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At least 649 records · Page 36

Key Information File for Radioactive Waste Repositories - Preliminary Tests - 16169

The Key Information File (KIF) is a document designed to provide a summary of the existence, location and content of an engineered facility for the permanent disposal of radioactive wastes. It should be recognized as part of an internationally integrated system of records and memory. Its primary function would be to maintain awareness of the site, and long-term confidence in the effectiveness of the disposal system, so that the likelihood of unnecessary human disturbance is minimized. In order to ensure this document is accessible to a large audience, it would need to be limited in size. A draft format for the KIF has been proposed, and testing is being conducted on two cases, that of the 'Centre de stockage de la Manche', the closed French Repository (surface) and that of the planned Swedish Spent Fuel Repository (geological). Preliminary results show the potential for such a document. Comments and questions raised by the 'Records, Knowledge and Memory preservation' (RK and M) initiative, coordinated by the Nuclear Energy Agency (NEA) of OECD, are presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Hot Isostatic Pressing of Radioactive Nuclear Waste: The Calcine at INL - 16348

Hot Isostatic Pressing (HIP) is a technology that has been around for 60+ years. By using high temperature and high gas pressure, dry metal and ceramic powders can be consolidated and a volume decrease can be achieved. This paper presents the efforts using the HIPing process at the Idaho National Laboratory to treat the calcine radioactive waste. Once loaded into collapsible canisters, the HIPing would be used to treat the waste before final disposal. The resulting volume reduction was shown to be 20-70% and the cost ratio vs vitrification is 1:1.74. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Detection and Mitigation of Hanford Site Radioactive Waste Tank Surface Water Intrusion - 16396

The last single-shell underground radioactive waste storage tanks at the U.S. Department of Energy's Hanford Site, near Richland, WA, were removed from service in November, 1980. At that time the single-shell tanks contained 153,300 m{sup 3} (40.5 million gallons) of waste, including 34,400 m{sup 3} (9.1 million gallons) of liquids. Subsequently the Interim Stabilization Program and Waste Retrieval/Closure Project have removed 24,600 m{sup 3} (6.5 million gallons) of liquid from the tanks to reduce the potential for leakage to the environment. After interim stabilization, each tank was isolated by cutting and capping waste transfer lines, filling drains with grout, and covering likely surface water entry points. The minimum expected lifetime for the isolation features was 20 years, consistent with planned retrieval of all single-shell tank wastes by 2018. Since then the retrieval completion date has been extended. In 2011 the Tank Operating Contractor began a systematic review of the aging isolation barriers to determine if surface water was entering the single-shell tanks. This paper summarizes the results of the 2011 - 2015 activities including single-shell tank inspection results, methodologies used to identify intrusions and waste leaks masked by intrusion, and eventual deployment of a prototype forced ventilation exhauster in 2015 to evaporate water intrusion pools on the waste surface. Further efforts to limit water intrusion, including deployment of surface barriers, are discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of Treatment Process for Radioactive Wastewater Generated from Molybdenum-99 Study - 16415

Molybdenum-99 (Mo-99) high acidic waste solution was securely stored at INER for numerous years. Until 2004, the procedures of waste solution draw-out and pressure-release were accompanied, along with construction of treatment facilities. In 2009, after simulated solution and small volume Mo-99 waste solution tests, the treatment procedure was determined. Then batch-scale treatment facilities were set up. The waste solution was preserved using the instrument, along with modification of some facilities simultaneously. Ultimately all of 12 tanks of Mo-99 inorganic waste solution stored in Building 014 were finished on July, 2012. And residential 5 tanks of those solution served in Building 015B were also completed on September, 2014. A practicable process has been developed for treating radioactive waste generated from Mo-99 study, and it can effectively remove radionuclides from a nitric acid solution. At first, after the waste was sucked from the bucket, the pH value was adjusted to about 10.0 through acid-alkali neutralization. Then, an adsorbent of natural zeolite powder was added to adsorb nuclides, and two adsorption columns filled with Cs-treat and Sr-treat adsorbents were used for removing nuclear species of Cs-137 and Sr-90. In this step, the nuclear species of Cs-137 and Sr-90 contained in the liquid waste are significantly reduced to a degree for manual operation allowable. Then, nitrate ion and mercury ion were removed through another procedure. As a result, the operations were safer and more efficient than other conventional methods. Additionally, the hazard of accidental leakage caused by corrosion of the bucket after long period of storage can be effectively prevented and potential pollution threats are also eliminated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Smart Drum Technology for Radioactive and Other Hazardous Materials - 18273

Smart drum technology is the result of a coupling between a robust drum-type package containing radioactive and other hazardous materials and the ARG-US radio frequency identification (RFID) system that enables 'unattended' remote monitoring of the State of Health of the package and contents - in real time, 24/7, with automatic alert/alarm capabilities. Recent educational demos of the system are highlighted. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Development of Hot-Isostatic Pressing for Treatment of UK ILW Radioactive Wastes - 18280

Hot-Isostatic Pressing (HIPing) has been proposed as an improved method for the treatment of radioactive wastes in the UK. Sludges and slurries housed on the Sellafield site are a prime candidate for HIP treatment due to their large volumes and issues associated with the use of the baseline cementation strategy. Options have been developed by GeoRoc Ltd to co-process these wastes, including a detailed conceptual plant design. This paper will describe research and development aimed at demonstrating the viability of the concept presented in this design. This paper summarizes two studies performed to address the knowledge gaps identified for treatment of these wastes using HIP technology. The first study related to the poorly characterized nature of the wastes, which requires any treatment option to be flexible in terms of the chemical and physical properties of the waste feed. The second study involved demonstrating that HIPing can be effectively utilized at a scale applicable to treat a waste inventory estimated to be in excess of 5,000 m{sup 3}. The first section of this paper presents a summary of research carried out to determine where variation in the waste stream composition may affect the processing and wasteform produced when HIPing Magnox sludges. This study aimed to qualify where restrictions on the process envelope may exist due to poorly defined waste-stream characteristics. A baseline wasteform was selected with the aim of achieving significant waste volume reduction, while producing a physically and chemically stable wasteform. A glass-bonded magnesium-titanate/magnesium-silicate wasteform was selected as the preferred option for the Magnox sludge wastes. A series of studies investigated the densification and phase formation of waste-forms by altering key variables in the Magnox sludge simulant. To test the extremes of the processable waste loading, samples were produced and characterized at 100 wt% waste-loading for Magnox sludge, Site-Ion Exchange Plant (SIXEP) sand/clinoptilolite and a 50:50 mix of the two waste-streams. Further studies involved a) altering the waste loading to additive mass ratio (to simulate changes in the waste loading/sludge solids content), b) varying the particle size distribution from the datum to include only small or large particulates in the sludge feed; and c) introducing up to 20 wt% metallic Magnox alloy, organic species or carbonate materials (to determine how their inclusion may affect the processing and densification of the materials). A U{sub 3}O{sub 8} loaded sample was also produced using the baseline Magnox sludge formulation. All the tests performed were shown to produce monolithic, well densified waste-forms with substantial volume reductions. Minimal variations from the baseline were noted but the waste-forms produced with these deviations was determined to be of comparable quality to of the baseline product. U{sub 3}O{sub 8} was shown to behave in a similar fashion to the ZrO{sub 2} surrogate used to simulate it in the process envelope studies and was encapsulated in the magnesium titanate/magnesium silicate matrix. This work demonstrates that HIP technology can successfully produce waste-forms across the broad process envelope expected in the Magnox sludge inventory. It is also possible to produce monolithic waste-forms at 100 wt% waste loading from Magnox sludges, SIXEP sand clinoptilolite slurries or a co-processed mixture of the two. The second part of this paper summarizes a commercially relevant demonstration of HIP technology for the treatment of sludge and slurry wastes housed at the Sellafield site (UK). This study culminated in production of the largest HIPed wasteform successfully produced to date. This study was performed in a series of stages. Investigations into formulation and properties of waste-forms were performed at a lab scale to evaluate suitable wasteform systems and identify suitable combinations of additives and waste loadings. Scale up demonstrations were then performed in stages with custom-designed HIP canisters. Scale-up and characterization of key wasteform properties were successfully performed with initial canister volumes of 50 ml, 400 ml and 8 L. Suitable front-end unit operations to produce a granular, free-flowing calcined simulant were demonstrated at an engineering scale. The demonstration was successful in processing a HIP canister with an initial starting volume in excess of 100 L. A 75 % volume reduction from the raw sludge was achieved at scale while demonstrating that HIP is applicable at suitably large scale for immobilization of ILW waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Role of Indian Tribes in the Safe, Secure, and Routine Transportation of Spent Nuclear Fuel and High-Level Radioactive Waste - 19463

Far too often federally recognized Indian Tribes are the last stakeholders invited to the table to discuss a particular issue with potentially wide-ranging impacts. Indian tribes were the first governments on this continent. With respect to the future transportation of spent nuclear fuel (SNF) or high-level radioactive waste (HLW), Indian tribes are potentially impacted in a myriad of ways: economics, cultural, environmental, health and safety. Yet few regulators fully understand and appreciate the legal and political standing of Indian tribes. This paper will examine the ways in which Indian tribes must be involved with the planning for safe, secure and routine SNF and HLW shipments and the ways in which the federal government is obligated to involve tribes. Indian tribes are similar to states in that they are responsible for the health and safety of its members (which they themselves define). Tribes, like states, make, implement, interpret, and enforce laws, and provide public services - healthcare, public education, social services, drug treatment, public safety and emergency management, transportation, managing natural resources, public housing, and environmental and cultural preservation. Many tribes also operate economic enterprises - manufacturing, farming, forestry, and gaming. The most basic principle of Indian law is the principle of 'sovereignty,' that is, an Indian tribe possesses all the powers of any sovereign state. Tribal governments are not subject to state or local jurisdiction but are subject to federal jurisdiction. The tribes enjoy a government-to-government relationship (and expectation) with the federal government (i.e., agencies, like the Department of Energy, the Nuclear Regulatory Commission). Of paramount concern to Indian tribes is the doctrine of Trust Responsibility, the general principle that the federal government's conduct 'in dealing with the Indians should therefore be judged by the most exacting fiduciary standards.' With respect to SNF/HLW this means that the federal government has an obligation (or a Trust Responsibility) to ensure that tribal people, lands and resources are protected. Equally important is the expectation that federal agencies will work with the Indian tribes on a government-to-government basis before undertaking any action with potential negative consequences on the tribe's people, land or resources. The concept of government-to-government consultation between Indian tribes and the United States is long standing and well recognized. As expressed in Executive Order 13175, 'Consultation and Coordination with Indian Tribal Governments': 'Consultation - Each federal agency must have an accountable process to ensure meaningful and timely input by tribal officials in the development of regulatory policies that have tribal implications.' The purpose of consultation is to ensure that Indian tribes are consulted before government officials make decisions that may have an impact on their lives and territories. The three agencies involved with SNF and HLW transportation - the US Department of Transportation (DOT), the US Nuclear Regulatory Commission (NRC), and the US Department of Energy (DOE) - have had various degrees of success with respect to direct consultation with Indian tribes. Given that there is plenty of lead-time before the first shipments commence, there should be ample time to develop a strategy for ensuring that tribes are equal partners when planning and preparing for shipments of SNF and HLW to the first interim storage facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Panel Session 110B: US DOE High-Level Radioactive Waste (HLW) Interpretation

This panel focused on the US DOE High-Level radioactive Waste (HLW) definition interpretation. Discussion integrated topics such as the issuance of the HLW interpretation Federal Registration Notice, the status of the draft Environmental Assessment that analyzes the treatment and commercial disposal of up to 10,000-gallons of SRS Defense Waste Processing Facility recycle wastewater in South Carolina, and the path forward. Panelists with presentations: Status on the Environmental Assessment for the Commercial Disposal of Defense Waste Processing Facility Recycle Wastewater from SRS (Theresa Kliczewski); Enhancing Stakeholder Engagement on the HLW interpretation (Kara Colton, Rick McLeod); Technical Aspects of Potential Disposal of SRS DWPF Recycle Wastewater (Kent Rosenberger)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Borehole Disposal of Radioactive Waste in Israel - 20388

Israel is assessing borehole disposal of radioactive waste. With limited geological options for disposal, intermediate-depth borehole disposal is being considered in the arid Yamin Plain region of the northeastern Negev desert at depths of several hundred meters below ground surface in the vadose zone. Unlike deep borehole disposal of several kilometers, which relies on emplacement below the depth of recirculating groundwater, the safety case for intermediate-depth borehole disposal relies more on the aridity of the vadose zone, the robustness of the waste package, and the other engineered barriers (e.g., seals, backfill materials) in the disposal borehole. As a tradeoff, the shallower depth may allow the use of a larger diameter borehole than would be possible for a deeper borehole. To study the suitability of the Yamin Plain region for borehole disposal, a small-diameter characterization borehole is being planned to retrieve core samples and to better understand the vadose zone geomechanical and hydrogeochemical properties and percolation flux. The information from the characterization borehole will inform the safety case, and together with performance assessment analyses will help to identify key areas of uncertainty and guide future research and development activities aimed at demonstrating the feasibility of the intermediate-depth borehole disposal concept in Israel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Experimental and Computational Investigation of the Emission of Radioactive Products from Irradiated Fuel of Fast Sodium-Cooled Reactor

The key experimental and computational studies of the emission of long-lived radioactive fission products and actinides from uranium dioxide irradiated in BN-600 into sodium and the gas phase in the range 150–800°C, performed on the experimental bench at the Leipunskii Institute of Physics and Power Engineering, are presented. The results of measurements of the properties of aerosols of certain radioactive fission products formed upon heating of irradiated nuclear fuel in sodium are presented. The experimental results are of interest for evaluating the radiation impact of reactor facilities on workers and the public during beyond design basis accidents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Computational Errors in the Calculation of Long Radioactive Decay Chains

For safety security of objects utilizing atomic energy, the effect of numerical errors on the calculation of the activity and integrals of the activity in time when using Bateman’s formula and Siewers’ algorithm to calculate radioactive decay chains is investigated. The smallest errors obtain when using Siewers’ algorithm and Bateman’s formula together with long arithmetic. It is shown that the standard calculations of radioactive decay chains according to Bateman’s formula with double precision lead to significant errors. However, these errors have almost no effect on the dose characteristics of mixtures of radionuclides. The analysis is based on data contained in ICRP Publication No. 107.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Stable and radioactive carbon isotope partitioning in soils and saturated systems: a reactive transport modeling benchmark study

This benchmark provides the first rigorous test of a three-isotope system [ 12 C, 13 C, and 14 C] subject to the combined effects of radioactive decay and both stable equilibrium and kinetic fractionation. We present a series of problems building in complexity based on the cycling of carbon in both organic and inorganic forms. The key components implement (1) equilibrium fractionation between multiple coexisting carbon species as a function of pH, (2) radioactive decay of radiocarbon with associated mass-dependent speciation demonstrating appropriate correction of the Δ 14 C value in agreement with reporting convention, and (3) kinetic stable isotope fractionation due to the oxidation of organic carbon to inorganic forms as a function of time and space in an open, through-flowing system. Participating RTM codes are CrunchTope, ToughReact, Hytec, and The Geochemist’s Workbench. Across all problem levels, simulation results from all RTMs demonstrate good agreement.

58 GEOSCIENCES↗

Thermal and Thermophysical Properties of the Radioactive Graphite–Inert Gas System in Heating. Computer Experiment

Consideration has been given to the capabilities of the TERRA software system for modeling thermodynamically the heating of a radioactive graphite–inert gas system in a nuclear reactor. As a result of such modeling, the authors have determined processes occurring in the reactor with the involvement of Eu, Pu, and U actinoids during the heating of the radioactive graphite–argon system at pressures of 10{sup 5} and 10{sup 6} Pa, and also equilibrium constants of chemical reactions and the influence of pressure on them. The thermophysical properties of the system have been studied, and the influence of pressure on its specific volume, the heat capacity, and the molar mass of a gas phase and the weight fraction of a condensed phase has been established.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Quantification techniques of mass-separation and ion yield for the detection of radioactive isotopes

This research demonstrates two methods of quantifying isotope separation efficiency using an inductively coupled plasma mass spectrometer. One method measured the accumulated charge on a conductive collection substrate in real-time while the second method performed a chemical dissolution of the substrate with subsequent post-processing and analysis. Our previous results demonstrated and compared these methods with a multi-element standard of stable isotopes. This research applies these methods to quantifying radioactive isotopes utilizing measurements of the stable isotope at the same mass-to-charge ratio. The mass-separated radioactive isotope were fission products produced from thermal neutron irradiation of a highly enriched 235U foil. The two methods are explained and the results of this new separation methodology are compared to available cumulative fission yields.

Dion, Michael P.↗

Using radioactive material to evaluate decontamination of contaminated electronics

Electronic materials are used everywhere and can get easily contaminated by their use in the field/laboratory. The goal of this project was to use radioactive material to track the effectiveness of a cleaning procedure using an off-the-shelf cleaning gel. Radioactive potassium bromide (KBr) was used as a model contaminant in four contamination scenarios to gauge the effectiveness of a cleaning gel in the decontamination of contaminated raspberry pi’s. Finally, the investigated decontamination technique was found to be 75–97% effective in removing contamination from the tested electronic devices. 95% of the contaminated electronic devices retained their functionality post-decontamination.

38 - RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCL↗

Single-Pass Flow-Through Test Elucidation of Weathering Behavior and Evaluation of Contaminant Release Models for Hanford Tank Residual Radioactive Waste

Contaminant release models are required to evaluate and predict long-term environmental impacts of even residual amounts of high-level radioactive waste after cleanup and closure of radioactively contaminated sites such as the DOE’s Hanford Site. More realistic and representative models have been developed for release of uranium, technetium, and chromium from Hanford Site tanks C-202, C-203, and C-103 residual wastes using data collected with a single-pass flow-through test (SPFT) method. These revised models indicate that contaminant release concentrations from these residual wastes will be considerably lower than previous estimates based on batch experiments. For uranium, a thermodynamic solubility model provides an effective description of uranium release, which can account for differences in pore fluid chemistry contacting the waste that could occur through time and as a result of different closure scenarios. Under certain circumstances in the SPFT experiments various calcium rich precipitates (calcium phosphates and calcite) form on the surfaces of the waste particles, inhibiting dissolution of the underlying uranium phases in the waste. This behavior was not observed in previous batch experiments. For both technetium and chromium, empirical release models were developed. In the case of technetium, release from all three wastes was modeled using an equilibrium Kd model. For chromium release, a constant concentration model was applied for all three wastes.

chromium↗

Experiments with a prototype titanium hot cavity surface ionization source intended for electromagnetic separation of radioactive samarium and other lanthanide elements

This paper reports experimental results of a prototype titanium surface ionization source. For the first time, a lanthanide ion beam has been produced with a surface ionizer composed completely of titanium metal. Titanium does not readily activate with neutron irradiation. This offers the potential for inserting an ion source made of titanium directly into a reactor with a pre-loaded non-radioactive lanthanide target. This seamlessly integrates target irradiation with isotope separation, eliminating post irradiation sample manipulation. Samarium ion beam currents up to 960 nA have been produced in an off-line test bench equipped with rudimentary beam optics. This is a crucial step toward the development of an ionization source adopted for the electromagnetic isotope separator (EMIS) facility, which has been designed for high throughput separations of radioactive 153 Sm and other lanthanides of interest in the field of nuclear medicine. The ion current and important factors affecting the performance of the ion source, such as the ionizer temperature and thermal gradient, are discussed. Lastly, the experimental results are presented together with a discussion of future modifications to optimize the overall surface ionization source performance.

07 ISOTOPE AND RADIATION SOURCES↗

Detection of Sleeping Beauty transposition in the genome of host cells by non-radioactive Southern blot analysis

The Sleeping Beauty transposon (SB-Tn) system is being used widely as a DNA vector for the delivery of therapeutic transgenes, as well as a tool for the insertional mutagenesis in animal models. In order to accurately assess the insertional potential and properties related to the integration of SB it is essential to determine the copy number of SB-Tn in the host genome. Recently developed SB100X transposase has demonstrated an integration rate that was much higher than the original SB10 and that of other versions of hyperactive SB transposases, such as HSB3 or HSB17. In this study, we have constructed a series of SB vectors carrying either a DsRed or a human β-globin transgene that was encompassed by cHS4 insulator elements, and containing the SB100X transposase gene outside the SB-Tn unit within the same vector in cis configuration. These SB-Tn constructs were introduced into the K-562 erythroid cell line, and their presence in the genomes of host cells was analyzed by Southern blot analysis using non-radioactive probes. Many copies of SB-Tn insertions were detected in host cells regardless of transgene sequences or the presence of cHS4 insulator elements. Interestingly, the size difference of 2.4 kb between insulated SB and non-insulated controls did not reflect the proportional difference in copy numbers of inserted SB-Tns. We then attempted methylation-sensitive Southern blots to assess the potential influence of cHS4 insulator elements on the epigenetic modification of SB-Tn. Our results indicated that SB100X was able to integrate at multiple sites with the number of SB-Tn copies larger than 6 kb in size. In addition, the non-radioactive Southern blot protocols developed here will be useful to detect integrated SB-Tn copies in any mammalian cell type.

60 APPLIED LIFE SCIENCES↗