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Due to increased interest in advanced reactor deployment and their associated potential new fuel cycles, the U.S. Department of Energy (DOE) Spent Fuel and Waste Science and Technology (SFWST) program has begun to evaluate the possible implications of long term management and final disposition of the spent nuclear fuel (SNF) generated. Safely managing and dispositioning this SNF, along with any other associated radioactive waste, is the primary focus of this initial preliminary assessment. This paper summarizes efforts to evaluate the characteristics and packaging options for three types of advanced reactor SNF forms: (1) tristructural isotropic (TRISO), (2) metallic, and (3) irradiated fuel salt presented in the report titled “Storage, Transportation, and Disposal of Advanced Reactor Spent Nuclear Fuel and High-Level Waste”. TRISO and metallic SNF and their associated waste streams were emphasized because of the near-term anticipated demonstrations of X-energy’s Xe-100 and TerraPower and GE Hitachi’s Natrium advanced reactors. Preliminary information on spent fuel salts discharged from molten-salt reactors (MSRs) was also examined to provide a baseline for future efforts. All calculations and assumptions were based on publicly available information. This paper identifies several different reactors that produce either TRISO or metallic SNF as well as a few of the reactor and fuel characteristics used for the assessments. Based on these characteristics, calculations were performed to determine the applicability of packaging SNF into existing canister designs. The evaluations included geometric (e.g., dimension, volume) and mass/weight considerations, known operational approaches and loading procedures, physical and chemical considerations/conditions for storage environments, and as-loaded radiation, thermal, and criticality analyses to identify constraints on storage, transportation, and disposal. Gaps in publicly available data pertaining to reactor operation and/or fuel composition provide increased uncertainty in some evaluations. Additionally, uncertainty in packaging and SNF management operations provide additional uncertainty. However, preliminary conclusions can still be assessed through this work and are presented in this paper.
The U.S. Department of Energy’s (DOE’s) goal is to develop solutions for the long-term, sustainable management of the nation’s spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The DOE has been exploring options for an integrated waste management system to transport, store, and dispose of SNF and HLW from commercial electricity generation, as well national defense activities. To support its efforts, DOE has been exploring ways to seek broad input from the public. The role of citizen science in a wide range of projects is increasing. In the U.S. 2013 and 2015 Open Government National Action Plans, the U.S. government encouraged all Federal agencies to pursue public participation through citizen science, and a large number of agencies and organizations have responded. This paper explores the potential for considering citizen science projects as part of an integrated waste management system, including the siting of facilities. A literature review was performed to gain insights into how citizen science projects can be effectively developed and deployed, including success factors and lessons learned. Results highlight the experiences shared by many people who have developed and implemented citizen science projects, which provide valuable context for potential application to a facility siting process.
An improved approach is presented for conducting complete and defensible radiological site-specific performance assessments (PAs) to support radioactive waste disposal decisions. The basic tenets of PA were initiated some thirty years ago, focusing on geologic disposals and evaluating compliance with regulations. Some of these regulations were inherently probabilistic (i.e., addressing uncertainty in a quantitative fashion), such as the containment requirements of the U.S. Environmental Protection Agency's (EPA's) 40 CFR 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, Chap. 191.13 [1]. Methods of analysis were developed to meet those requirements, but at their core early PAs used 'conservative' parameter values and modeling approaches. This limited the utility of such PAs to compliance evaluation, and did little to inform decisions about optimizing disposal, closure and long-term monitoring and maintenance, or, in general, maintaining doses 'as low as reasonably achievable' (ALARA). This basic approach to PA development in the United States was employed essentially unchanged through the end of the 20. century, principally by the U.S. Department of Energy (DOE). Performance assessments developed in support of private radioactive waste disposal operations, regulated by the U.S. Nuclear Regulatory Commission (NRC) and its agreement states, were typically not as sophisticated. Discussion of new approaches to PA is timely, since at the time of this writing, the DOE is in the midst of revising its Order 435.1, Radioactive Waste Management [2], and the NRC is revising 10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste [3]. Over the previous decade, theoretical developments and improved computational technology have provided the foundation for integrating decision analysis (DA) concepts and objective-focused thinking, plus a Bayesian approach to probabilistic modeling and risk analysis, to guide improvements in PA. This decision-making approach, [4, 5, 6] provides a transparent formal framework for using a value- or objective-focused approach to decision-making. DA, as an analytical means to implement structured decision making, provides a context for both understanding how uncertainty affects decisions and for targeting uncertainty reduction. The proposed DA approach improves defensibility and transparency of decision-making. The DA approach is fully consistent with the need to perform realistic modeling (rather than conservative modeling), including evaluation of site-specific factors. Instead of using generic stylized scenarios for radionuclide fate and transport and for human exposures to radionuclides, site-specific scenarios better represent the advantages and disadvantages of alternative disposal sites or engineered designs, thus clarifying their differences as well as providing a sound basis for evaluation of site performance. The full DA approach to PA is described, from explicitly incorporating societal values through stakeholder involvement to model building. Model building involves scoping by considering features, events, processes, and exposure scenarios (FEPSs), development of a conceptual site model (CSM), translation into numerical models and subsequent computation, and model evaluation. These are implemented in a cycle of uncertainty analysis, sensitivity analysis and value of information analysis so that uncertainty can be reduced until sufficient confidence is gained in the decisions to be made. This includes the traditional focus on hydrogeological processes, but also places emphasis on other FEPSs such as biotically-induced transport and human exposure phenomena. The significance of human exposure scenarios is emphasized by modifying the traditional acronym 'FEPs' to include them, hence 'FEPSs'. The radioactive waste community is also recognizing that disposal sites are to be considered a national (or even global) resource. As such, there is a pressing need to optimize their utility within the constraints of protecting human health and the environment. Failing to do so will result in the need for additional sites or options for storing radioactive waste temporarily, assuming a continued need for radioactive waste disposal. Optimization should be performed using DA, including economic analysis, invoked if necessary through the ALARA process. The economic analysis must recognize the cost of implementation (disposal design, closure, maintenance, etc.), and intra- and inter-generational equity in order to ensure that the best possible radioactive waste management decisions are made for the protection of both current and future generations. In most cases this requires consideration of population or collective risk. (authors)
This panel focused on the progress towards closure of the legacy ponds and silos on the Sellafield Site. The discussion included progressive learning, technical challenges, safety, and regulatory flexibility required to achieve cleanup. The panel included NDA, regulators and Sellafield Ltd. Presentations proposed include: 1 - Developments in the Cleanup of Sellafield: An Owners Perspective (Adrian Simper, Nuclear Decommissioning Authority UK (United Kingdom)); 2 - Sellafield Ltd (Lee Peck, Sellafield Ltd. (United Kingdom)); 3 - Developments in the Clean-up Mission at Sellafield (Ian Dallas, Office for Nuclear Regulation (United Kingdom)); 4 - Developments in the Clean-up Mission of Sellafield, UK (Steven Thomson, National Nuclear Laboratory (United Kingdom))
The U.S. Department of Energy (Department) Office of Legacy Management (LM), established in 2003, manages the Department’s postclosure responsibilities and ensures the future protection of human health and the environment. During World War II and the Cold War, the Federal government developed and operated a vast network of industrial facilities for the research, production, and testing of nuclear weapons, as well as other scientific and engineering research. These processes left a legacy of radioactive and chemical waste, environmental contamination, and hazardous facilities and materials at well over 100 sites. Since 1989, the Department has taken an aggressive accelerated cleanup approach to reduce risks and cut costs. At most Departmental sites undergoing cleanup, some residual hazards will remain at the time cleanup is completed due to financial and technical impracticality. However, the Department still has an obligation to protect human health and the environment after cleanup is completed. LM fulfills DOE’s postclosure obligation by providing long-term management of postcleanup sites which do not have continuing missions. LM is also responsible for sites under the Formerly Utilized Sites Remedial Action Program (FUSRAP). Currently, the U.S. Army Corps of Engineers (USACE) is responsible for site surveys and remediation at FUSRAP sites. Once remediation is completed, LM becomes responsible for long-term management. LM also has responsibility for uranium processing sites addressed by Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA). UMTRCA Title II sites are sites that were commercially owned and are regulated under a U.S. Nuclear Regulatory Commission (NRC) license. For license termination, the owner must conduct an NRC-approved cleanup of any on-site radioactive waste remaining from former uranium ore-processing operations. The site owner must also provide full funding for inspections and, if necessary, ongoing maintenance. Once site cleanup is complete, LM accepts title to these sites on behalf of the United States and assumes long-term management.
The aim of the present work is to assess the risk of explosion in closed containments used for the transportation of nuclear materials or nuclear waste. Indeed, it is very well-known that hydrogen can be produced due to (i) the radiolysis of different materials within the containment, (ii) the thermal decomposition of mainly the organic part in the containment. Since hydrogen has a very low ignition energy and a very wide flammability domain, it is important to determine the risk of ignition of the subsequent mixture produced by the aforementioned mechanisms. The quantity of the hydrogen that can be produced can vary depending on the containment type and on the state of the material/waste being transported. It is then mandatory to have a very good knowledge not only of the flammability domain of hydrogen in air diluted with water vapor. To do so, an experimental work on the flammability domain of multiple ternary mixtures containing H{sub 2}/air/H{sub 2}O{sub vap} is being conducted at the CNRS-ICARE laboratory within collaboration with the CEA. The present work will focus on the experimental determination of the lower flammability limit of H{sub 2}/air/H{sub 2}O{sub vap} using the spherical method. These experiments have been carried out for an initial temperature of 200 deg. C and at three different initial pressures: 0.3, 1 and 2 bar. The monitoring of the flammability conditions has been conducted using fast flame imaging and pressure measurements. (authors)
Melter feeds for high-level nuclear waste (HLW) typically contain a large number of constituents that evolve gas on heating, Multiple gas-evolving reactions are both successive and simultaneous, and include the release of chemically bonded water, reactions of nitrates with organics, and reactions of molten salts with solid silica. Consequently, when a sample of a HLW feed is subjected to thermogravimetric analysis (TGA), the rate of change of the sample mass reveals multiple overlapping peaks. In this study, a melter feed, formulated for a simulated high-alumina HLW to be vitrified in the Waste Treatment and Immobilization Plant, currently under construction at the Hanford Site in Washington State, USA, was subjected to TGA. In addition, a modified melter feed was prepared as an all-nitrate version of the baseline feed to test the effect of sucrose addition on the gas-evolving reactions. Activation energies for major reactions were determined using the Kissinger method. The ultimate aim of TGA studies is to obtain a kinetic model of the gas-evolving reactions for use in mathematical modeling of the cold cap as an element of the overall model of the waste-glass melter. In this study, we focused on computing the kinetic parameters of individual reactions without identifying their actual chemistry, The rough provisional model presented is based on the first-order kinetics.
The Port Granby Project is part of the Port Hope Area Initiative (PHAI) through which Canadian Nuclear Laboratory (CNL) is managing 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 1) constructing a new, secure long-term waste management facility (LTWMF) and supporting infrastructure, and 2) 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. This paper provides an overview of the challenges of water management (control of storm and groundwater at the LTWMF and PGWMF) on a large-scale construction project with changing and unpredictable weather patterns. Consistent with increased observations of extreme weather events, over two consecutive construction seasons, weather changed from one of the driest on record (high temperatures, dry conditions and strong wind) to one of the wettest on record (short duration high-intensity rainfall that flooded the LTWMF, excavations and on-site PGWMF and LTWMF retention ponds). This paper describes the collaborative approach among Wood (Constructor), AECOM (Engineer of Record), and CNL to adjust to extremes in weather to control dust and excessive water during construction. The summer of 2016, one of the driest summers on record in southern Ontario, was followed by one of the wettest in 2017. In 2016, the project team used water from the uncontaminated on-site storm water ponds, applied a polymer product for dust control across the site to augment water applications, obtained a Permit to Take Water from the provincial government to pump water from Lake Ontario, procured large tanks for storage of Lake Ontario water to meet peak demand, and retained additional water trucks/operators for day to day site operations. In stark contrast, to accommodate heavy rains in 2017 the project team installed ∼6 million liter aboveground storage tanks with secondary containment for contaminated water storage; procured 100+ mobile water storage tanks to supplement the on-site storage, utilized LTWMF cells sumps, excavations and retention ponds to store and manage onsite water; and adjusted additional staff resources (more staff and longer shifts) during anticipated periods of high-intensity rainfall, at times managing water around the clock. Recognizing the need to minimize the release of contaminated soil and water from the construction site at all times, these efforts illustrate the importance of anticipating and reacting to extreme weather events on a more regular basis. (authors)
The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste forms include: •Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. •Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. •Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. •Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. •An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. •An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. •Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. •Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). •Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported.
UCOR, an AECOM-led partnership with Jacobs, manages the cleanup of the 2,200-acre East Tennessee Technology Park for its client, the U.S. Department of Energy (DOE), in Oak Ridge, Tennessee. The site was contaminated with radioactive, hazardous and industrial wastes generated by more than 40 years of national defense and energy missions. While successfully conducting its cleanup responsibilities, DoE's Oak Ridge Office of Environmental Management and UCOR are strongly committed to developing the current and future workforce through the application of multiple partnering opportunities and the use of innovative learning techniques that are adaptive to different generations of learners. The goal is to maintain a highly-skilled workforce for cleanup activities while establishing bridges for the development and placement of our future workforce. The training and professional development of the UCOR workforce is provided by multiple sources, including UCOR in-house training, trade unions, regional higher educational institutions and regional technical schools. In concert with these partners, UCOR is working to maintain and expand the existing talent base, enhance safety, improve processes, and resolve challenges through aggressive and well-coordinated initiatives designed to develop the future workforce. While UCOR's efforts are primarily focused on the cleanup of DOE facilities, its workforce development programs can also benefit the commercial nuclear power industry - both in ongoing plant operations and in decommissioning aging plants as they come to the end of their useful life. UCOR's Workforce Development Program is built on four distinct pillars: - Higher Education; - Community Colleges; - Apprenticeship Readiness Program; - K-12 STEM Program. (authors)
The Savannah River Site (SRS) was constructed during the 1950's and became one of the major producers of plutonium for the United States during the Cold War. During its production life, the F/H Area Seepage Basins received approximately 1.8 billion gallons of acidic waste containing radionuclides and dissolved heavy metals. This led to the creation of a highly contaminated groundwater plume with uranium (VI) as a key contaminant of concern within the plume. Now, as a designated hazardous waste management facility, the main activities taking place at the site are nuclear storage and remediation of contaminated soil and groundwater from radionuclides. Humic substances (HS) are major components of soil organic matter having the ability to influence migration behavior and fate of heavy metals and are being investigated for potential use in environmental remediation at SRS. Essentially, HS are polyfunctional organic macromolecules formed by the chemo-microbiological decomposition of biomass or dead organic matter. HS are able to interact with both metal ions and organic compounds based on solubility. Humic acid (HA) represents the fraction of HS soluble at pH greater than 3.5. HA carries a large number of functional groups, provides an important function in ion exchange and is a metal complexing ligand with a high complexation capacity. Silica is the term applied to solid forms with the stoichiometric composition of SiO{sub 2}. The colloidal silica used in this work is amorphous and non-porous in suspension; silica's net surface charge for pH values higher than 3 is negative, due to the isoelectric point being 1.5 - 2.5. Humic acid has been previously used in remediation techniques and silica is found naturally in the soil. This investigation studied the synergistic interactions between four key components: U(VI), humic acid (HA), colloidal silica and SRS sediment under varying pH conditions ranging from 3 to 8. Multi-component batch systems were constructed to effectively analyze the removal of U(VI). The batches were pH adjusted, left to equilibrate for a minimum of 48 hours on a shaker, and then centrifuged. A set of samples, filtered to remove any colloids in suspension, and a set of unfiltered samples were analyzed. For the samples at acidic pH, the uranyl cation is the dominant species; with increasing pH, the mononuclear and polynuclear hydrolyzed uranyl ions become dominant based on speciation modeling. In filtered samples, uranium removal is highest at low pH with the negatively charged moieties of HA readily binding to the uranyl cation. Percent removal of U(VI) steadily decreased with an increase in pH and then remained comparable at a neutral/alkaline pH with the change in uranyl speciation. The percentage of colloidal silica removal was observed to be approximately 80% for both sediment- and non-sediment-bearing samples. As the pH increased, the silica removal in the non-sediment set showed a slight decrease and was in the range of 70% - 75%. Uranium removal in the unfiltered samples showed similar trends as compared to the filtered samples; however, the removal of uranium in the unfiltered samples was slightly lower compared to the filtered samples. Colloidal silica removal in non-sediment-bearing batches showed a decline with increasing pH, from 90% to 50%; conversely, the silica removal remained constant for the sediment-bearing batches at approximately 60%. Colloidal silica does not seem to have a significant effect on the removal of U(VI) before filtration due to U(VI) bound to colloidal silica remaining in suspension. The presence of humic acid, silica colloids, and SRS sediment was shown to synergistically effect the removal of U(VI). (authors)
About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)
This panel focuses on the young professional views from around the world on radioactive waste management. An informal panel will enable these future leaders of our industry to share ideas and experiences and develop contacts within peer groups on a genuine international level. The informal and interactive panel format will stimulate valuable discussion among the participants and the audience. The panel will be led by young professionals who will actively encourage participation from all those attending. Mentoring will be discussed. This panel aims to encourage fresh thinking and provide an opportunity for an open and frank discussion of issues. Presentations proposed include: 1 - International Youth Nuclear Congress (Alexander Borovskis, AusYGN (Australia)); 2 - AECOM Early Career Professionals Technology Exchange (Natalie Young, Washington River Protection Solutions (USA)); 3 - Young Professionals in Nuclear Science and Engineering - An International Perspective (Shona Thompson, Canadian Nuclear Safety Commission (Canada))
The Nuclear Research and consultancy Group is located at Petten, the Netherlands, and operates the High Flux Reactor (HFR). The HFR is used for isotope production for medical purposes and material research. During its exploitation several types of radioactive waste are produced, which partly have accumulated at the reactor site. In addition, NRG operated the Low Flux Reactor (LFR), a small 30 kWmax Argonaut type reactor which was used for research and education purposes. Over the past decade several projects on waste management and decommissioning were carried out requiring various approaches, including the radiochemical characterisation of the different emerging waste streams as required by the Dutch regulator. This paper focuses on the practices and technical aspects of the destructive and non-destructive characterization of various types of low- and intermediate-level waste arising from a number of nuclear activities performed at the premises of NRG, i.e. the Dutch legacy waste management program, the treatment of ion exchange resins, and the decommission of the Low-Flux Reactor. (authors)
The European Commission under their Instrument for Nuclear Safety Cooperation (INSC) has initiated a project supporting Iraq's efforts for the safe disposal of short lived low- and intermediate-level radioactive waste (SL-LILW) arising from the decommissioning of Iraqi nuclear installations. The INSC project includes both the development of an initial conceptual level basic design and preliminary safety assessment (PSAR) report and the subsequent detailed engineering design and supporting intermediate safety assessment report (ISAR). The project is being executed by a Consortium led by NUKEM Technologies Engineering Services GmbH and joined by DBE Technology GmbH of Germany; JAVYS a.s. of Slovakia and MCM Environmental Services Ltd of the United Kingdom. The current paper focuses on the development of the conceptual level basic design. In previous decades Iraq had a significant nuclear program which encompassed ten nuclear sites spread out across the country. The largest site was located at the Al-Tuwaitha site near Baghdad where 18 nuclear facilities were located. The facilities suffered significant damage from bombing and subsequent looting during the first Gulf War in 1991. Currently, the Ministry of Science and Technology (MoST) is the radioactive waste management operator in Iraq and also the license holder for the ten Iraqi nuclear sites, all of them heavily damaged. Because of the ongoing security concerns in Iraq MoST has selected the Al Tuwaitha Nuclear Center as the preferred and best suited location for development of an SL-LILW repository. As an initial step in development of the conceptual design various repository design options for the site were examined and assessed. The assessment considered similar facilities that have been implemented internationally. At the conclusion of the assessment and in conjunction with the Iraqi Beneficiaries from MoST, a near-surface vault design was selected for the facility. The design of the facility relies on experience gained from existing low-level waste repositories at Centre de l'Aube in France and the El Cabril solid radioactive waste disposal facility in Spain as well as on more recent experience gained in the design of the Bulgaria National Disposal Facility (NDF) under construction near the Kozloduy Nuclear Power Plant. Similar to these reference designs, the concept for the Al Tuwaitha Disposal Facility (ATDF) will rely on a series of reinforced concrete vaults into which containerised waste will be placed. Each vault will consist of two chambers designed specifically to hold the waste. Central drainage points in each waste chamber will lead to an infiltration water collection system designed to detect any failures in the waste containment. Each disposal vault will be sealed after filling by installation of a reinforced concrete closure slab. Once all of the waste has been disposed of the site will be decommissioned and a multilayer protective cover will be installed over the sealed disposal vaults. The multilayer cover will act to further isolate the waste from the environment by limiting the access of water to the waste. Due in part to the presence of shallow groundwater at the site and to avoid potential damage to the facility by flooding from the nearby Tigris River the concept for the Al Tuwaitha Disposal Facility requires construction of the disposal vaults on a raised platform several meters above ground level. Construction of this platform will be integrated into the conceptual safety design and materials will be selected which further enhance the isolation function of the disposal system by hindering or delaying the release of radionuclides to the environment. The infiltration water collection system, which will be constructed in a specifically designed accessible gallery embedded and fully integrated into the platform, will provide access to inspection points located beneath each disposal chamber. The ATDF when completed will be a further demonstration of the adaptability of the disposal vault concept for use in isolating SL-LILW from the environment. (authors)