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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Innovation in Contracting Strategy for a Master Services Agreement for Remediation of Low Level Radioactive Waste - 18381

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. It is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from several locations, including the Port Hope Harbour, a closed municipal landfill and approximately 800 residential properties. This complex project requires contracting the services of qualified, skilled and experienced environmental remediation contractors to meet the diverse needs of the myriad tasks that comprise the remediation and restoration of all properties. This paper will highlight how the CNL contracting services team developed an innovative approach to a contracting strategy designed to deliver the safe execution of work and keep the project on time and within budget. The approach addresses the complexities of a scope of work that consists of multiple work packages to be executed both concurrently and in tandem, and that require highly specialized trades/skills, as well as high-level project management expertise to ensure seamless coordination among the various contractors. The author will discuss CNL's strategy to engage the market with a Request for Expression of Interest to identify potential qualified suppliers and gauge their level of interest, issue a Request for Prequalification (RFPQ) to determine eligible contractors and, ultimately, issue a Request for Proposal for a Master Services Agreement (MSA) that would competitively evaluate each contractor's direct experience, technical expertise, quality management programs, safety systems compliance and track record. Using CNL's experience as a case study, the paper will highlight how an innovative approach can positively affect project timelines by describing how CNL addressed the challenge associated with the limited time available to implement procurement processes and deliver a series of executed agreements while ensuring that work would begin on time and key project milestones would be met. The author will summarize how CNL's strategy put the project back on schedule, without compromising the process guideline/requirements, and ensured that a qualified and experienced set of contractors, committed to a strong safety culture, will be available to perform the work required when the cleanup in the Port Hope community begins in 2018. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Removal of Historic Low-Level Radioactive Waste from the Highland Drive South and Alexander Street Ravines - 19155

In Port Hope, the Highland Drive South Ravine (HDSR) is the discharge area for groundwater originating from the closed Highland Drive Landfill site. The presence of low-level radioactive waste (LLRW) in the HDSR is the result of a contaminated groundwater plume flowing from the landfill site. The contaminant plume contains elevated concentrations of uranium and arsenic and extends from the landfill south to discharge groundwater to shallow soils in a wet discharge area within the ravine. The groundwater discharges to Hunt's Ponds and Highland Drive South Creek, which are immediately downstream of the wet discharge area and will be included in the remediation design. Remediation and environmental management plans for HDSR have been developed within the framework of the Port Hope Project. Recent subsurface investigations have identified LLRW at greater depths than previously anticipated. Other investigations of surrounding properties have identified LLRW that is primarily from filling activities in backyards. The steep topography of the site and the presence of the creek and seepage face present remediation challenges. Also in Port Hope, the LLRW in the Alexander Street Ravine Area is the result of end dumping, filling of private backyards and stack deposition from the former refinery. The original site extent was focused on the deeper contamination due to end dumping and included a portion of Alexander Creek. However, recent investigations have confirmed other fill areas and the extent of stack deposition that are generally contiguous with the end dumping area. Restricted access to the site will result in all remediation in this area taking place via a single access point along the narrow existing Hayward Street. While both of these sites are ravines in Port Hope containing LLRW, the origins of contamination and remediation approaches are very different. This paper will compare these two challenging remediation sites. (authors)

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DOE OSTI2014

The Performance of Underground Radioactive Waste Storage Tanks at the Savannah River Site: A 60-Year Historical Perspective

The Savannah River Site produced weapons-grade materials for nearly 35 years between 1953 and 1988. The legacy of this production is nearly 37 million gallons of radioactive waste. Since the 1950s, the liquid waste has been stored in large, underground carbon steel waste tanks. During the past 20 years, the site has begun to process the waste so that it may be stored in vitrified and grout forms, which are more suitable for long-term storage. Over the history of the site, some tanks have experienced leakage of the waste to the secondary containment. This article is a review of the instances of leakage and corrosion degradation that the tanks and associated equipment have experienced since the first tanks were built. Furthermore, the activities that the site has taken to mitigate the degradation and manage the service life of the tank for its anticipated lifetime are reviewed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Quantitative comparison between PGNAA measurements and MCNP calculations in view of the characterization of radioactive wastes in Germany and France

The Forschungszentrum Juelich GmbH (FZJ), together with the Aachen University Rheinisch-Westfaelische Technische Hochschule (RWTH) and the French Alternative Energies and Atomic Energy Commission (CEA Cadarache) are involved in a cooperation aiming at characterizing toxic and reactive elements in radioactive waste packages by means of Prompt Gamma Neutron Activation Analysis (PGNAA). The French and German waste management agencies have indeed defined acceptability limits concerning these elements in view of their projected geological repositories. A first measurement campaign was performed in the new Prompt Gamma Neutron Activation Analysis (PGNAA) facility called MEDINA, at FZJ, to assess the capture gamma-ray signatures of some elements of interest in large samples up to waste drums with a volume of 200 liter. MEDINA is the acronym for Multi Element Detection based on Instrumental Neutron Activation. This paper presents MCNP calculations of the MEDINA facility and quantitative comparison between measurement and simulation. Passive gamma-ray spectra acquired with a high purity germanium detector and calibration sources are used to qualify the numerical model of the crystal. Active PGNAA spectra of a sodium chloride sample measured with MEDINA then allow for qualifying the global numerical model of the measurement cell. Chlorine indeed constitutes a usual reference with reliable capture gamma-ray production data. The goal is to characterize the entire simulation protocol (geometrical model, nuclear data, and postprocessing tools) which will be used for current measurement interpretation, extrapolation of the performances to other types of waste packages or other applications, as well as for the study of future PGNAA facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Safe Packaging of Chemically Reactive Radioactive Waste - Addressing the Data Needs - 17487

Industry-standard thermal hazard screening is an effective, cost efficient approach to quickly obtain the required data typically utilized for safe scale-up of chemical processes and to accommodate changes to process recipes. Such thermal hazard screening is directly relevant to the packaging, transport, and storage of radioactive waste that is or can become chemically reactive. For such waste streams it is vital to identify safe temperature and pressure conditions and quantify adiabatic heat and gas generation rates in order to safely accommodate (or preclude) thermal instability within the waste package or storage facility. This paper illustrates widely-used thermal hazard screening bench-scale techniques that lend themselves to quickly identifying reactive hazards while providing directly scalable data for package/storage facility design. Example data are presented with discussion of how the data are analyzed for application to safe packaging and storage. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Analysis of Glycolate in Radioactive Waste by Ion Chromatography (IC) and Proton Nuclear Magnetic Resonance (H NMR)

Here, in preparation for implementing the Nitric-Glycolic (NG) acid flowsheet for the Savannah River Site (SRS) Liquid Waste System (LWS), analytical methods for determining glycolate at low concentration, below 20 mg/L in radioactive samples, were developed to support system management and safety. To accommodate the wide range of LWS matrix conditions, two alternative methods were developed, refined, and demonstrated for glycolate analysis in radioactive waste samples: ion chromatography (IC) and a proton nuclear magnetic resonance (H NMR). Investigators validated IC and H NMR methods for glycolate analysis, defined the range of applicability, and demonstrated key supporting analytical protocols. The deployed IC method is applicable in low to moderate ionic strength samples and requires sample pretreatment using a Dionex OnGuard II H + cartridge. The deployed H NMR method is more labor intensive but provides options for a broader range of matrices. Based on the results, high quality glycolate analysis of the Defense Waste Processing Facility (DWPF) condensate in Tank 22 is feasible by IC down to approximately 12 mg/L. Using H NMR, glycolate may be determined to 8 mg/L or lower depending on the run time with the potential for broader applicability of the method to higher ionic strength conditions in other tanks of the SRS LWS.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Design and performance testing of a novel far field Gamma system to assay radioactive waste in 400 and 200 litre drums - 15339

A Far Field Gamma ray Measurement system has been developed for measuring and assaying both 400 and 200 litre waste drums with a wide range of activity at nuclear power stations. These include radioactive waste drums with surface dose-rates of less than 2 mSv/h (200 mrem/h) and, at the higher range of activity, 400 litre drums with surface dose-rates of up to 50 mSv/h (5 rem/h). The assay system incorporates an electro-mechanically cooled high purity Germanium profile detector and advanced digital spectroscopic analysis electronics. A novel optimised conical-trapezoidal collimator has been incorporated to reduce the effect of the radioactive background in addition to the conventional detector lead shield with tin and copper graded lining to reduce the effect of lead X-rays. Two features have been incorporated to reduce detector dead time and maintain detector resolution when measuring high activity drums. The first is the inclusion of two automated tungsten filters of different thickness as an integral part of the detector collimator. The filters are used in conjunction with an automated rail system on which the detector platform is mounted. The detector can be positioned at different distances from the drum surface in order to reduce detector dead time. The deployment of the filters and the adjustment of the drum - detector position are automatically controlled based on a user defined dose-rate limit table. The data is provided by two Geiger-Muller dose-rate detectors, which are used to measure the drum surface dose-rate. This data is subsequently used to adjust the detector position and in cases of higher dose-rate deploy one or other of the tungsten filters. The drum rotation platform incorporates a load cell to determine drum weight and hence the drum density. This information is used to calculate an attenuation correction based on the assumption of uniform drum density and uniform distribution of activity within the drum. Where the waste drums have a more complicated and known regular internal structure, such as a small radioactive region surrounded by an annulus of shielding material, the analysis algorithms incorporated into the far field geometry and attenuation correction spectroscopic data analysis code are able to make a more accurate determination of drum activity than would be possible assuming a completely uniform drum. The detector energy and efficiency calibration is achieved using a point source with multiple gamma ray energy peaks. The response of the high purity Germanium detector is also modelled using the Monte Carlo Neutron Photon code and the model is benchmarked using the measured point source calibration data. The performance of the instrument has been determined for a range of uniform density matrix drums containing a set of Eu-152 line or rod sources located in re-entrant tubes positioned on an equal volume basis. When rotated, the test drums with volume distributed line sources simulate waste drums of different density each with a uniformly distributed radioactive source. The measurements have been validated using Monte Carlo simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Processes in Salt Repositories for Radioactive Waste Disposal

This document summarizes the key processes (thermal, hydrological, mechanical, and chemical; THMC) impacting the features of a deep geological repository for radioactive waste in salt. Some processes are natural and on-going whether the repository is there or not, and other processes are driven by the perturbation associated with the repository. The features considered here include both engineered and natural components of the repository system. The engineered barrier system (EBS) in a salt repository is quite different from those implemented for a repository in clay or crystalline rocks, because it is comprised mostly of granular salt and salt-compatible cements, rather than bentonite. When compared to other rocks (i.e., silicates), salt has unique properties that make it an excellent potential host rock. Openings and fractures in salt creep closed readily. Salt has high thermal conductivity, which can reduce peak temperatures. Additionally, far away from the excavations the porosity of salt is unconnected, which leads to essentially zero advective or diffusive transport. The small amount of hypersaline brine occurring in salt minimizes microbial activity, reduces colloid-assisted transport, and eliminates in-package criticality (i.e., chloride is a neutron poison). At the end of the report, we present a brief outline for a potential salt repository, including considerations avoided in previous repository disposal concepts. We propose considering higher-temperature processes in future disposal concepts, rather than trying to minimize the thermal perturbation of the repository. Since hot salt is drier, a dry repository would limit corrosion, gas generation, and solute transport. Openings and fractures creep shut faster in hot salt. Therefore, higher temperatures could be seen as beneficial, rather than something to minimize, through increased spacing between waste packages (increasing repository costs).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Operational History of the Area 3 Radioactive Waste Management Site

Report presents information based on characterization and monitoring data, historical documentation and records, modeling studies, evaluations, and process knowledge for the Area 3 Radioactive Waste Management Site on the Nevada National Security Site.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

On the long-term migration of uranyl in bentonite barrier for high-level radioactive waste repositories: The effect of different host rocks

Repositories for high level radioactive waste hosted in argillite or granite with bentonite backfill are the two disposal concepts that been extensively studied. A general perception is that granite has very low sorption capacity and therefore is disadvantageous in retarding the migration of radionuclides in comparison with argillite. However, when there is bentonite backfill, does the sorption capacity of granite really matter for the migration of radionuclides? How different host rocks would affect the migration of U(VI) through the bentonite backfill of an engineered barrier system (EBS), and which properties of the host rocks cause those differences in migration, given that migration of radionuclides is a critical measure in assessing the performance of a repository? Here we present two coupled thermal, hydrological, and chemical models for the transport of U(VI), each with an identical setup except for the type of host rock (argillite vs granite rock). Comparisons between our models show that the different host rocks exert their influence on migration of U(VI) via regulating the chemical conditions in the bentonite, which affect the concentration of U(VI) at the source and the adsorption of U(VI) in the bentonite. The key chemical conditions are pH, Ca + , and HCO 3 - concentration in pore-water, which are strongly affected by soluble carbonate minerals. Our models also show the occurrence of illitization (dissolution of smectite and precipitation of illite) in the bentonite, which affects the migration of U(VI) through changing pH and the quantity of adsorbents. Although generalization of current model results for granite and argillite should be done carefully, the simulations highlight the importance of pore-water chemistry in host rock, as well as bentonite-host rock interactions, when assessing the performance of a repository. In comparison to properties such as thermal conductivity, permeability, and adsorption capacity, properties that have been studied intensively, the pore-water chemistry of host rocks deserve equal or even more attention because of its strong influence on radionuclide migration.

58 GEOSCIENCES↗
DOE OSTI2014

Characterization and Delivery of Hanford High-Level Radioactive Waste Slurry

Two primary challenges to characterizing Hanford’s high-level radioactive waste slurry prior to transfer to a treatment facility are the ability to representatively sample million-gallon tanks and to estimate the critical velocity of the complex slurry. Washington River Protection Solutions has successfully demonstrated a sampling concept that minimizes sample errors by collecting multiple sample increments from a sample loop where the mixed tank contents are recirculated. Pacific Northwest National Laboratory has developed and demonstrated an ultrasonic-based Pulse-Echo detection device that is capable of detecting a stationary settled bed of solids in a pipe with flowing slurry. These two concepts are essential elements of a feed delivery strategy that drives the Hanford clean-up mission.

critical velocity↗
DOE OSTI2017

UNREVIEWED DISPOSAL QUESTION EVALUATION: Development of Inventory Screening Limits for the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

This Unreviewed Disposal Question Evaluation estimates the inventory of specific radionuclides that represent negligible risk for the Area 5 Radioactive Waste Management Site on the Nevada National Security Site. Radionuclides exceeding these screening limits are candidates for entry into the performance assessment and composite analysis model. Screening limits are developed for an expanded list of radionuclides. The screening limits are estimated as the total site inventory that would cause a receptor to receive 1/1000th of the 0.1 milliseivert (mSv) air pathway or 0.25 mSv all-pathways performance objectives. The receptor is assumed to be exposed to undiluted waste for a period of one year.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Spent Nuclear Fuel and High-Level Radioactive Waste Inventory Report

This report provides information on the inventory of commercial spent nuclear fuel (SNF), as well as Government-managed SNF and high-level radioactive waste (HLW). Actual or estimated quantitative values for current inventories are provided along with inventory forecasts derived from examining different future commercial nuclear power generation scenarios.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

The need for a characteristics-based approach to radioactive waste classification as informed by advanced nuclear fuel cycles using the fuel-cycle integration and tradeoffs (FIT) model

This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. Because heat generation is generally the most important factor limiting geological repository areal loading, this analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. Waste streams generated in different fuel cycles and their possible classification based on the current U.S. framework and international standards are discussed. It is shown that the effects of separating waste streams are neglected under a source-based radioactive waste classification system. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

New High-Throughput Measurement Systems for Radioactive Wastes Segregation and Free Release - 16166

The European Metrology Research Program (EMRP) is a coordinated research-and-development program that pursues the closer integration of national research programs. The EMRP Metrology for Decommissioning Nuclear Facilities 'MetroDecom' Joint Research Project ENV54 (JRP ENV54), a 36-month project started in September 2014, delivers research products addressing all aspects of the decommissioning process. This includes the characterization of waste materials present on the decommissioning site, pre-selection prior to free release or repository acceptance measurement, free-release measurement, the measurement of waste packages thermal power prior to the repository storage, and the monitoring of stored wastes and repositories in the very long term. This paper describes the track record of two JRP ENV54 working packages, which address the measurement facilities for pre-selection of waste materials prior to measurement for repository acceptance or possible free release (segregation measurement system); and free release (free release measurement system), based on a single standardized concept characterized by unique, patented lead-free shielding. The key objective is to improve the throughput, accuracy, reliability, modularity and mobility of segregation and free-release measurement. This will result in more reliable decision-making with regard to the safe release and disposal of radioactive wastes into the environment and, resulting in positive economic outcomes. (authors)

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