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DOE OSTI2016

Evaluation and Impacts of Mercury in the SRS Liquid Waste System - 16121

The Savannah River Site (SRS) Liquid Waste System (LWS) stores radioactive waste in 43 underground tanks. The radionuclides in the waste are removed through a series of separation processes and the low-level fraction is immobilized in a grout waste form while the high level fraction is disposed in a glass waste form. Mercury originated from decades of canyon processing (used as a catalyst for dissolving the aluminum cladding of reactor fuel) and is present throughout the LWS (∼60 metric tons). Mercury has long been a consideration in the LWS, from both hazard and processing perspectives. Mercury is removed from the LWS in many ways including: 1) the Tank Farm evaporator condensates; 2) at the Defense Waste Processing Facility (DWPF) where it is steam-stripped and removed during the feed preparation processes; 3) it is removed at the Effluent Treatment Plant (ETP); and it is immobilized in the low-level, grout waste form. An integrated, system-wide evaluation and assessment of mercury behavior in the LWS is being performed. The recently completed Phase I evaluation focused on mercury inventory and speciation in the LWS; mercury holdup and chemical processing behavior; mercury impact Identification, including worker safety and equipment degradation; and mercury removal and disposal options. The mercury removal system at the DWPF has been unable to remove mercury successfully due to equipment and chemistry challenges since the 2008 timeframe. Phase I evaluations indicate that approximately 43 percent of the mercury in waste sent to the DWPF is now returning to the Tank Farm in the recycle stream and is concentrating to higher than expected levels (∼500 mg/l) around the 2H evaporator system. The chemical behavior of mercury is further complicated by the formation of higher than expected quantities of organomercury species, specifically methyl mercury, now present in the DWPF recycle evaporator system (2H Tank farm evaporator), which is used to volume-reduce the recycle stream. Phase II activities are building on the Phase I mercury activities, including additional sampling and characterization activities, a re-assessment of overall system knowledge, ranking and prioritizing critical gaps/information in mercury behavior across the flowsheet, assessing impacts of mercury removal and disposal options, and documenting an action plan for overall mercury management in the LWS. Phase II activities include two System Engineering Evaluations to 1) 'Reestablish Mercury removal Capability within DWPF', and 2) 'Determine the Best Alternative Mercury Removal Location with the LWS'. This paper provides the results of the Phase I and II mercury evaluations for the SRS LWS. (authors)

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DOE OSTI2016

Using Green's Functions to Compute 3 Dimensional Neutron Flux Maps - 16016

The knowledge of the radionuclide content of radioactive waste is of utmost importance for safety and waste management reasons. Numerical simulations are used by EDF-DP2D to anticipate dismantling and radioactive waste management. The main numerical activation schemes include the computing of a 3 dimensional neutron flux map, obtained on the basis of a neutron propagation calculation. The codes used solve the transport equation called the Boltzmann equation without simplification. The Monte Carlo method is used to solve the Boltzmann equation. Random series of numbers are used to simulate the lives of millions of neutrons. The codes follow each neutron individually, from its birth to its disappearance by leakage, absorption, or fission, and then the neutron flux map is calculated at the nominal power rating conditions. The main drawback of the Monte Carlo codes is the large computing time needed to converge within acceptable statistical criteria. This extended time period is incompatible with the multiple fuel managements used by EDF to operate its NPPs because it is necessary to make a Monte Carlo calculation for each fuel loading pattern. The solution developed by EDF-DP2D is to use the importance factors of Green's Functions to simplify the computing of the 3 dimensional neutron flux maps. Physicists talk about Green's Functions where mathematicians talk about elementary solutions of linear equations with constant coefficients. With an activation scheme based on the Green's Functions approach the 3D neutron flux is obtained on the basis of the convolution of the normalized source contributions (i.e. normalized response matrices) with the neutrons emitted by the fuel assemblies. The response matrices to unitary sources are computed with the Monte Carlo code MCNP for different tallies of interest and the neutron sources emitted by the fuel are computed with a core code for different fuel loading patterns. The principal interest of this approach is the calculation of the response matrices to unitary sources that can be used for all power distributions, thus limiting the Monte Carlo computing time. (authors)

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DOE OSTI2019

Oxidation of Organics in the Defense Waste Processing Facility Condensate

Defense Waste Processing Facility (DWPF): The state of South Carolina currently holds 35 million gallons of radioactive waste in 51 underground tanks. DWPF was built in an effort to convert the waste into a more stable form. It turns the sludge waste from the tanks into glass, which will keep the waste stable for 10,000 years. Currently, DWPF uses formic acid in its flowsheet and there is one major problem associated with its use: it decomposes into hydrogen, a flammable gas. Glycolic acid, however, has been found as a possible alternative that performs similarly to formic acid but does not directly form hydrogen. One of the main issues with the glycolic acid flowsheet is the potential for there to be glycolate and other organics in DWPF's Recycle Collection Tank (RCT). These organics could lead to hydrogen generation later in the downstream evaporators and must be destroyed in the RCT in order for glycolic acid to be a suitable replacement for formic acid. Oxidants can be added to the RCT to destroy the organics. Two oxidizers, permanganate and Fenton's reagent (iron catalyzed hydrogen peroxide), are possible candidates. Find optimum processing conditions for the oxidation of organic species in a typical RCT simulant. A reactor served as a small-scale version of the RCT. During each run, an RCT simulant with 8 added organics, including glycolate, was created in the reactor. Either permanganate or peroxide was slowly added for each run through a syringe pump. Each experiment lasts 6 hours and a total of 23 runs are scheduled. Variables of Interest: Oxidizer (Permanganate or Peroxide); Temperature (15 deg. C to 50 deg. C); Oxidant addition time (10 mins to 2 hours); Amount of oxidant (1 x to 1.5 x stoichiometric requirements); Solution pH (3 to 13); Glycolate concentration (125 to 250 mg/kg). Offgas Analyzers: Mass spectrometer and FTIR spectrometer. Samples: pulled at 7 different times throughout the run to determine the amount of organic and inorganic carbon and concentrations of various organics through ion chromatography (IC). Only 3 of the runs have been completed at this point, all with permanganate. Based on the first 3 runs, permanganate seems to oxide the majority of the organics and the run at pH 8 and 32.5 deg. C had the most effective oxidation. The rest of the runs will need to be performed to determine the best oxidant and optimum values of the other variables of interest. Once the experiment is completed, DWPF will select one of the oxidants. Testing will continue to determine its performance during foam-over and other non-ideal conditions. Finally, the oxidant will be tested with actual radioactive waste to confirm its effectiveness.

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DOE OSTI2021

FOURTH QUARTER / ANNUAL TRANSPORTATION REPORT FISCAL YEAR 2021

Waste Shipments to and from the Nevada National Security Site, Radioactive Waste Management Complex. This report satisfies the U.S. Department of Energy (DOE) commitment to prepare a quarterly summary of waste shipments to the Nevada National Security Site (NNSS) Radioactive Waste Management Complex (RWMC) in Area 5.

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DOE OSTI2017

Thermal Testing of Waste Packages for a Storage Environment

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard TRU waste to the Waste Isolation Pilot Plant (WIPP) for disposal. The POC consists of an inner Pipe Container (PC) surrounded by organic fiber dunnage inside of a 7A drum. The PC was design to maintain separation of fissile material and to provide shielding from radiation. The POC has many design features that are similar to Type B packages that do provide protection from engulfing fires. However, it was not specifically designed to prevent release of stored material during a long-duration fire. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites (e.g., Waste Isolation Pilot Plant). Radioactive waste storage operators would like to be able to claim some level of protection is provided by the POCs for thermal assaults that could occur within storage facilities when these POCs contain PCs with combustibles. In 1997 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provide during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drums generating sufficient internal pressure to pop off its lid and expose the top of the PC to the fire environment. However, these tests are not bounding for POC/PC containing combustibles. With the addition of combustible materials in the PC, the possibility exists for generation of higher quantities of gases and much higher pressures inside the PC, which can lead to early rupture of the PC filter and higher radioactive aerosol release fractions. Therefore, a new series of tests was required to determine the temperature response of the exposed PC and determine the amount of material that could be released. DOE contacted SNL in 2015 to discuss the 1997 SNL tests and came to the conclusion that those tests were not bounding for POC/PC containing combustibles Therefore, DOE commission a new set of tests at SNL. Starting in 2015, SNL conducted fire tests to examine whether the contents of the PC would reach a temperature that would result in generation of sufficient gas to cause overpressurization of the PC, and subsequent release of its content. This paper will describe the various test conducted thus far, and present preliminary results from these tests.

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DOE OSTI2019

Panel Session 16: Hot Topics and Emerging Issues in US Commercial LLRW Management and Disposal

This panel focused on emerging issues in commercial LLW management in the US from the perspective of active members of the Low-Level Radioactive Waste Forum, Inc. State, compact, federal and industry officials will share their views on a variety of timely and significant topics related to low-level radioactive waste management, disposal and related issues. Presentations proposed include: 1 - Overview of NRC Activities and Initiatives (John Tappert, US NRC (USA)); 2 - Advocates for Responsible Disposal in Texas (Edward Selig, Advocates for Responsible Disposal in Texas (ARDT) (USA)); 3 - Hot Topics and Emerging Issues in U.S. Commercial LLW Management and Disposal (Joe Weismann, US Ecology (USA)); 4 - DOE/NNSA Office of Radiological Security (ORS) Sealed Source Management and Disposal (David Martin, Energetics Incorporated (USA), Temeka Taplin, US DOE NNSA (USA))

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DOE OSTI2015

Understanding the consequences of hypothetical post-closure transient criticality in a geological disposal facility - 14439

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher-activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of criticality safety for the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal all waste packages will be safely sub-critical. Containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is not possible for some time after disposal. However, as waste packages degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Although the likelihood of a critical configuration developing is demonstrably low, it cannot be assumed to be zero over long post-closure timescales. The UK environment agencies' Guidance on Requirements for Authorisation for a GDF includes a specific requirement to consider the consequences of hypothetical 'what-if' post-closure criticality. RWM has undertaken a programme of research to model and understand the consequences of hypothetical post-closure criticality events. Initially focused on the scenario of fissile material relocation and accumulation following intermediate level waste (ILW) disposal, two bespoke computer models were developed. The quasi-steady-state (QSS) model considered critical systems maintained by accumulation of fissile materials where the temperature feedback remains negative throughout the transient. The rapid transient model (RTM) considered systems with positive temperature feedback. The QSS model development included a successful comparison with the Oklo natural reactors. With a requirement to consider post-closure criticality for a wider range of potential waste types under different illustrative disposal concepts, the initial transient criticality models had some limitations. For the QSS model this was the assumption of fixed geometry, while the complex functional inputs required in the RTM were not readily available for different materials. For the QSS model this paper describes the model development to allow changing volume; this has enabled its application to 'what-if' scenarios of flooding within a spent fuel waste package and the slumping of a stack of ILW packages. For rapid transient criticality this paper explains how the development of a simpler 'bounding approach' model to complement the RTM has enabled transient criticality analysis for a wider range of host materials. The role of this research in demonstrating post-closure criticality safety of a GDF in the UK is presented in a companion paper which also draws on the results of a parallel research project to demonstrate that the likelihood of post-closure criticality is low. (authors)

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DOE OSTI2016

Secondary Waste Form Development and Data Package for Hanford's Integrated Disposal Facility Performance Assessment - 16479

The Hanford Low-Activity Waste (LAW) Vitrification facility is planned to immobilize Hanford tank waste for on-site, near-surface disposal at Hanford's Integrated Disposal Facility (IDF). During this process, liquid and solid secondary wastes will be generated which are also slated for disposal at the IDF. In order to receive authorization to dispose of the waste in the IDF, a Performance Assessment (PA) must be completed in accordance with DOE Order 435.1, Radioactive Waste Management, and its accompanying Manual. The PA must include calculations of potential doses to a member of the public for a 1,000-year period to provide a reasonable expectation that performance objectives will not be exceeded and that the waste will be managed in a manner that is protective of worker and public health and safety, and the environment. Liquid and solid secondary wastes are planned to be immobilized within cementitious waste forms similar to those used elsewhere in the United States and around the world for permanent disposal of low-level radioactive waste (LLW). Data packages and supporting documents are being developed to provide the technical underpinning for the analyses of contaminant releases to the environment from these waste forms for the IDF PA. Traditionally, releases from cementitious waste forms have utilized a diffusion-limited release model using contaminant-specific diffusivities measured in standard leach tests as inputs. More mechanistic modeling techniques are under development to increase the rigor of the PA analyses. In addition, recent testing has shown diffusivities of key contaminants can be reduced by several orders-of-magnitude through optimized formulation of the waste form and use of additives to bind specific contaminants. (authors)

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DOE OSTI2016

Evaluation of Current Knowledge for Building the Dutch Salt Safety Case - 16024

This paper provides an overview of an evaluation for building the Safety Case of geological disposal of radioactive waste in rock salt in the Netherlands. The evaluation comprises the present knowledge about the safety and feasibility of geological disposal designs considered in the past, thereby classifying the information taking into account the methodology and structure of a Safety Case. The evaluation is performed as part of the Dutch research programme OPERA (OnderzoeksProgramma Eindberging Radioactief Afval - Dutch acronym for Research Programme into Geological Disposal of Radioactive Waste) by a consortium consisting of GRS from Germany, and TNO and NRG from the Netherlands. Basis of the evaluation is the available Dutch and international scientific-technical information about geological disposal in rock salt, especially from the German and US programmes. The evaluations have shown that although much knowledge on the topic exists, the available information need to be integrated in order to develop a Dutch Safety Case on rock salt. For example, the characteristics of the Dutch radioactive waste investigated for disposal have changed considerably since the last Dutch research programme ended in 2000. This implies that the previously considered Dutch facility concepts for geological disposal in rock salt need to be adapted to the newly regarded waste types such as spent fuel from research reactors and large amounts of LLW, including depleted uranium. Introducing these aspects into the iteration towards a salt Safety Case would also require an update of the post-closure safety assessment. (authors)

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DOE OSTI2018

Andra's Underground Research Laboratory as a Tool to Prepare Monitoring System of Cigeo Project - 18070

Andra, as French radioactive waste management agency, plans to build an underground disposal facility, named Cigeo, dedicated to high and intermediate level, long lived radioactive wastes. To ensure the safety during the disposal operations that includes the principle of reversibility, the design of the associated monitoring system will be based on specific research and development program and feedback obtained over more than 15 years on tests performed in representative conditions. Andra's Underground Research Laboratory (URL), located at 500 m deep, is excavated in a clay-rich formation. It is dedicated to (i) assess the thermo-hydro-mechanical, chemical and radiologic behaviors of the rock and supporting layers (concrete and steel), (ii) test envisaged monitoring system, (iii) provide feedback on operational organization and safety management and (iv) qualify operating methods (digging, concrete pouring, monitoring system design, etc.) to be implemented in the future disposal. Data collected from (i) characteristics of samples retrieved from boreholes, (ii) experts' observations, (iii) sensors installed in situ and (iv) manual measurements, are integrated in a database. This database is dedicated to samples management (more than 73 600 classified samples), data measurements collection, and conversion to physical values and available to plots or exports. Each day, approximately 1.6 million of values are integrated in the database. Finally, a geo-referenced database is linked to the main database in order to allow the spatial localization, both on the surface or in the underground galleries. All these tools give confidence in Andra's capacity to monitor Cigeo and be ready for the future industrial pilot phase. The monitoring system will also allow, step by step, optimizing the disposal design, performing regular safety control, helping in decision making for reversibility and improving long term safety. (authors)

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DOE OSTI2015

Analysis of the likelihood of post-closure criticality in a geological disposal facility - 14349

Radioactive Waste Management (RWM) is responsible for implementing geological disposal of the UK's higher activity radioactive wastes. These wastes include significant quantities of fissile nuclides and, therefore, demonstration of the criticality safety of the wastes under disposal conditions forms an important component of RWM's disposal system safety case. At the time of disposal, controls will ensure that all waste packages are sub-critical, and the containment provided by the engineered barrier system of a geological disposal facility (GDF) will ensure that criticality is prevented for some time after disposal. However, as waste packages begin to degrade, fissile and other materials may be mobilised, which could affect the potential for criticality. Therefore, the possibility for the evolution of conditions in a GDF to result in criticality requires assessment. RWM established the 'Likelihood of Criticality' research project to develop qualitative and quantitative arguments to demonstrate that criticality after closure of a GDF is unlikely. The project included consideration of the criticality safety of disposal concepts for intermediate and high level waste, spent fuel, separated plutonium and highly-enriched uranium in different generic geological settings. The results are underpinned by the construction of post-closure criticality scenarios that have been evaluated using high-level judgments about scenario credibility as well as modelling of evolving conditions in a GDF. In the latter case, a probabilistic modelling approach has been undertaken to address uncertainty in the calculation of concentrations and masses of fissile material that could occur at potential accumulation sites. The potential for criticality has been judged based on comparisons of the calculated fissile masses with minimum values required for criticality at different locations in the barrier system. This paper explains the novel approach taken to evaluating the likelihood of criticality in a GDF based on the probabilistic modelling of waste package evolution and plutonium and uranium migration. The role of this research in RWM's work to demonstrate the criticality safety of a GDF in the UK is presented in a companion paper, which also draws on the results of a parallel research project that demonstrates that, even if a criticality were to occur, then its consequences would be insignificant to GDF performance. (authors)

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DOE OSTI2018

Pathways towards Valid Data: Assessment of Optimized Packaging Strategies for Core Components - 18158

Every company that runs nuclear power plants needs a strategy for disposal of the radioactive waste and for decommissioning and dismantling of its nuclear reactors. The dismantling progress of several German NPPs is already in an advanced stage. Hence, particularly highly activated core components have to be disposed of as radioactive waste. Nevertheless, this is also an issue during the operating phase due to necessary replacements of highly activated core components such as PWR control elements or RPV heads. The conditioning of waste packages for final disposal in the German repository Konrad requires the core components to be dismantled, packaged and dried. Therefore, a qualified packaging strategy is needed to ensure a reliable compliance with the protection objectives of the repository and an optimized utilization of prescribed contingents. Exemplified by PWR control elements a pathway is drawn how to perform an optimized packaging strategy in compliance with the protection objectives of the repository based on only few relevant radionuclides derived from activation calculations. For these calculations, particularly the influence of trace impurities in the materials has to be considered and their amounts have to be verified or analyzed by sampling. Furthermore, low nuclide activities from contamination cannot simply be neglected because some of them are obliged to be declared or their radiological relevance has not been sufficiently determined. Due to various influence factors, qualified sampling strategies need carefully considered concepts. Therefore, detailed planning facilitates the qualification and approval process and reduces the costs for analyses and the number of containers. (authors)

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DOE OSTI2016

Design and Construction of a Loess-Cement Cushion as an Integral Component of an SL-LILW Repository - 16077

The Bulgarian State Enterprise for Radioactive Waste Management (SERAW) is planning the construction of a near-surface repository for short-lived low and intermediate level radioactive waste, the National Disposal Facility (NDF), adjacent to the Kozloduy Nuclear Power Plant (KNPP). The NDF will be constructed as a reinforced concrete vault-type repository system, where prefabricated cubic waste packages (2 m on a side) will be emplaced in large reinforced concrete disposal cells. Waste emplacement operations will be conducted in three phases over a total of 60 years of operations. After the last waste packages have been disposed, an engineered multilayer cover will be installed over the site to further isolate the waste from the accessible environment. The Radiana Site where the NDF will be constructed is a sloping area located between the second and sixth loess terraces above the River Danube with elevations ranging from approximately +40 m to elevation +95 m. The unsaturated zone at the location selected for waste disposal ranges from 13 m to 16 m in thickness. After construction, the unsaturated zone will consist of a 5 m thick, engineered loess-cement cushion installed over naturally occurring Pliocene clays and sandy clayey sediments of the Brusarci Formation with a thickness above the top of groundwater ranging from 8 m to 11 m. The loess-cement cushion is a key component of the 'engineered fill' barrier as identified in the Intermediate Safety Assessment Report (ISAR) for the NDF. This barrier consists of engineered fill both below the disposal cells, i.e., the loess-cement cushion, and above the disposal cells, i.e. the final multilayer cover to be installed at closure of the facility. The loess-cement cushion component of this barrier will act to provide a geo-technically stable foundation for the disposal system. In addition, the barrier will raise the base of the disposal cells and increases the unsaturated zone thickness above the top of the groundwater table, as well as to serve as a chemical barrier that will act to sorb radionuclides thereby retarding their migration to the underlying groundwater. In addition to its assigned barrier functions the loess cement cushion also houses key embedded components required as part of the overall NDF safety concept, including systems necessary for operational and post-closure performance monitoring as well as the support foundation structures required for the mobile roofs, which are designed to handle and protect the waste during emplacement operations. The post-closure monitoring period, i.e., the period of institutional control, is given by as 300 years. While similar although less massive cushions have been previously constructed as foundations for various structures at the KNPP, the loess-cement cushion planned for construction at the NDF including the embedded structures will represent a first of its kind construction in terms of both its physical dimensioning as well as its requirements for long-term performance. This paper describes the design of the cushion, its functions and its construction as planned for the NDF as an integral component of the site safety concept. (authors)

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DOE OSTI2017

Web-Based Learning as a Tool for Public Education and Outreach - 17503

As internet access grows worldwide, so do online learning opportunities. Academic institutions have provided Web-based courses for many years, more are joining this set each year, and online training is common in the workplace. Web-based learning represents an opportunity for sharing information with members of the public interested in learning more about nuclear energy and approaches for managing associated spent fuel and radioactive wastes. A literature review was conducted to assess current practices and opportunities for Web-based learning relevant to nuclear energy and radioactive waste management, and to identify insights toward enhancing public education and outreach. Results indicate that substantial learning is being achieved online. A review of several online training materials related to nuclear waste management found that while informative, some are text-heavy with few supporting graphics. Many educational materials designed for the workforce could also serve as a foundation for informational materials for the broader public. To address this expanded objective, basic concepts could be illustrated with graphics, the use of video clips could be expanded, a tiered approach could be applied to present increasing levels of detail, and additional practical exercises could be included to reflect field situations. Such enhancements could further strengthen the effectiveness of Web-based learning as a tool for education and outreach, both for the nuclear workforce and for citizens interested in these topics. (authors)

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DOE OSTI2018

Optimising National WM Strategies through early consideration of Nuclear Safeguards - 18607

International Atomic Energy Agency (IAEA) safeguards stand alongside safety and security as one of the three pillars for prudent management of all nuclear material, including nuclear waste. Although safety considerations, including public and environmental radiation protection, are the traditional drivers of waste management strategies, there is a requirement for nuclear safeguards in shaping and quantifying long-term solutions for special classes of radioactive waste. This is especially important when considering the disposal of nuclear material in a repository that could be released from regulatory oversight for radiation protection after a period of institutional control, as some safeguards oversight would still be required. Failure to consider nuclear safeguards at the early stages of planning may alternatively lead to excessive volumes for disposal or the need to recondition already treated wastes. At the very least, safeguards measures should be considered before countries seek to dispose of legacy nuclear materials, decommissioning wastes and other materials containing uranium, thorium or plutonium in an operating repository. Nuclear safeguards are the set of tools and techniques used by the IAEA to verify that countries comply with their international obligations not to divert nuclear materials to nuclear explosive devices. Under the Nuclear Non-Proliferation Treaty (NPT), nuclear safeguards apply to all uranium, thorium and plutonium held in non-weapons NPT states, with limited exceptions. In nuclear-weapon states, much of their material is likewise subject to IAEA safeguards through conditions in agreements with respective suppliers of the uranium. Under IAEA safeguards, such nuclear material must remain accounted for and accessible for periodic verification-analogous to the safety concepts of reversibility and retrievability. Nevertheless, safeguards requirements can be terminated - provided that the State and the [IAEA] agree that such nuclear material is practicably irrecoverable - through dilution or immobilisation in solid form (such as cement, glass or ceramic). This means candidate material can be disposed of as non- nuclear material, with no further safeguards requirements, provided this criterion is met. Unlike radiological wastes where repository design strongly influences pathways for radionuclide migration and overall safety performance, repository design is not a factor in considering the termination of safeguards on nuclear material. This paper looks at similarities and differences in the safety, security and safeguards regulatory requirements for the management of radioactive waste and nuclear material in Australia. In particular, it outlines recent experience gained from participation in Australia's waste acceptance criteria working group for a national low-level waste repository and intermediate-level waste store, as well as from Australia's overall waste management strategy. These items of work have coincided with the IAEA update of its internal policy on termination of safeguards on nuclear material. Australia's experience shows how inclusion of safeguards criteria in the planning process is leading to optimisations in choice of facility concepts, project sequencing and equitable distribution of economic risk. By considering the three pillars-safety, security and safeguards-early on in the planning process, Australia's waste management strategy will provide a benchmark for other states in establishing and updating their own waste acceptance criteria for the disposal of nuclear material. (authors)

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DOE OSTI2025

RANGERS: Methodology for Design and Performance Assessment of Engineered Barrier Systems in a Salt Repository for HLW/SNF: Synthesis Report

Salt formations have long been recognized as a highly favorable host rock for the final disposal of high-level radioactive waste (HLW) in deep geological repositories. Their unique properties, including exceptional impermeability, self-healing capabilities, and thermal conductivity, make them a reliable natural barrier for the deep disposal of radioactive waste. This report focuses on the development and application of a methodology for assessing the integrity and per formance of the Engineered Barrier System (EBS) within salt-based repositories, a critical component of the multi-barrier system ensuring safe radioactive waste disposal.

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DOE OSTI2016

Thermo-Hydraulic-Mechanical Modelling of the Full-Scale Emplacement (FE) Experiment at the Mont Terri Rock Laboratory - 16490

The Full-Scale Emplacement Experiment at the Mont Terri underground research laboratory is a heater test carried out in a gallery constructed in Opalinus Clay. The aim of the experiment is both to develop emplacement procedures for radioactive waste canisters and their surrounding backfill as well as understanding the thermo-hydro-mechanical effects of disposing of radioactive waste on the engineering materials and the host rock. Coupled thermo-hydro-mechanical models have been used to aid planning of the locations of sensors, to inform decisions on the schedule for turning on the heaters as well as to make predictions of the temperatures and water pressures in the bentonite and the Opalinus Clay. The FE Experiment was preceded by a number of other experiments at the Mont Terri laboratory at smaller scales, but with similar materials, which provided crucial input to the models and gave confidence in the general process understanding. Blind predictions were made prior to turning on the heaters, of the temperature and relative humidity in the bentonite surrounding the heaters. The comparison between the blind predictions and early data from the experiment show that knowledge gained from smaller scale experiments on similar materials has enabled good blind predictions to be made and gives confidence that temperature in the experiment is unlikely to rise to a level that could damage the sensors. It also demonstrates that the thermal properties of the disposal system are understood well enough to produce reliable estimates of temperature in the backfill. (authors)

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DOE OSTI2016

A New Consolidated Storage Facility for Used Nuclear Fuel in Andrews County, Texas - 16418

On February 6, 2015, Waste Control Specialists LLC (WCS) announced its plan to file a license application with the U.S. Nuclear Regulatory Commission (NRC) to construct and operate a Consolidated Interim Storage Facility (CISF). Upon issuance of the license, WCS would be authorized to receive Used Nuclear Fuel (UNF) and reactor related Greater Than Class C Low Level Radioactive Waste (referred hence forth as UNF) at the CISF in late 2020. WCS based its decision to license a CISF based on strong support from the State of Texas and the local community's willingness to host such a facility in Andrews County, Texas. WCS' decision to license a CISF in a supportive community is consistent with the consent-based licensing approach recommended by the Blue Ribbon Commission on America's Nuclear Future. WCS intends to use existing technology developed by AREVA and NAC International that have already been approved by the NRC to transport UNF from existing commercial nuclear reactors across the nation for interim storage at its 14,000 acre site in Andrew County, Texas. In the pre-application licensing meetings held with the NRC, WCS underscored that UNF would be removed first from the shutdown decommissioned reactor sites and transported to the CISF. Approximately eighty percent of the UNF currently residing at the shutdown decommissioned reactor sites can be safely placed into the storage systems developed by AREVA and NAC International that are currently licensed by the NRC. WCS plans to store up to 40,000 Metric Tons of Heavy Metal (MTHM) at the CISF that would be constructed in eight phases over the next 20 years. WCS intends to submit its license application for the CISF to the NRC by April 2016. The license application will request authorization to possess up to 5,000 MTHM of UNF for Phase One which can accommodate storing the fuel located at over ten shutdown decommissioned reactor sites. As discussed with the NRC, the license application will be developed following NRC regulatory guidance to comply with the technical requirements specified in Title 10 of the Code of Federal Regulations (CFR) Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste. In addition, to filing the license application, WCS is also seeking legislative changes to the Nuclear Waste Policy Act of 1982 requiring the U.S. Department of Energy (DOE) to take title of the UNF at the commercial nuclear reactor sites. Under such legislative changes, the DOE would also be responsible for the transport of UNF to the CISF. In addition to preparing the CISF License Application, WCS and its partners AREVA and NAC International are prepared to work with DOE to facilitate the development of the UNF transportation infrastructure in the timeframe needed to support CISF operations. The WCS site and technical approach give it several advantages which uniquely position it to become the first successful commercial CISF in the U.S. These advantages include: broad consent-based support at the local, state, and Congressional level; an operating waste disposal site with an existing transportation infrastructure; and the use of already licensed systems to address 80% of the UNF currently in dry storage at shutdown sites in the U.S. WCS firmly believes this initiative will provide an economically viable option for consolidated interim storage of UNF until a permanent geologic repository becomes available. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗