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At least 937 records · Page 52
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Knowledge Management in the Field of Nuclear Energy - How to Bridge the Gap between Retiring Workforce and Next Generations - 17101

The current phase-out decision of the German Federal Government and the 'Energiewende' - resulting from the Fukushima Daiichi Accident in 2011 - changed the nuclear market significantly. Together with the change in political and social attitude towards nuclear energy since 2002 it has contributed to a decline in technical expertise. This as well as the generation change in the nuclear technology lead to the risk of losing accumulated knowledge and experience which still are essential for the future - not only in Germany but in other countries - even though the focus in Germany will shift from operating to decommissioning, dismantling and radioactive waste management. A long-term strategy in order to preserve technical expertise to solve the forthcoming tasks in Germany was required. This strategy should include e.g. the implementation of technical programs not only for internal training of the staff but also for authorities. Together with authorities TUV NORD developed individual training concepts for a broad range of customers to preserve a transfer of know-how with the state-of-the-art in science and technology and even support international projects, especially in countries with less experience in the nuclear fields. Additionally, special data bases were developed in order to ensure the accessibility of long-term nuclear data storage. Furthermore technical conferences in different fields like decommissioning and radioactive waste disposal were established as an opportunity for knowledge exchange. Based on given examples it is being demonstrated how technical expertise can be preserved even though the field of nuclear energy in Germany has changed. (authors)

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DOE OSTI2016

Deep Borehole Field Test Laboratory and Borehole Testing Strategy

Deep Borehole Disposal (DBD) of high-level radioactive wastes has been considered an option for geological isolation for many years (Hess et al. 1957). Recent advances in drilling technology have decreased costs and increased reliability for large-diameter (i.e., ≥50 cm [19.7”]) boreholes to depths of several kilometers (Beswick 2008; Beswick et al. 2014). These advances have therefore also increased the feasibility of the DBD concept (Brady et al. 2009; Cornwall 2015), and the current field test design will demonstrate the DBD concept and these advances. The US Department of Energy (DOE) Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013) specifically recommended developing a research and development plan for DBD. DOE sought input or expression of interest from States, local communities, individuals, private groups, academia, or any other stakeholders willing to host a Deep Borehole Field Test (DBFT). The DBFT includes drilling two boreholes nominally 200m [656’] apart to approximately 5 km [16,400’] total depth, in a region where crystalline basement is expected to begin at less than 2 km depth [6,560’]. The characterization borehole (CB) is the smaller-diameter borehole (i.e., 21.6 cm [8.5”] diameter at total depth), and will be drilled first. The geologic, hydrogeologic, geochemical, geomechanical and thermal testing will take place in the CB. The field test borehole (FTB) is the larger-diameter borehole (i.e., 43.2 cm [17”] diameter at total depth). Surface handling and borehole emplacement of test package will be demonstrated using the FTB to evaluate engineering feasibility and safety of disposal operations (SNL 2016).

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DOE OSTI2015

The evolution of the Nevada National Security site waste acceptance criteria - 15312

The Nevada National Security Site Waste Acceptance Criteria (DOE/NV--325) is an evolution of requirements spanning almost three decades of waste acceptance at the Nevada National Security Site (NNSS). The original waste acceptance criteria (WAC) document was written in 1978 (document NVO-185) and underwent multiple revisions as time passed and disposal experience was gained. This document came into being when the NNSS first started receiving significant quantities of low-level radioactive waste (LLW) from offsite U.S. Department of Energy (DOE) generators, such as the Mound and Rocky Flats Plants. The original WAC was far less prescriptive and rigorous than the current criteria since many of the current DOE order directives, Federal/State laws, and regulations did not exist. The current WAC reflects the results of numerous revisions in order to address the nuances of the varied waste streams and package configurations accepted for disposal at the NNSS. This includes waste streams that exhibit challenges not previously faced or covered by the WAC and therefore, from time to time, require new information and or changes to be made to the WAC. These wastes now represent a growing percentage of the remaining waste at cleanup sites across the DOE Complex. These waste streams present challenges, such as the technical basis for ensuring safe transportation to and disposal at the NNSS. The current WAC (Rev. 10 dated June 2013), is Available online at: http://www.nv.energy.gov/library/publications/Environmental/DOENV{sub 3}25.pdf). Other challenges for the NNSS include incorporating new and emerging technologies and addressing stakeholder concerns. Addressing these challenges sometimes results in changes to the NNSS WAC. Drivers for revising the NNSS WAC over the years have included operating experience improvements, the need to specify waste origin/ownership, and identifying preferred transportation routing. This paper will discuss various aspects of the NNSS WAC, and how it has evolved based upon DOE orders and requirements, acceptance of offsite mixed wastes, acceptance of non-radioactive classified wastes, stakeholder concerns and agreements, and the changing face of the remaining legacy waste at DOE sites across the United States. (authors)

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DOE OSTI2025

DECOVALEX-2023: An international collaboration for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems

The DECOVALEX initiative is an international research collaboration (www.decovalex.org), initiated in 1992, for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. DECOVALEX stands for “DEvelopment of COupled Models and VALidation against EXperiments”. The creation of this international initiative was motivated by the recognition that prediction of these coupled effects is an essential part of the performance and safety assessment of geologic disposal systems for radioactive waste and spent nuclear fuel. DECOVALEX emphasizes joint analysis and comparative modeling of the complex perturbations and coupled processes in geologic repositories and how these impact long-term performance predictions. The most recent phase of the DECOVALEX Project, here referred to as DECOVALEX-2023, started in early 2020 and ended in late 2023. More than fifty research teams associated with 17 international DECOVALEX partner organizations participated in the comparative evaluation of eight modeling tasks covering a wide range of spatial and temporal scales, geological formations, and coupled processes. This Virtual Special Issue on DECOVALEX-2023 provides an in-depth overview of these collaborative research efforts and how these have advanced the state-of-the-art of understanding and modeling coupled THMC processes. While primarily focused on radioactive waste, much of the work included here has wider application to many geoengineering topics.

Coupled processes↗
DOE OSTI2023

Spectroscopic identification of Ca-bearing uranyl silicates formed in C–S–H systems

Portland cement-based grouts used for radioactive waste immobilisation contain a Ca- and Si-rich binder phase, known as calcium–silicate–hydrate (C–S–H). Depending on the blend of cement used, the Ca/Si ratio can vary considerably. A range of C–S–H minerals with Ca/Si ratios from 0.6 to 1.6 were synthesised and contacted with aqueous U(VI) at 0.5 mM and 10 mM concentrations. Solid-state 29 Si MAS-NMR spectroscopy was applied to probe the Si coordination environment in U(VI)-contacted C–S–H minerals and, in conjunction with U L III -edge X-ray absorption spectroscopy analysis, inferences of the fate of U(VI) in these systems were made. At moderate or high Ca/Si ratios, uranophane-type uranyl silicates or Ca-uranates dominated, while at the lowest Ca/Si ratios, the formation of a Ca-bearing uranyl silicate mineral, similar to haiweeite (Ca[(UO 2 ) 2 Si 5 O 12 (OH) 2 ]·3H 2 O) or Ca-bearing weeksite (Ca 2 (UO 2 ) 2 Si 6 O 15 ·10H 2 O) was identified. This study highlights the influence of Ca/Si ratio on uranyl sequestration, of interest in the development of post-closure safety models for U-bearing radioactive waste disposal.

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DOE OSTI2019

Remediation of Irradiated Graphite Using Microorganisms Isolated from Oil Contaminated Soil - 19643

The deployment of nuclear energy globally is hindered by the fear cleared by nuclear reactor disasters such as the Chernobyl and Fukushima reactor meltdowns. Additionally, accumulation of radioactive waste globally will, in the near future, present a significant threat to humans and the environment. Notably, the majority of nuclear waste from the energy generation industry is generated as Low Level Waste (LLW) in which the small radioactive component renders the bulk of the waste unsuitable for conventional treatment and disposal. In the case of radioactive graphite, the small radioactive component comprising radiocarbon-14 (C-14) and metallic pollutants can be removed to allow safe disposal or reuse as of the graphite. In the current study, mixed-cultures of organisms isolated from oil contaminated sites showed potential of selective bioaccumulation of C-14 from a powdered C-12/C-14 graphite matrix. The study was conducted in a continuously mixed reactor (growth vessel) constructed from Pyrex panels (30 cm x 20 cm x 15 cm) with finely ground graphite as the principal carbon source (1500-2000 mg.L{sup -1}) and spikes of glucose (500 mg.L{sup -1} every 48 hours). The growth vessel was connected to a biofilter unit by submersible pump where particulate matter was isolated for further analysis. Results showed that approximately 68% more C-14 was released from ignited dry cells than from respiration product. C-14 accumulated in cells corrected using a trap filter system, with a marked decrease of C-14 in carbon dioxide than vegetative suspended cells in the main tank. The only mechanism that could account for the differential C-14 distribution between the liquid medium and cells was possible fractionation during uptake of carbon compounds. The discriminating factor suggested in this study is the kinetic isotope effect (KIE) which allows heavier atoms to have different affinities for reaction sites on catalysis and biocatalysis. (authors)

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DOE OSTI2017

Waste Management Strategies for Production of Mo-99

Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Program for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expanded to incrementally reduce the dependence on a subcontractor.

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DOE OSTI2015

Safety assessment for the radon-type surface disposal facility at Saakadze, Georgia - 15406

The Saakadze disposal facility is located about 40 km from Tbilisi and was planned and constructed during Soviet times as Radon-type surface facility to accommodate up to 600 m{sup 3} of low and intermediate level waste generated on the territory of Georgia. The disposal facility was operated from 1963 until its closure in 1995. The vast majority of waste in the Saakadze facility has been disposed of in two near surface concrete vaults with typical Radon-type dimensions, which cover layers already show significant structural damages. In order to assess the present safety of the site and provide the competent Georgian authorities with the necessary basis to take decisions in regard to the future of the repository, the European Commission launched and financially supported a project, which was assigned to a consortium of TUV-Nord ENSYS GmbH and DBE Technology GmbH. The safety assessment of Saakadze facility, which was an essential part of this project and which is the subject of this paper was carried out by DBE Technology GmbH. The Government of Georgia has been gradually increasing efforts to improve the regulatory processes and the legislation on nuclear and radiation safety, and, in particular, on safe management of radioactive waste and sealed radioactive sources. Currently, however, there are still areas where specific Georgian laws and regulations do not exist. Accordingly, in certain aspects IAEA recommendations and best practices were used as guidelines for the methodology of the safety assessment. In spite of the large uncertainties to be considered, especially in regard to the inventory, preliminary results of the safety assessment indicate that scenarios related to the groundwater pathway do not lead to radiological impacts that are above or near to limiting values. However, human intrusion scenarios might lead to dose rates for the critical group that are exceeding these values. This is mainly due to the fact that today there is no proper engineered cover on top of the concrete vaults. Based on the comparison of existing options and considering the results of the long-term safety assessment, it is recommended to backfill the remaining voids inside the concrete vaults and to construct an engineered cover to reduce the probability of human intrusion scenarios and to minimize their potential impact. (authors)

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DOE OSTI2020

AVANTech's Liquid Radwaste (LRW) Processing System Upgrade at the Farley Nuclear Plant - 20328

This paper looks at the Liquid Radioactive Waste (LRW) processing system configuration at Joseph M. Farley Nuclear Plant (Plant Farley). A combination of factors led the staff at Plant Farley to seek a better way to manage and process liquid waste. These factors were: - Reduce effluent activity; - Reduce waste volume; - Improve cost efficiency. The LRW processing system changes made at Plant Farley are identified and discussed. The information and technologies cited in this paper are directly applicable to other nuclear power plants who might contemplate an upgrade of LRW processing systems or of individual system components. (authors)

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DOE OSTI2015

Decontamination of uranium contaminated metals - 15 years of experience - 15528

Uranium contaminated waste is somewhat of a special case among radioactive waste as the uranium, if enriched, also is a valuable asset. Incineration and pyrolysis of uranium contaminated dry active waste (DAW) was covered during WM'14. Uranium contaminated metals is however somewhat different when it comes treatment and recovery of uranium. Since the late 1990's the Metals Treatment and Recycling Facility in Studsvik Sweden has treated approximately 1500 tonnes of metals contaminated by uranium. Depending on the metal, melt-decontamination may or may not be a way of salvaging metals if the pre-treatment was not enough. The only metals that currently are subject to melt decontamination are iron based materials as well as lead. Other metals such as aluminium, brass and copper must all be cleaned prior to melting. Recovery of uranium from the secondary waste is most likely economically feasible only from the first step, the pre-melting decontamination, as the slag is just that, a slag, which is molten at a high temperature, and thereafter cooled under non-controlled conditions so uranium is trapped inside in an unspecified chemical form. This paper will describes experiences and lessons learned from 15 years of treatment, including melt-decontamination, of uranium contaminated metals. (authors)

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DOE OSTI2016

Overview of PGSFR Project

The long-term plan for advanced SFR development was authorized by Korean Atomic Energy Commission (KAEC) in 2008 and updated in 2011, which will be carried out toward the construction of a SFR prototype plant called PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) by 2028. The primary purpose of PGSFR is to demonstrate the technology reducing radioactive waste of spent nuclear fuel from nuclear power plants (NPP) in Korea. There are 24 NPPs in operation nationwide covering 30% of domestic electricity demand. The spent nuclear fuel is currently stored in the on-site storage pool of NPPs. Then, it has become a critical issue to ensure technical options to handle spent nuclear fuel in the long term basis to secure the sustainable operation of NPPs. PGSFR is a pool type sodium-cooled fast reactor with the power of 150 MWe or 400 MWth. PGSFR uses metallic fuel to make use of inherent safety characteristics of metallic fuel. The spent nuclear fuel will be processed using pyro-process to produce U-TRU-Zr fuel for PGSFR. But the initial core shall be loaded by low enriched U-Zr fuels. PGSFR has two intermediate loops with two intermediate heat exchangers (IHX) for each loop to deliver the generated heat from the core to power conversion system. There a steam generators producing super-heated steam in each loop. There are four decay heat removal circuits; two passive and two active ones. The reactor building is built on a seismic isolation base mat which can significantly mitigate seismic vibration of major components in the reactor building. The reactor building has a containment structure to isolate the reactor building under specific accident conditions and also provide structural integrity at the time of aircraft impact accident. PGSFR project is the national development program of Korea to construct the first SFR in Korea by 2028. The primary goal of PGSFR is the demonstration of reduction of radioactive waste from spent nuclear fuel by incineration in PGSFR. SFRA will flexibly cope with the licensing environment to minimize the uncertainty in licensing and safety concerns. The successful construction of PGSFR will bring Korean nuclear industry a step closer to guarantee the sustainable operation of NPPs nationwide and finally most importantly will solve the spent nuclear fuel issue of Korea. (authors)

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DOE OSTI2016

Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammable gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.

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DOE OSTI2018

Novel Platform for Digitization of Radiation Measurements, Decommissioning and Dismantling and Nuclear Waste Management - 18229

An innovative Platform for a seamless digitization of the information produced during the characterization of radioactive waste, contaminated objects, and in general radioprotection measurements has ben developed by CAEN SyS. The Platform is based on a spectroscopic handheld instrument integrating RFID technology, barcode reader, camera, voice recorder, and geo-localization, a selection of radiation tolerant RFID tags tested above 300 Gy integrated dose, ruggedized portable RFID readers and a secure database framework capable to collect data also from external sources such as gamma scanners and imaging systems. The Platform allows for real-time measurement of dose rate, radiation spectroscopy, tagging, tracking and inventorying of items. All data produced by the Platform are securely stored into the non-volatile memory of the RFID tags and in the database for both onsite and remote access to the information by operators. This paper presents the current development of the Platform and shows the advantages of such approach to achieve an easy yet reliable waste traceability, waste bagging optimization, and a structured methodology for radiation measurements, meeting an excellent quality control as increasingly required by Safety Authorities. Applications here envisaged and explored are dismantling and decontamination operations, radioprotection measurements and patrols, nuclear material management, safeguards and verification inspections. More in general, the Platform can address the needs of all plants, research laboratories and facilities that are looking to improve their Safety Standards, decrease operational costs, have a precise knowledge of the contaminated environment and objects as well as optimization for personnel dose minimization. (authors)

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DOE OSTI2020

Panel Session 90: US DOE Mixed Waste: Proposals for Dealing with Problematic Waste Streams and Policy Changes

This panel focused on generation and management of low level and mixed low-level radioactive wastes, challenges for disposition, innovative solutions applied at DOE sites, and collaboration among sites to ensure that waste disposition pathways are understood. In addition, potential policy initiatives were discussed including depleted uranium, elemental mercury, and treatment of legacy mixed waste components. As part of the introduction, Theresa Kliczewski provided an overview of why this panel is relevant: problematic mixed waste streams still exist across the DOE cleanup complex. Theresa noted that the DOE EM Headquarters Office of Waste and Materials Management (EM-4.2), for which she is employed with, continues to work with the DOE cleanup sites on a path forward for these problematic mixed waste streams. As part of this effort, EM-4.2 requested that the Energy Facility Contractors Group (EFCOG) perform a review of all the DOE sites problematic mixed waste streams. Panelists with presentations: Challenging Waste Streams: Disposition of Depleted Uranium Oxide Conversion Product (Douglas Tonkay); EFCOG WMWG Challenging Waste Subgroup (Tammy Monday); Mercury Waste Challenges (John Wrapp); Legacy Large Mixed Waste Components (Kalli Shupe)

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DOE OSTI2015

Technetium Incorporation in Glass for the Hanford Tank Waste Treatment and Immobilization Plant

A priority of the United States Department of Energy (U.S. DOE) is to dispose of nuclear wastes accumulated in 177 underground tanks at the Hanford Nuclear Reservation in eastern Washington State. These nuclear wastes date from the Manhattan Project of World War II and from plutonium production during the Cold War. The DOE plans to separate high-level radioactive wastes from low activity wastes and to treat each of the waste streams by vitrification (immobilization of the nuclides in glass) for disposal. The immobilized low-activity waste will be disposed of here at Hanford and the immobilized high-level waste at the national geologic repository. Included in the inventory of highly radioactive wastes is large volumes of 99Tc (~9 × 10E2 TBq or ~2.5 × 104 Ci or ~1500 kg). A problem facing safe disposal of Tc-bearing wastes is the processing of waste feed into in a chemically durable waste form. Technetium incorporates poorly into silicate glass in traditional glass melting. It readily evaporates during melting of glass feeds and out of the molten glass, leading to a spectrum of high-to-low retention (ca. 20 to 80%) in the cooled glass product. DOE-ORP currently has a program at Pacific Northwest National Laboratory (PNNL), in the Department of Materials Science and Engineering at Rutgers University and in the School of Mechanical and Materials Engineering at Washington State University that seeks to understand aspects of Tc retention by means of studying Tc partitioning, molten salt formation, volatilization pathways, and cold cap chemistry. Another problem involves the stability of Tc in glass in both the national geologic repository and on-site disposal after it has been immobilized. The major environmental concern with 99Tc is its high mobility in addition to a long half-life (2.1×105 yrs). The pertechnetate ion (TcO4-) is highly soluble in water and does not adsorb well onto the surface of minerals and so migrates nearly at the same velocity as groundwater. Long-term corrosion of glass waste forms is an area of current interest to the DOE, but attention to the release of Tc from glass has been little explored. It is expected that the release of Tc from glass should be highly dependent on the local glass structure as well as the chemistry of the surrounding environment, including groundwater pH. Though the speciation of Tc in glass has been previously studied, and the Tc species present in waste glass have been previously reported, environmental Tc release mechanisms are poorly understood. The recent advances in Tc chemistry that have given rise to an understanding of incorporation in the glass giving rise to significantly higher single-pass retention during vitrification are presented. Additionally, possible changes to the baseline flowsheet that allow for relatively minor volumes of Tc reporting to secondary waste treatment will be discussed.

Kruger, Albert A.↗
DOE OSTI2018

Commercial SNF Pickup Queue Under the Standard Contact and Analysis of Hypothetical Alternate Acceptance Strategies

The Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961) establishes the terms and conditions under which the U.S. Department of Energy (DOE) will make available nuclear waste disposal services to the owners and generators of spent nuclear fuel (SNF) and high-level radioactive waste (HLW), as provided in section 302 of the Nuclear Waste Policy Act of 1982 (NWPA) as amended, and by which DOE will take title to, transport, and dispose of SNF and/or HLW delivered to DOE by those owners or generators who execute the contract. The Standard Contract establishes the process for allocating the federal government’s finite waste acceptance capacity among those various owners/generators. This acceptance allocation, also known as the acceptance queue, orders acceptance and transportation priority in accordance with the principle of “oldest fuel first” (OFF). Under the OFF methodology, the oldest SNF, as measured from the date of permanent discharge from the reactor, is given the highest priority in the acceptance queue. While the date of discharge of spent fuel forms the basis of allocations, the owner/generator can send an equivalent Metric Tons of Uranium (MTU) of SNF that the owner/generator owns in its place, as long as it meets the “Standard Fuel” specifications in Appendix E of the Standard Contract. This is a technical paper that does not take into account the contractual limitations under the Standard Contract. This paper also does not specify the party or parties responsible for any costs discussed herein. This paper provides analysis of hypothetical strategies for accepting SNF and no inferences should be drawn from this paper regarding future actions by DOE. To the extent that discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail.

Joseph III, Robby↗
DOE OSTI2026

Corrosion of 316 SS in direct contact with Zircaloy-4 in repository relevant conditions

The effects of interfacial contact between 316 stainless steel (316 SS) waste canister and spent fuel cladding material (Zircaloy-4) in case of water intrusion into a deep geological repository are investigated. Hydrothermal corrosion experiments of 316 SS and Zircaloy-4 in direct contact were performed at elevated temperatures in MilliQ water and simulated Mont Terri groundwater. Hematite, magnetite, chromite, and baddeleyite form as secondary phases. pH of the Mont Terri solution increases with corrosion duration and magnetite forms preferentially instead of hematite. Thermodynamic modeling shows that iron oxidation in 316 SS controls the pH and redox potential of the aqueous environment and is in good agreement with experimental Mont Terri results, showing that experimental Mont Terri results trend to equilibrium. All oxides except hematite are predicted to exist at thermodynamic equilibrium based on calculations, implying that experimentally formed hematite will be reduced or transformed to magnetite at equilibrium. The thermodynamically predicted chromite was observed to form only at the interface where 316 SS was in contact with Zircaloy-4 at 250°C, showing that contact with Zircaloy-4 seems to accelerate the 316 SS corrosion to thermodynamic equilibrium. Finally, Mont Terri groundwater readily trends towards thermodynamic equilibrium, leading to a pH and redox buffered system that is dominated by iron in solution coming from the accelerated corrosion of 316 SS caused by Zircaloy-4 contact.

316 stainless steel↗
DOE OSTI2018

Issues Affecting the Assessment of Impacts of Disposal of Radioactive and Hazardous Waste - 18177

BIOPROTA is an international collaborative forum that seeks to address the key uncertainties in long term assessments of radionuclide and other contaminant releases into the environment arising from solid radioactive waste disposal. (see www.bioprota.org). Two workshops have been organized through the BIOPROTA forum to consider the non-radiological post-disposal impacts of radioactive waste disposal. In these workshops, it was noted that, ideally, the assessment of radionuclides and hazardous materials should be such that consistent assumptions and criteria are used in risk evaluation and that in cases where this is not possible, a clear understanding of the reasons why needs to be provided. Accordingly, a detailed review was organized through BIOPROTA to provide information to support development of a consensus on how to address this issue. The study suggests that there is a continuing need for some common measure of hazard that supports identification of risk management priorities for mixed hazardous waste. Without it, effective optimization of management of these mixed hazardous wastes will be restricted. (authors)

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