a community-based solution for the long-term management of historic low level radioactive waste (LLRW) resulting from 60 years of uranium and radium processing operations in the Town of Port Hope, Ontario. The project aims to remediate approximately 1.7 million cubic metres LLRW through the consolidation of waste within two Long-term Waste Management facilities to be constructed within the municipalities of Port Hope and Clarington. Within the Town of Port Hope, surveys are being conducted at approximately 5000 individual properties to confirm the presence of LLRW, delineate its extent and develop cost estimates for remediation. Identification of the presence of LLRW is based on multiple lines of evidence including reviews of historical information, gamma radiation surveys and sampling for four signature contaminants. Of the 5000 properties, a number are municipal rights of way, largely municipal roads, that may have been impacted through spillage of LLRW when it was historically transported through the community for disposal and/or through its use as construction backfill. This paper describes the completion of gamma radiation surveys on these municipal rights of way and the challenges that needed to be addressed working on travelled roadways in an active community. To maximize efficiency and minimize inconvenience to motorists the gamma surveys are conducted using a mobile platform that provides high resolution spectral gamma logging coupled to GPS for geo-spatial referencing. The system employs large volume, spectroscopy grade, sodium iodide detectors that permit the collection radiological spectra allowing for effective discrimination of gamma radiation signals emitted from naturally occurring background sources (e.g. potassium 40) from the gamma emissions associated with the specific contaminants of concern (e.g. Ra-226). The system can be deployed on any platform including a backpack. For the majority of the roadways surveyed a small, self-contained, road-legal trailer attached to a motorized tow vehicle (e.g. side-by-side UTV or small pick-up truck) is used as the survey platform. This accommodates the mounting of multiple detectors and allows the surveys to be conducted without the need for closing traffic lanes thus improving productivity and safety while minimizing inconvenience to motorists. Data quality objectives are readily achieved through the use of multiple passes over the same target area providing a geo-spatially rich dataset for post-processing and imaging. Topics covered in the paper include: the logistics of completing high quality gamma surveys on travelled roadways; health, safety and traffic considerations; community outreach and response; and a comparison of isotopically-speciated data vs total gamma readings evaluating the influence of background sources. The lessons learned will be applicable to anyone planning on conducting similar surveys in urban settings. (authors)

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Three-dimensional mapping of crystalline ceramic waste form materials

Here, we demonstrate the use of synchrotron-based, transmission X-ray microscopy (TXM) and scanning electron microscopy to image the 3-D morphologies and spatial distributions of Ga-doped phases within model, single- and two-phase waste form material systems. Gallium doping levels consistent with those commonly used for nuclear waste immobilization (e.g., Ba 1.04 Cs 0.24 Ga 2.32 Ti 5.68 O 16 ) could be readily imaged. This analysis suggests that a minority phase with different stoichiometry/composition from the primary hollandite phase can be formed by the solid-state ceramic processing route with varying morphology (globular vs. cylindrical) as a function of Cs content. Our results represent a crucial step in developing the tools necessary to gain an improved understanding of the microstructural and chemical properties of waste form materials that influence their resistance to aqueous corrosion. This understanding will aid in the future design of higher durability waste form materials.

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Status Update on the NCRP Scientific Committee SC 5-1 Report: Decision Making for Late-Phase Recovery from Nuclear or Radiological Incidents - 13450

In August 2008, the U.S. Department of Homeland Security (DHS) issued its final Protective Action Guide (PAG) for radiological dispersal device (RDD) and improvised nuclear device (IND) incidents. This document specifies protective actions for public health during the early and intermediate phases and cleanup guidance for the late phase of RDD or IND incidents, and it discusses approaches to implementing the necessary actions. However, while the PAG provides specific guidance for the early and intermediate phases, it prescribes no equivalent guidance for the late-phase cleanup actions. Instead, the PAG offers a general description of a complex process using a site-specific optimization approach. This approach does not predetermine cleanup levels but approaches the problem from the factors that would bear on the final agreed-on cleanup levels. Based on this approach, the decision-making process involves multifaceted considerations including public health, the environment, and the economy, as well as socio-political factors. In an effort to fully define the process and approach to be used in optimizing late-phase recovery and site restoration following an RDD or IND incident, DHS has tasked the NCRP with preparing a comprehensive report addressing all aspects of the optimization process. Preparation of the NCRP report is a three-year (2010-2013) project assigned to a scientific committee, the Scientific Committee (SC) 5-1; the report was initially titled, Approach to Optimizing Decision Making for Late- Phase Recovery from Nuclear or Radiological Terrorism Incidents. Members of SC 5-1 represent a broad range of expertise, including homeland security, health physics, risk and decision analysis, economics, environmental remediation and radioactive waste management, and communication. In the wake of the Fukushima nuclear accident of 2011, and guided by a recent process led by the White House through a Principal Level Exercise (PLE), the optimization approach has since been expanded to include off-site contamination from major nuclear power plant accidents as well as other nuclear or radiological incidents. The expanded application under the current guidance has thus led to a broadened scope of the report, which is reflected in its new title, Decision Making for Late-Phase Recovery from Nuclear or Radiological Incidents. The NCRP report, which is due for publication in 2013, will substantiate the current DHS guidance by clarifying and elaborating on the processes required for the development and implementation of procedures for optimizing decision making for late-phase recovery, enabling the establishment of cleanup goals on a site-specific basis. The report will contain a series of topics addressing important issues related to the long-term recovery from nuclear or radiological incidents. Special topics relevant to supporting the optimization of the decision-making process will include cost-benefit analysis, radioactive waste management, risk communication, stakeholder interaction, risk assessment, and decontamination approaches and techniques. The committee also evaluated past nuclear and radiological incidents for their relevance to the report, including the emerging issues associated with the Fukushima nuclear accident. Thus, due to the commonality of the late-phase issues (such as the potential widespread contamination following an event), the majority of the information pertaining to the response in the late-phase decision-making period, including site-specific optimization framework and approach, could be used or adapted for use in case of similar situations that are not due to terrorism, such as those that would be caused by major nuclear facility accidents or radiological incidents. To ensure that the report and the NCRP recommendations are current and relevant to the effective implementation of federal guidance, SC 5-1 has actively coordinated with the agencies of interest and other relevant stakeholders throughout the duration of the project. The resulting report will be an important resource to guide those involved in late-phase recovery efforts following a nuclear or radiological incident. (authors)

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Options for the Removal of Cesium and Strontium from Acidic Uranyl Sulfate Solutions from Molybdenum-99 Production: Part II The Effect of Sodium Carbonate Adjustment to pH 9 on Strontium Removal

A small laboratory-scale column process aimed at removing fission products from prototypical acidic sulfate solution wastes from a prototypical Mo-99 recovery process has been operated in a radioactive materials hood at Savannah River National Laboratory. This equipment has been operated with realistic concentrations of uranium (0.4 wt % 235 U), plutonium ( 239 Pu), neptunium ( 237 Np), non-radioactive surrogates of common fission products (including Cs and Sr) and two promising absorbents – AMP-PAN and CST R9120-B. The goal of this work was to demonstrate the removal and concentration of high specific activity isotopes which are expected to dominate the classification of the low-level waste from the production of 99 Mo. Both AMP-PAN and CST R9120-B are attractive absorbents for the removal of 137 Cs from an acidic sulfate waste stream. They do not provide a solution for the removal of 90 Sr from the acidic solutions. CST was shown to be useful at removing both Cs and Sr after adjustment of the acidic sulfate solution with sodium carbonate to pH 9. Limited removal of transuranic species and almost no removal of uranium was observed for CST R9120B absorbent. This could be important due to the relatively large amount of uranium expected to be present in such waste streams. The capacity of these absorbents for 137 Cs will likely be dependent on the Cs concentration in the solutions processed but loading in excess of 0.7 mg Cs per gram of CST absorbent while loading 0.14 mg Sr per gram of CST absorbent appears to be achievable from pH 9 carbonate solution.

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Remote Dismantling and Packaging of the RPV and Thermal Shield at the Obrigheim NPP - 16073

The Obrigheim Nuclear Power Plant (KWO) is located in the Federal State of Baden- Wuerttemberg in Germany and was operated by the EnKK-KWO. The Nuclear Power Plant - a light water pressurized water reactor with 357 MW of electrical power - was operated from 1968 to 2005. In June 2009 the Energiewerke Nord GmbH (EWN) was charged with the dismantling planning and dismantling of the reactor. According to the German nuclear and radiation protection law, several licenses will be required for the realisation of this task. This also applies to the disposal of the waste produced by dismantling, which must finally be stored in the German final repository for low and medium-radioactive waste (underground repository 'Konrad'). In spring 2010 EWN completed all application documents for the dismantling of the reactor pressure vessel (RPV) and the RPV internals and EnKK-KWO delivered them to the responsible Environmental Ministry of Baden-Wuerttemberg. Since then it has been worked intensively on the design, construction and manufacturing of the facilities for dismantling. With a separate notification of amendment, it was applied for the license to install and commission the facilities for the dismantling of the RPV and the RPV internals at the Environmental Ministry of Baden-Wuerttemberg in October 2010. This notification of amendment was confirmed in summer 2012. In September 2011 the work process and test plan and the technical information about the campaign for the removal of the radioactive waste resulting from the RPV with internals, peripheral components and secondary waste with the aim of final disposal in the Federal Repository 'Konrad' was submitted to the Federal Office for Radiation Protection. The evaluation and assessment of the work process and test plan was concluded in August 2013, when it was audited by TUV Nord EnSys without any further measures and indications. On September 10 the work process and test plan was released by the Federal Office for radiation protection. The dismantling of the RPV internals upper core and lower core structure started in September 2013 in the wet cutting area in room 01.106/306 and were finished in October 2014. In the presentation 'Dismantling of the Obrigheim NPP Reactor and Waste Management' held on the WM Symposia in Phoenix, 2014, an overview about the remote cutting and packaging of the upper and lower core structure was given. In today's presentation the subsequent dismantling activities will be outlined and evaluated. The presentation will start with the short description of the basic frame conditions for the nuclear licensing procedure, important procedural principles, and necessary plant-specific conditions. After the remote cutting and packing of the reactor internals of the upper and lower core structure was completed, the wet cutting area 'reactor room 01.106/306' was cleaned and the reactor room rebuilt into a cutting area for peripheral components. After the peripheral components were dismantled, all preconditions were set to lift the RPV from its mounting position and to arrange it in the upper reactor room 01.306. After these activities were completed, the lifting of the RPV was prepared and the cutting area 'Dismantling of the RPV insulation in the reactor room 01.106/306' was set up. To carry out this task, a strand jack unit with connection block was installed on the bridge of the reactor building crane. The opening created by the dismantling of the pool plate was closed with a flap door. The manipulators used for the wet cutting of the RPV internals were also used for the remote dismantling of the RPV insulation and mounted in the cutting area. The remote facilities were handled from a control centre and/or control stand. The air-technical housing which has already been used for the remote dismantling of the RPV internals was designed for the dismantling of the RPV insulation and was extended for this dismantling scope with appropriate ventilation systems. A special feature is the fixation of the RPV insulation at the RPV corpus. The RPV insulation consists of insulation wool, wire mesh and metal-sheet insulation. This 'Sandwich Construction' is detached with a steel corset at the RPV corpus and partly tightened. The insulation of the RPV bottom dome is tightened at the supporting frame. Due to the activation of the material in the core area and the resulting high ambient dose rate the RPV insulation can only be dismantled remotely. Additionally, as asbestos fibers are expected to be found in the insulation the dismantling has to be performed according to the Technical Rules for hazardous substances. The demolition, restoration or maintenance works have to be performed remotely according to the Technical Rules for hazardous substances when the RPV is located in the cutting area. But in case of dismantling of a power reactor no such experience can be found. Beginning with the explanation of the technical, radiological and substantial boundary conditions this presentation will not only impart knowledge about the dismantling of the RPV insulation but also about the remote controlled packaging of the metal sheet insulation and the manual conditioning of the RPV insulation. Subsequently, reports on the remotely controlled cutting and packaging of the RDB are given. Here is particularly interesting that cutting methods have been used and how the activity determination of the segments was carried out. On the basis of the evaluation of the realisation the subsequent dismantling activities will be lined out. In parallel to the dismantling tasks, the waste package documentation for the final repository was issued, checked and confirmed for the containers loaded with parts of the RDB internals. The most important stages are exemplified in this presentation. (authors)

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Enhancing e-waste estimates: Improving data quality by multivariate Input–Output Analysis

Highlights: • A multivariate Input–Output Analysis method for e-waste estimates is proposed. • Applying multivariate analysis to consolidate data can enhance e-waste estimates. • We examine the influence of model selection and data quality on e-waste estimates. • Datasets of all e-waste related variables in a Dutch case study have been provided. • Accurate modeling of time-variant lifespan distributions is critical for estimate. - Abstract: Waste electrical and electronic equipment (or e-waste) is one of the fastest growing waste streams, which encompasses a wide and increasing spectrum of products. Accurate estimation of e-waste generation is difficult, mainly due to lack of high quality data referred to market and socio-economic dynamics. This paper addresses how to enhance e-waste estimates by providing techniques to increase data quality. An advanced, flexible and multivariate Input–Output Analysis (IOA) method is proposed. It links all three pillars in IOA (product sales, stock and lifespan profiles) to construct mathematical relationships between various data points. By applying this method, the data consolidation steps can generate more accurate time-series datasets from available data pool. This can consequently increase the reliability of e-waste estimates compared to the approach without data processing. A case study in the Netherlands is used to apply the advanced IOA model. As a result, for the first time ever, complete datasets of all three variables for estimating all types of e-waste have been obtained. The result of this study also demonstrates significant disparity between various estimation models, arising from the use of data under different conditions. It shows the importance of applying multivariate approach and multiple sources to improve data quality for modelling, specifically using appropriate time-varying lifespan parameters. Following the case study, a roadmap with a procedural guideline is provided to enhance e-waste estimation studies.

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Update on Full-Scale Retrieval and Transfer System Testing for Pneumatic Conveyance of High-Level Calcine Waste - 20241

The Idaho National Laboratory processed spent nuclear fuel from 1953 to 1994. The process used solvent extraction to separate uranium, resulting in a liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive solid powder, referred to as calcine. The calcine is stored in stainless steel bins contained within concrete vaults called the Calcined Solids Storage Facilities. There are a total of seven CSSFs. The Calcine Retrieval Project is tasked with retrieving the calcine from the bins of CSSF 1 and pneumatically transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. Once the calcine is removed from CSSF1, that facility may be closed, reducing the footprint of waste storage at the INL. The same process for removing Calcine from CSSF1 may then be utilized to remove the calcine from the remaining CSSFs for final treatment and transport to a long-term storage facility. To overcome these inherent challenges with accessing CSSF 1 and extracting the calcine in a timely manner, a full-scale integrated mockup has been constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. All essential elements of the transport design are represented in the Mockup, including motive and makeup air supply (air compressor), a replica of the CSSF6 cyclone, pre-filter, HEPA filter, vacuum pump, and retrieval control devices. Construction on the Mockup was completed in 2018 and testing began in February 2019. It has already proven very beneficial to the CRP. Experience in running the Mockup have aided in the development of retrieval and transport methodologies and the selection of process parameter values. It has also uncovered some unforeseen technological challenges and will continue to be valuable to the testing team as these hurdles are overcome and the process is proven effective. Operations personnel have also benefitted from its use as a training tool and will continue to use the Mockup as a simulator to reproduce problems encountered during extraction and transfer of Calcine from CSSF 1. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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An Assessment of Potential Dose Impacts from External Contamination on Naval Reactors Facility Waste Canisters (Special Analysis)

This Special Analysis (SA) was performed to address a request by Naval Reactors Facility (NRF) Waste Programs for a permanent exception to limits of removable surface contamination on the exterior of waste canisters shipped to the Remote-Handled Low-Level Waste (RHLLW) Disposal Facility as specified in the waste acceptance criteria (WAC) (PLN-5446). The purpose of this SA is to determine if the NRF-requested levels for removable surface contamination on the exterior of all NRF waste canisters is within the bounds of the current performance assessment (PA). This was done by calculating the groundwater all-pathways dose contribution from surface contamination on the exterior of NRF canisters for the following cases: (1) the exteriors of all NRF waste canisters are contaminated to the 10 CFR 835 Appendix D allowable limits in the current WAC, and (2) the exteriors of all NRF waste canisters are contaminated to the limits requested by NRF Waste Programs. Dose impacts for each case were compared to each other and to the all pathways dose for the PA base case. Dose impacts were also compared to the all pathways dose limit specified in DOE O 435.1. A simple assessment of the potential impacts of the increase in surface contamination using the NRF-requested limits on the biotic, air, and inadvertent intruder pathways was also performed. Based on the results of this SA, the increase in canister exterior contamination limits requested by NRF are well within the bounds of the current PA and will not result in a violation of performance objectives. The results also show the increased limits do not reflect or necessitate a fundamental change to the PA conceptual model, nor a change to the way exterior contamination is not included in PA dose calculations. Therefore, it is recommended the NRF request for an exception to the current limits for external surface contamination be accepted and the revised limits for NRF-generated waste canisters be added to the WAC. The monitoring plan will also be revised to identify external canister contamination as a potential mobile source term that may be detected by monitoring earlier than potential releases from waste. The PA, composite analysis (CA), closure plan, and PA/CA maintenance plan do not require revision. Recommendations are also included for inclusion of NRF procedures used to limit water pool radionuclide variability (and thus canister surface contamination variability) to the waste generator certification process.

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Hanford Site Composite Analysis: Radionuclide Selection for Groundwater Pathway Evaluation.

The updated Hanford Site Composite Analysis (CA) will provide an all-pathways dose projection to a hypothetical future member of the public from all planned low-level radioactive waste disposal facilities and potential contributions from all other projected end-state sources of radioactive material left at the Hanford Site following site closure. Its primary purpose is to support the decision-making process of the U.S. Department of Energy (DOE) under DOE O 435.1-1, Radioactive Waste Management, related to managing low-level waste disposal facilities at the Hanford Site. A key aspect of conducting a CA is selecting the radionuclides to be included in the quantitative analysis.

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Updates to a Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 20420

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The U.S. Department of Energy (DOE) was required to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980 including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the West Valley Site that evaluates potential future consequences for human health and the environment. This forms the core of the spatial domain of the West Valley PPA Model. The PPA Model, developed using the GoldSim system modeling software, is a tool intended to provide support for decision making that evaluates uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-grouped contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of high level radioactive waste resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The PPA Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife could be exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the PPA Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. PPA Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes. With this information, the decision makers can drive informed, defensible decisions regarding decommissioning of the Site. This paper includes an update of the information presented at WM2019 [1]. (authors)

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Remediation of Residential Properties Contaminated with LLRW as Part of the Port Hope Area Initiative - 17456

The Port Hope Area Initiative (PHAI) is a community-based solution for the long-term management of historic low level radioactive waste (LLRW) resulting from 60 years of uranium and radium processing operations in the urban setting of Port Hope, Ontario. The Eldorado refinery, on the shores of Lake Ontario, began refining Ra-226 from pitchblende ore, later transitioning to the refining of uranium ore. Within the community of Town of Port Hope, surveys are being conducted at approximately 5000 individual properties to identify the presence of LLRW resulting from fugitive emissions from the refinery and/or the use of refinery residue as construction backfill within the community. Of the 5000 properties being surveyed, it is expected that approximately 10% will contain some degree of LLRW requiring physical remediation. The design phase for remediation of the first group of 44 residential properties to be included in the small scale sites program and the various challenges associated with working in a well-established community are presented. In developing the design packages for individual properties it is important to understand that this remedial phase represent the final and most intrusive aspect of the PHAI for those homeowners affected. As such, the success of the program is dependent on how well the needs and desires of the community and individual property owners are incorporated into the remedial designs and how well those designs are implemented to remove LLRW and restore properties to their original (or better) condition. (authors)

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Radiological Verification for the Remediation of LLRW at the Port Granby Project - 19210

The Port Granby Project is part of the Port Hope Area Initiative (PHAI) that Canadian Nuclear Laboratory (CNL) is using to manage long-term historic low level radioactive waste (LLRW) resulting from 60 years of uranium and radium processing operations in the urban setting of Port Hope, Ontario. Port Granby remediation scope includes constructing a new, secure long-term waste management facility (LTWMF) and supporting infrastructure and excavating and relocating LLRW and marginally contaminated soil from the Port Granby Waste Management Facility (PGWMF) which accepted waste until the 1980's to the newly constructed LTWMF. The PGWMF contains an estimated 450,000 cubic meters of LLRW deposited in unmarked and unlined trenches throughout the southern portion of the site that borders Lake Ontario. Project requirements include excavating contaminated material, verifying that the trench bottoms are below action levels using Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) guidance, and final restoration. This paper describes challenges of meeting remediation verification requirements using traditional field and laboratory confirmation methods in areas of deep trench areas located in rugged terrain with ponded water and mud and presents a process that has expedited the traditional MARSSIM method for verification with greater productivity, and with enhanced safety and reliability as compared to the use of traditional methods alone. This enhanced process has been successfully employed for in-situ verification so that the excavated area can be backfilled and returned to a vegetative state. This enhanced process has been examined by federal regulatory staff through the course of their onsite audits and outcomes from these examinations have all been positive. The method involves a visual inspection and traditional walkover of the area to identify any areas with radiation measurements above the region's background followed by a full area survey performed using Wood's proprietary Orion ScanPlotSM system. The ScanPlotSM data is evaluated, mapped and overlaid on top of an aerial image of the surveyed area allowing remaining radiological material at surface or near surface to be quickly and easily identified for removal. Field staff then collect soil samples from evenly distributed points within the surveyed area and field screen the samples using an X-ray fluorescence (XRF) analysis for the three primary contaminates of potential concern (CoPCs) (arsenic, uranium and lead) that are indicative of the historical contamination. Samples passing XRF field screening are packaged and sent to a local laboratory for a full suite of analyses and checked against the CoPC list to assess whether the soil matches the profile for nearby native soils and passes the remediation goals. (authors)

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Performance Assessment for the E-Area Low-Level Radioactive Waste Disposal Facility at the Savannah River Site: Chapter 3

This chapter summarizes safety functions and features, events, and processes; conceptual models which apply to all DUs; property data packages; and modeling tools developed and implemented in this PA to analyze ELLWF performance. Conceptual models of GW flow and transport in the VZ, which are specific to type of DU, are presented in Chapter 4.

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Summary of Savannah River Site FY24 Salt Waste Qualification Data

Savannah River Mission Completion (SRMC), the Liquid Waste Operations subcontractor at SRS, and Savannah River National Laboratory (SRNL) analyzed samples from Savannah River Site (SRS) Waste Tanks 21H, 41H, and 42H to support qualification of Salt Waste Processing Facility (SWPF) Waste Batches 10, 11, 12, and 13 for processing (the FY24 Salt Batch Qualification samples). These Tanks (i.e. 21H, 41H and 42H) are blend tanks for feed to SWPF. This report focuses on the characterization of the Salt Batch Qualification sub-samples (i.e., any adjustments after qualification or as a part of processing are documented elsewhere). None of the samples displayed any unusual or unexpected characteristics such as large amounts of solids, floating solids, or unusual color. Characterization of these samples confirmed similar chemical composition and characteristics to previous salt waste batches. The analytical results (both rapid, typically 4 weeks, and long-term, typically 8 weeks) for Salt Batches 10, 11, 12, and 13 are now summarized and discussed in this technical report. SRMC-Analytical Laboratory (SRMC-AL), for the first time, provided all the short-term analyses data used in the qualification evaluations of SWPF Salt Batches 10, 11, 12, and 13. These SRMC short-term results are presented along with the results for the long-term analyses provided by SRNL.

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Modelling radionuclide transport in fractured media with a dynamic update of K d values

Radionuclide transport in fractured crystalline rocks is a process of interest in evaluating long term safety of potential disposal systems for radioactive wastes. Given their numerical efficiency and the absence of numerical dispersion, Lagrangian methods (e.g. particle tracking algorithms) are appealing approaches that are often used in safety assessment (SA) analyses. In these approaches, many complex geochemical retention processes are typically lumped into a single parameter: the distribution coefficient (Kd). Usually, the distribution coefficient is assumed to be constant over the time frame of interest. However, this assumption could be critical under long-term geochemical changes as it is demonstrated that the distribution coefficient depends on the background chemical conditions (e.g. pH, Eh, and major chemistry). In this study, we provide a computational framework that combines the efficiency of Lagrangian methods with a sound and explicit description of the geochemical changes of the site and their influence on the radionuclide retention properties.

58 GEOSCIENCES↗