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DOE OSTI2019

Implementation of Increased Salt and Decreased Sludge Volumes in the Defense Waste Processing Facility - 19597

The Savannah River Site (SRS) Liquid Waste System (LWS) safely stores and treats high-level radioactive waste. The LWS consists of 51 waste storage tanks (eight of which are operationally closed and filled with grout), waste evaporators, treatment facilities, and solidification facilities such as the Defense Waste Processing Facility (DWPF) and Saltstone Production Facility (SPF). The disposition of high-level waste (HLW) into a glass waste form poured into canisters at the DWPF began in 1996. DWPF processed a sludge only flowsheet until 2008 when the Actinide Removal Process (ARP) and Modular Caustic Side Solvent Extraction Unit (MCU) began treating the salt waste from the underground liquid storage tanks by striking with monosodium titanate (MST) to remove actinides, when necessary to meet downstream acceptance criteria, and removing the MST and entrained sludge solids (SS) by filtration before using an organic solvent to strip the cesium from the salt, creating two new concentrated HLW streams. This change caused the DWPF to begin processing under both a sludge-only and a coupled operations flowsheet where salt streams would only be added when available. This strategy allows for continuous glass canister production at DWPF. Recently, DWPF completed a 10-month melter replacement in December 2017 and the Concentrate, Storage, and Transfer Facility (CSTF) has completed repairs to the 3H Evaporator system. These facility outages have caused delays in preparing the next batch of sludge feed for DWPF. Concurrent with these facility outages, the Salt Waste Processing Facility (SWPF) has been undergoing system testing and commissioning, working toward startup. SWPF is designed to process salt waste in the same manner as ARP/MCU, but with much higher throughput volumes produced per unit time. In order to process the SWPF salt streams efficiently, DWPF will implement a coupled operation only flowsheet. Based on the facility outages mentioned above, efforts have been made to identify strategies to minimize sludge volumes and maximize salt volumes in the DWPF until the next sludge batch is ready for processing in DWPF. Savannah River Remediation (SRR) recently performed an evaluation that examines the effects of reducing the volume of sludge and maximizing the addition of the cesium-rich concentrated salt solution in each Sludge Receipt and Adjustment Tank (SRAT) batch in order to extend the existing sludge batch as long as possible until the next sludge batch is ready. These effects include a reduction of weight percent solids in the DWPF processing vessels, an increase in cycle time, a new canister heat generation value that remains within regulatory limits, and higher gamma source that remains within bulk facility gamma shielding. Balancing the above parameters resulted in a small reduction of sludge volume per SRAT batch under SWPF operation and maximizing the salt volumes for both ARP/MCU and SWPF operation. This slight reduction in sludge volume and increase in salt volumes allows SRR to mitigate some of the schedule impact incurred by the facility outages. This strategy was evaluated for both ARP/MCU and SWPF operation. This strategy has been successfully implemented under ARP/MCU operation using small step changes in the current process. This paper provides an overview of the evaluation performed for DWPF with ARP/MCU and DWPF with SWPF operations. (authors)

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

Radiation Leakage and the Precision of Characterizing Nuclear Material Using Calorimetric Assay - 19233

For measuring a waste drum containing multi-nuclide radioactive material calorimetric assay is often combined with gamma ray and neutron measurements. To reduce the methodological uncertainties of the inventory determination in a multi-nuclide matrix, the escaping radiation must be quantized as much as the encapsulated gamma-rays and α-, β-, n- particles depositing radiological heat inside the drum. To differentiate the possible variety of heat sources in a radioactive waste drum, Monte Carlo simulations are utilized resembling the waste drum and the calorimeter in both, their geometries as well as the material compositions. MCNP6-based numerical imaging of the Large Volume Calorimeter (LVC) designed by KEP Nuclear(France) allows simulating various scenarios for different particles types, energies and source locations to accompany and support the interpretation of calorimetric KEP-LVC measurements anticipated within the CHANCE project. For the analyses, the particle flux and energy deposition in each layer of the calorimeter are displayed. The results yield that a significant part of the emitted radiation leaves the system and does not contribute to the heat deposited inside the drum. As an example, the particle flux study of a 200 L mock-up drum placed inside the KEP-LVC calorimeter is discussed. A tally 4 with mesh option for both gamma and neutrons up to 5 MeV simulates the mean flux in each cell and for various source location and source distribution patterns. The expected energy deposition is cumulated for each layer, and over the whole energy range it reveals that actually a small but well known fraction of particles escapes the LVC calorimeter. This makes it a very suitable apparatus for the anticipated experiments on large and heterogeneous waste drums. The results suggest that the most interesting cases would be polyethylene, bitumen and concrete matrices [3, 5]. However, it must be kept in mind that the high-energy part of the gamma and neutron radiation flux can reach the reference chamber of the calorimeter and deposit some energy there, which may compromise the calibration and cause double-bias. The simulations showed that, depending on source location and power, neutron and high-energy gamma radiation leakage can influence the heat measurements differently. In general, part of the simulation results suggest that for waste drums with a high emission rate the accuracy of calorimetric assay can be decreased due to radiation escape. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Removing of Core Constructions of the MR and RFT Reactors in NRC 'Kurchatov Institute' - 16054

In frame of the FTP 'Providing nuclear and radiation safety of Russia in 2008 and for the period till 2015' works of a decommissioning of the MR and RFT research reactors are carried out in National Research Center 'Kurchatov Institute'. Within these works, the equipment of 9 loop-back facilities of the MR reactor was dismantled. In the conditions, which are most approached to natural, one researched a heat, hydrodynamic and strength characteristics of elements of cores and the capital equipment of the majority of power reactors at these facilities. In 2014-2015 dismantle of core constructions of the MR reactor was carried out and dismantle of a graphite laying of the RFT reactor is begun. The beryllium blocks, which were constructional elements of the core, and the graphite blocks playing a reactor reflector role were also taken, packed into cases or containers and removed: part on an accumulative platform, and part are placed in storage of high level radioactive waste of the Center for decay and further sending for long storage. Works were performed by remote controlled mechanisms, control of their performance was exercised by means of remote diagnostics: Gammavizor, the scanning spectrometer systems. The following methods and the organizational principles were applied to providing of radiation safety of these works: - the constant radiation control including individual control of a dose of external radiation, control of volume activity of radionuclides in air, annual control of internal radiation doses of the personnel; - dust suppression and fixing of radioactive contamination by dispersion of polymeric structures; - carrying out decontamination on a shift basis rooms with use of detergents; - use of the remote controlled mechanisms equipped with hydraulic scissors, nippers, application of the diamond are sharp that allowed to perform the majority of technological operations remotely, using 'cold' methods of fragmentation; - development on each type of works of the operation project and schedules. As a result, of works of 2011-2014 of the decommissioning of the MR and RFT research reactors more than 630 t of the equipment were dismantled. From them more than 455 t of radioactive waste was. The total activity of radwaste exceeded 300.0 GBq. From the pool of the reactor, about 160 channels of loop-back installations and near 230 beryllium and graphite are taken, fragmented and packed. The total activity on radionuclides {sup 60}Co and {sup 137}Cs of them was near 55 TBq. From the territory of the center about 2550 m{sup 3} of firm and 4600 m{sup 3} of liquid radwaste are removed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

High Performance Teams Enable Mission Success for SRNL - 16363

The accomplishment of the DOE EM mission of cleaning up the radioactive waste generated as a byproduct of the development and maintenance of nuclear materials over the last 60 years is a daunting task. SRNL specializes in working directly with the 'owners' of the cleanup challenges to identify the desired outcome and develop a solution that can be put into practice. 'We put science to work' is not just a 'tag line' rather it is a concise statement of our values and approach to applying science and engineering to the cleanup objectives. The management of radioactive waste poses multi-decade challenges to DOE EM and the Nation. The ability to harness the depth of knowledge, experience, expertise and passion for safely and effectively solving the technical challenges has reached a critical junction. Perhaps too late to be a natural and organic transition the industry finds itself critically devoid of resources to effectively and efficiently keep focused on the highest order task: cleanup. When project managers plan the resources necessary to perform the planned scope they typically are focused on financial, human, and industrial resources that would be beneficial to project execution. The critical shortage we will be evaluating is experienced and skilled leaders, mentors and coaches in the workforce. The production of educated scientists and engineers is a primary focus at all levels of educational institutions with Science Technology Engineering and Mathematics (STEM) focused curricula beginning in elementary school. The need for leaders who are skilled in inspiring organizational creativity, adept in harnessing and focusing individual energy while effectively communicating to the work force and the clientele equally compelling messages, has never been greater, or their absence in the workforce more ominous. Managers trained in the classical management theory (insert references here) are aging out of the workforce. In many cases these managers have not retained close connection to advancements in their primary field of study but rather have succumbed to the administrivia burdens that typify highly regulated work. The current technically trained workforce while interested in the 'perks' of management (i.e. visibility, pay increases, advancing status) are driven to keep close connection to their technical 'roots' as a fall-back plan should they need to change their career path either by choice or necessity. The challenge is how to achieve the productive hum of a finely tuned organizational model that blends executive vision and drive with technical leadership and excellence in execution. If any of these three tenets of a high performance team falters the ability to recover erodes the performance of the team and begins to weaken the cultural foundation essential to sustaining high performance work teams. At SRNL, recognition of the need to adapt the management dynamics and the business model and to harness the energy of the workforce to sustain - and in fact increase - the level of performance is a driving force in the transformation of the organization. Examples of changes in organizations structure, business systems and employee engagement are discussed as they relate to the impact on accomplishing the cleanup mission and fostering the next generation of great leaders and innovators. (authors)

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

Engineering Design of the Geological Disposal Facility for the UK - 18074

This paper discusses plans for implementing UK government policy for the safe geological disposal of higher activity radioactive waste. In order to progress the programme for geological disposal in the absence of a specific site, RWM has developed generic, illustrative disposal system designs for three host geological environments. These geological environments are typical of those being considered in other countries and have been chosen because they represent the range that may need to be addressed when developing the Geological Disposal Facility (GDF) for the UK. This paper provides an overview of the updated 2016 design of the UK geological disposal facility and describes the processes of construction, transport, waste package receipt, handling and emplacement and the design characteristics that the disposal facility will need to include for the waste inventory. The paper provides examples of the approach to ensure confidence in safety over the GDF life cycle and discusses the challenges in the design, construction and operation of a nationally significant infrastructure project with a scheduled waste disposal of more than 150 years. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Review of the Safety Analyses of the Konrad Repository - 20530

The Bundesgesellschaft fuer Endlagerung (BGE) is the license holder of all geological repository sites in Germany since 2017. Besides the construction of the Konrad repository for the emplacement of low- and intermediate-level waste (LILW), the operation of the Asse II mine including the retrieval of the emplaced waste, and the preparation for the closure of the Morsleben repository, BGE is responsible to carry out the operative tasks of selecting a site for a high-level waste (HLW) repository. The Konrad repository is Germany's first repository for radioactive waste with negligible heat generation, which was fully licensed under atomic law within the scope of a plan approval procedure from the original application in 1982 until the license was granted in 2002 and finally confirmed by the Federal Administrative Court in 2007. The emplacement of LILW was licensed up to a volume of 303,000 m{sup 3} with a total β/γ-activity of 5 x 10{sup 18} Bq and a total α-activity of 1.5 x 10{sup 17} Bq. Currently, the former iron ore mine Konrad is being converted to a repository and is expected to come into operation in 2027. But what if there is no safety case in place nor a regulatory framework tailored for a change of documents in order to ensure the safety according to the state-of-the-art of science and technology during the evolution of a long-term project such as a geological repository? In this presentation, a review procedure of the safety analyses, that have been the basis of the plan approval decision for the Konrad repository at that time, will be described. Furthermore, the first results of this graduated review procedure for updating the safety analyses will be presented. (authors)

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

Identification of Waste Streams and Chemicals of Concern for CNL's Near Surface Disposal Facility - 18057

Canadian Nuclear Laboratories (CNL) has a 70-year history in nuclear Research and Development (R and D) and the original support to the Canada Deuterium Uranium (Candu) reactor fleet. The present laboratory and operations include multiple facilities including reactors, laboratories, and buildings at various locations in Canada. The onsite facilities produced and still produce numerous and complex waste streams which have historically been placed in waste storage areas at their sites. CNL has completed the design and is in the process of licensing a Near Surface Disposal Facility (NSDF) for CNL low level waste (LLW). The current plan, pending regulatory approvals would complete construction and commissioning in 2021. CNL has identified the waste streams that are proposed to be disposed of to determine which may or may not require treatment prior to disposal and the Chemicals of Concern (COCs) that require identification and control to protect the facility, workers, public and the environment. CNL gathered the information into a Waste Stream spreadsheet that identifies the waste. Each waste stream was then evaluated as to whether or not it should meet the NSDF Waste Acceptance Criteria (WAC), and if not, which treatment technologies would be appropriate to make it ready for disposal. Understanding the hazards and associated COCs of the waste is required to safely dispose the waste. COCs present at Chalk River Laboratories (CRL) need to be identified and subsequently evaluated to determine if they pose any type of risk due to their hazardous properties or quantities. The COC data base was modified into a tool to assist waste generators. CNL has developed waste stream and COC data bases to support safe disposal of radioactive waste. Identification of waste streams and COCs for disposal is important for any disposal facility. The lessons learned by CNL as to what information is important and how to obtain the information would be useful to many facilities. (authors)

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

Indoor Radon Measurements from Residential Buildings with Basements in a Community with Potential LLRW Contamination - 19566

The Port Hope Area Initiative (PHAI) represents the Government of Canada's commitment to the cleanup and safe, local, long-term management of historic low-level radioactive waste (LLRW) in two Southern Ontario municipalities-Port Hope and Clarington. The PHAI (see reference for internet address) is based on community-recommended solutions for the cleanup and safe long-term management of approximately 1.7 million cubic meters of LLRW. It is currently one of Canada's largest environmental remediation projects at a cost of over one billion dollars. Canadian Nuclear Laboratories (CNL) is implementing the PHAI on behalf of Atomic Energy of Canada Limited, a Federal Crown Corporation. An initial characterization and remediation program in the urban area of Port Hope was conducted from 1976 to 1981. The waste is the result of radium and uranium processing in Port Hope by the former Federal Crown Corporation Eldorado Nuclear Limited between 1933 and 1988. One contaminant of concern in the LLRW is radium-226 (Ra-226) which may lead to elevated indoor levels of radon-222 (Rn-222) indoors. As part of CNL's implementation of the PHAI, Arcadis is conducting the Port Hope Project Property Radiological Survey of approximately 4,800 small-scale sites to identify 1) properties that have not been affected by historic operations and 2) properties that have residual historic low-level radioactive waste impacts requiring remediation either for soils or for interior contamination or both. Over the last six years, a component of the Small Scale Sites (SSS) characterization surveys have measured radon concentrations in essentially all buildings in Port Hope using alpha-track detectors. All properties had subsurface characterization including analytical characterization and lithology following the radon measurement reporting to measure if soil concentrations exceeded the PHAI Cleanup Criteria (PHAI CC) for Ra-226 of 0.29 Bq/g. Buildings with radon gas exceeding a trigger level of 200 becquerels per cubic meter (Bq/m{sup 3}) or those with ventilation systems required investigation of soil concentrations around the perimeter of the building. Ra-226 concentrations above the background level of 0.05 Bq/g due to LLRW sources were forwarded to create remedial action plans (RAPs) to remove the contamination near the building. The SSS surveys provide a substantive data set on the approximately 3,000 properties measured to date with interior Rn-222 gas measurements and descriptions of the physical characteristics of the buildings, the lithology of soils near the buildings and the measurements of Ra-226 concentration in the soils on the property. Consideration of the potential for contribution from non-LLRW Ra-226 from natural sources or coal ash was required on some properties. This paper describes the data and exploratory analysis of the relationship(s) between LLRW and non-LLRW factors that affect the variation in indoor radon concentrations. Historical information on the urban area, the measured Ra-226 concentration and the age of the building were correlated with radon concentrations. A large variety of native soils are found in the complex terrain and the type of native soil was correlated with the radon concentrations. A statistical analysis to address the impact of co-related explanatory variables to model indoor radon concentrations is planned when the SSS activities are complete. (authors)

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

Management of Metals from Decommissioning of Nuclear Facilities - 18411

The EDF group companies within the Cyclife platform have for decades performed treatment and recycling of metals. With different approaches but all considered international Good Practice, as valuable materials are recovered for safe recycling and volumes in the repositories for radioactive waste are saved. For an efficient decommissioning project, a well-defined waste management process is one of the critical paths, not least when it comes to routes for clearance and recycling of materials. It is easy to underestimate the complexity and bottlenecks caused by shortage of logistic space. In order to mitigate these risks, activities such as strengthening categorization, quantification and adopting different routes based on the material properties should be implemented. By categorization of the of waste and materials considered/classified as contaminated or potentially contaminated it is obvious that the management of the metals is a key parameter in nuclear decommissioning. Analyses performed by the EDF companies shows that a majority of the materials, classified as contaminated, from a BWR/PWR plant are metals. Experience gives that a high percentage of these metals from decommissioning can be subject to recycling either directly upon completion of the clearance process or after certain treatment. Non-technical key parameters like the regulatory requirements on applying the waste hierarchy, the impact on repository programs as well as, not at least, the regulatory and public acceptance of recycling of materials from the nuclear industry are, together with costs essential factors when forming the routing of metals and other waste materials from a decommissioning project. Today the EDF group companies are positioned to provide a wide range of services for management of the contaminated metals from decommissioning projects, spanning from industrial management of metals which directly will be subject general clearance after verification measurements up to decontamination and melting of selected reactor internals for recycling to public domain or within the nuclear industry. In this paper the typical waste material composition for BWR and PWR plants, the different ways to treat/manage the radioactive and potentially radioactive metals including a comparison of the different approach commonly used in a technical, logistical and economical perspectives. A few success stories will be resented as well as lessons learned. (authors)

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

Modular Platforms for Advanced Inspection, Locomotion, and Manipulation - 17150

Robots can provide remote access, manipulation, and inspection capabilities to augment human workers and improve safety in potentially dangerous decommissioning, radioactive waste management, and emergency response operations. However, such activities may require navigating challenging and unstructured environments, such as those with uneven terrain, obstacles, or loose debris. This remains a difficult task, even for modern robots. What's more, a robot's physical design can inhibit it from accessing hard-to-reach areas such as the tight spaces around, underneath, or inside piping and equipment. To address these challenges, we have developed a series of physically robust hardware modules that can be configured into a variety of robot morphologies, specialized to support different operational needs. Each hardware module consists of a number of onboard sensors and a high performance actuator with a series elastic element to sense and control interaction forces for improved locomotion and manipulation. Using this hardware, we have designed several robotic platforms that support the access, manipulation, and inspection requirements of decommissioning and radioactive waste management. To highlight current capabilities and potential opportunities for technological improvement, we present the outcomes from field demonstrations of our modular platforms at the Office of Environmental Management's Science of Safety Portsmouth Gaseous Diffusion Plant Robotics Challenge. Specifically, results include a serpentine robot climbing vertical structures such as pipes, posts, and supports to demonstrate advanced inspection. Additionally, we discuss field trials of a similar modular configuration that yields a highly dexterous manipulator arm and camera system capable of adding manipulation and inspection functionality to existing structures. Results are presented from inspection tasks in which the manipulator is installed on another mobile robotic platform. Finally, we show our hardware modules can be reconfigured into platforms capable of withstanding significant impacts and of alternative means of locomotion, as may be required to cross terrain too challenging for traditional wheeled robots. Results describe the performance of a hexapod robot that uses proprioceptive feedback to locomote in outdoor trials reminiscent of emergency response conditions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Bio-sorption of thorium by bone meal - 15435

In the recent years, there is an increasing interest in the application of biological materials for the removal of heavy metal ions from aqueous solutions. There are a variety of biological materials that have been proven efficient bio-sorbents such as agriculture residues, plants, algae, etc. The use of these materials is attractive and advantageous because they are inexpensive in comparison with commercial adsorbents. Bone meal is a biological material that reunites low cost and efficiency. It is a natural material that contains large amount of calcium phosphate in the form of hydroxyapatite. Studies with hydroxyapatite showed that this material is efficient in removing heavy metals from polluted soils due to their ability to adsorb molecules and can be used to remove radionuclides from liquid aqueous solutions, as liquid radioactive wastes generated in many activities. Bone meal was purchased in local market and chopped and sieved to obtain particle size between 0.297 mm and 0.125 mm. Batch bio-sorption experiments were performed to determine the metal uptake capacity and equilibrium time. Fixed concentrations of thorium solutions were prepared by dissolving thorium nitrate in distillate water. The bone meal was suspended in 5 mL of thorium solutions in a ratio of 0.2% w/v and left in contact during different times. The contact times adopted were 30 min, 1, 2 and 4 hours. After the contact time, the bone meal was removed by filtration and the supernatant, analyzed by inductively coupled plasma optical emission spectrometry (ICP-OES). The results were evaluated using isotherms and kinetics models. The maximum capacity for the bone meal was 11.5 mg/g in 2 h. These results suggest that bio-sorption with bone meal can be applied in the treatment of aqueous solutions and liquid radioactive wastes containing thorium. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Adsorption and Transport Properties of Zeolite SAPO-34 for Krypton/Xenon Separations

Nuclear power is an economical means of providing carbon-free energy, but requires effective ways to deal with the radioactive waste products. As part of the overall process for recycling used nuclear fuel (UNF), it is necessary to develop technologies to separate the off-gases formed during the fission of uranium and plutonium in nuclear power reactors. The off-gas contains {sup 129}I (radioisotope), NO, NO{sub 2}, CO{sub 2} containing {sup 14}C (radioisotope), {sup 3}H{sub 2}O (radioactive tritium oxide), {sup 85}Kr (radioisotope), and {sup 136}Xe. Separating and capturing each gas is essential to reduce the volume of radioactive waste for storage. Currently there exist economical technologies to separate {sup 129}I and {sup 14}CO{sub 2}. However, Kr and Xe separation is challenging because they are chemically inert and physically very similar. The Kr/Xe mixture is approximately 10/90 by volume and hence the separation of Kr would reduce the volume of radioactive {sup 85}Kr waste by a factor of 10. One way of separating them is by cryogenic distillation wherein they are condensed out as pure streams at very low temperature (Kr boiling point: -153 deg. C, Xe: -108 deg. C). Cryogenic distillation has intensive power requirements and a large system volume, resulting in high capital, operating, and disposal costs. Membrane-based separation using radiation-stable materials such as nanoporous zeolites is an attractive alternative, if membranes with sufficient Kr flux and selectivity can be fabricated. An ideal membrane process would permeate a stream of pure Kr for storage and also produce a retentate of pure Xe which has commercial value (> $8/g at 99+% purity from Airgas{sup R}). Zeolites are nanoporous crystalline aluminosilicate materials with pores that are capable of separating molecules by shape and size. SAPO-34 is a silicoaluminophosphate zeolite with pore size of 0.38 nm. The SAPO-34 is a molecular sieve composed of 3D networks with CHA frame structure containing uniform micropores. Since this pore size is intermediate between the kinetic diameters of krypton (0.36 nm) and xenon (0.396 nm), it was hypothesized that SAPO-34 membranes can separate krypton and xenon by diffusivity differences. Previously we fabricated SAPO-34 membranes and showed that they are capable of separating Kr from Xe. Here we describe a more detailed study of the fundamental adsorption and transport properties of SAPO34. In particular, the permeabilities of Kr and Xe in SAPO-34 are dependent on both their adsorption and diffusion characteristics in the nanoporous zeolite. Typically, nanoporous materials show stronger adsorption of Xe than Kr (due to stronger van der Waals forces) but faster diffusion of Kr (due to smaller kinetic diameter). Therefore the overall membrane selectivity towards Kr will be determined by competition between these two effects. A main objective of this study is to quantify and understand adsorption, diffusion, and permeation in detail and thereby obtain: (1) fundamental parameters useful for design calculations of a Kr/Xe separation membrane process, and (2) an experimental and modeling framework that can provide guidance on the expected limits of performance of SAPO-34 and other zeolite materials. (authors)

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

IAEA Initiatives that Integrate Stakeholders, Models and Decision Making - 18646

The International Atomic Energy Agency (IAEA) has initiated an effort to provide stakeholder support services as part of Phase II of the Constraints to Implementing Decommissioning and Environmental Remediation (CIDER II) programs. The intent of this sub-group of CIDER II is to provide members states support in areas of stakeholder communication, engagement and participation, with the goal of building trust in decommissioning and environmental remediation (D and ER) decisions for radioactive contamination and facilities. IAEA has also initiated a follow-on effort to address Modeling and Data for Radiological Impact Assessments (MODARIA II), which is aimed at enhancing the capabilities of Member States to simulate radionuclide transfer in the environment, and to ensure appropriate levels of protection from ionizing radiation. Working Group 1 of MODARIA II is focused on the integration of modeling and decision analysis. These two groups have complementary and supporting objectives, which provides an opportunity for collaboration and mutual benefit. The technical integrator is the approach of structured decision making, which follows a five-step process that combines stakeholder values with models through a technically defensible decision analysis. Briefly, the steps are: 1. Provide the context and decision context for the problem; 2. Obtain costs, values and concerns from stakeholders, to develop objectives that need to be achieved (e.g., improve human health, minimize costs, increase jobs); 3. Identify options that might achieve the objectives (e.g., cleanup, institutional controls, education); 4. Develop probabilistic models that connect options to objectives so that the objectives can be evaluated collectively (e.g., a fate and transport and risk model to address the objective of reducing human health risk, a cost model to address costs of taking action); and, 5. Model evaluation, including sensitivity analysis, to determine if an action can be taken with sufficient certainty, or if collection of more data is needed to decrease uncertainty before a decision is made. This approach also supports adaptive management as new data are collected. This approach clarifies the roles of stakeholders and subject matter experts in the decision-making process, where stakeholders specify values, and subject matter experts focus on the models that are needed to evaluate the options. It also provides a technically defensible, reproducible, transparent and traceable approach to all of the components necessary to properly solve the types of complex decision problems that plague radioactive waste management, decommissioning and environmental remediation. This paper introduces the major challenges to implementing such a broad scale of decision analysis, and approaches to overcoming those challenges, which include not only the technical challenges, but also the need to develop capacity in a range of skills not usually accommodated in environmental modeling and includes the need for the IAEA to address decision making in different countries, and with different cultures, religions, political affiliations, and languages. This paper explores the potential for IAEA to benefit from both the CIDER II and MODARIA II programs, and hence to develop a fully integrated approach to effective decision making for radioactive waste management and environmental remediation. Note that there are many people involved in these efforts at the IAEA, and credit belongs to those people for the progress that has been made to date. Our intent here is simply to introduce these IAEA programs at the WM2018 Symposium to provide our perspective and some awareness of some of the directions in which IAEA appears to be headed regarding stakeholder engagement, environmental and risk modeling, data science, and decision analysis. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Disposing of waste package with respect to physico-chemical characteristics - 15041

The waste inventory to be taken into account for the design of the Cigeo project is defined in the industrial program for waste management established in cooperation between nuclear operators (EDF, AREVA, CEA) and Andra. This waste inventory has been constituted with a list of different families of radioactive waste packages from High Level Waste (HLW) to Intermediate Level Long-lived Waste (IL-LLW). IL-LLW include a large variety of items such as structural elements from fuel assemblies (cladding from the fuel rods called 'hulls', end pieces called 'end caps' and assembly spacer grids, etc.), sludge from effluent treatment or miscellaneous equipment (filters, pumps, etc.). Most ILLW (except solidified sludges) are basically metallic but organic and inorganic components such as plastics (cellulose, PVC...) and salts may also be included. The rules used for designing Cigeo are based on the separation of waste having different physico-chemical characteristics in different cells. However it could be possible to dispose of waste packages with different physico-chemical characteristics in the same cell if we can demonstrate that a type of waste does not generate significant physico-chemical perturbations on other co-stored families. The first step of our analysis has been the identification of categories of materials present in waste likely to interact with each other. The second step consisted in analysing qualitatively the physico-chemical interactions between these materials. In the third step, we defined criteria allowing evaluation of the consequences of these interactions on (i) gas production (mainly H{sub 2}), (ii) radionuclide release and transfer and (iii) degradation of concrete components of the disposal cell. We built then a table allowing qualitative comparison of the consequences of the interactions between the materials types defined in the first step. The conclusion of this study shows that it is necessary to pay a very careful attention before disposing of waste packages containing organic components nearby waste packages containing actinides radionuclides. The geochemical phenomena coupling must also be quantitatively evaluated so as to analyse possible interaction processes between the released organic compounds and the cementitious components of the cells, the near-field clay-stone and the radionuclides migration. (authors)

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

Performance Assessment of a New Gamma Spectroscopy Toolkit: Peak Doctor and the Spectral Nondestructive Assay Platform (SNAP) - 17397

High resolution gamma spectroscopy (HRGS) is a widely used technique for non-destructive characterization of nuclear materials, and assay of radioactive waste. The ability to accurately evaluate raw spectral data is an important step in producing high-quality results. A spectral analysis application known as Peak Doctor has been developed under a Cooperative Research and Development Agreement (CRADA) between LANL and Pajarito Scientific Corporation (PSC). The software has undergone several years of testing and validation by the Radioactive Waste Characterization Team of the LANL Waste Management Services Group (WM-SVS). PeakDoctor is designed to work with the Spectral Nondestructive Assay Platform (SNAP{sup TM}) to provide a complete HRGS analysis engine. SNAP{sup TM} performs model specific corrections for attenuation and geometry. These applications are the major components of a new toolkit designed primarily for portable far-field non-destructive assay. Peak Doctor performs a very accurate peak fit analysis with the goal of improving upon the capabilities of existing commercially available routines. It is designed to be user-configurable and can be customized to any given type of detector and any given type of spectrum. Peak Doctor was developed to resolve many of the inherent complexities in gamma ray spectra including overlapping peaks, complex multiplet regions, large step peaks, backscatter regions and Compton edges. Unlike most spectral analysis platforms, Peak Doctor does not rely on the energy calibration and detector resolution parameters that are saved by the operator in the spectrum file. Instead, key parameters are pre-defined for individual detectors. The software fits a continuum using a series of splines between a series of junctions or 'knots'. Fitting techniques are used to join the splines together to create a near-continuous multi-element curve. Optimizing the location of the knots is one of Peak Doctor's most powerful features. The knots are initially configured using a pre-defined list and are removed or added automatically during the fitting process. Knots are also created when new features are identified such as Step Peaks and Compton Edges. The user can also manually specify knots in order to improve the continuum fit. The initial fit is improved on by removing any outlier regions (which are converted to peaks) and the continuum fitting process is repeated. At this stage, an experienced user can work interactively with the fit by manually converting peak regions to continuum or vice versa. The final step is to fit a series of peaks above the continuum based on detector parameters. Single and multiplet peaks are then processed by either the tailing (piecewise) or non-tailing (Gaussian) Levenberg Marquardt non-linear peak fitting routine depending on the user defined selection. Only a rough knowledge of the detector characteristics (energy calibration and resolution) is required for the fitting process to proceed with good accuracy. Any peaks that are missed can be manually added by the user and will also be fitted. Discontinuities in the peak / continuum fit can be rapidly identified by reviewing the 'residuals scale' plot. For waste streams that are well characterized, it is often the case that routinely encountered spectra will have similar peaks and continuum features. For these cases a handy 'One Step' method is available that proceeds in an automated fashion without user interaction. Peak Doctor has been configured with a set of default spectrum types including plutonium / transuranic waste (including Weapons Grade Pu, Heat Source Pu, Am-241, Am-243 and Np-237), depleted or enriched uranium, nuclear power plant / intermediate level waste (with fission and activation products such as Cs-137 and Co-60) as well as Naturally Occurring Radioactive Materials (NORM). The user can easily add new spectrum types for specific waste streams or other types of radiological / special nuclear materials. The output from PeakDoctor (including net peak area, centroid energy and a list of minimum detectable activity levels at pre-defined regions) is used by SNAP{sup TM} to complete the data analysis process. Isotope activities are calculated using configurable template-driven geometry and attenuation correction methods. The software accounts for matrix and container wall effects, collimator / angular corrections, detector efficiency (based on calibration curves) and stand-off distance. SNAP{sup TM} uses published algorithms incorporating a series of decision points during its operation to allow the analyst to provide logical input to guide the software during the analysis of each unique spectrum. The Peak Doctor / SNAP{sup TM} toolkit has been extensively tested and validated under a variety of conditions using 'benchmark' reference spectra. Test scenarios reported include HPGe-based measurements of well characterized uranium and plutonium standards and radioactive sources with a variety of different geometries and matrix materials including surrogate drums and boxes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Reconsolidated Salt as a Geotechnical Barrier

Salt as a geologic medium has several attributes favorable to long-term isolation of waste placed in mined openings. Salt formations are largely impermeable and induced fractures heal as stress returns to equilibrium. Permanent isolation also depends upon the ability to construct geotechnical barriers that achieve nearly the same high-performance characteristics attributed to the native salt formation. Salt repository seal concepts often include elements of reconstituted granular salt. As a specific case in point, the Waste Isolation Pilot Plant recently received regulatory approval to change the disposal panel closure design from an engineered barrier constructed of a salt-based concrete to one that employs simple run-of-mine salt and temporary bulkheads for isolation from ventilation. The Waste Isolation Pilot Plant is a radioactive waste disposal repository for defense-related transuranic elements mined from the Permian evaporite salt beds in southeast New Mexico. Its approved shaft seal design incorporates barrier components comprising salt-based concrete, bentonite, and substantial depths of crushed salt compacted to enhance reconsolidation. This paper will focus on crushed salt behavior when applied as drift closures to isolate disposal rooms during operations. Scientific aspects of salt reconsolidation have been studied extensively. The technical basis for geotechnical barrier performance has been strengthened by recent experimental findings and analogue comparisons. The panel closure change was accompanied by recognition that granular salt will return to a physical state similar to the halite surrounding it. Use of run-of-mine salt ensures physical and chemical compatibility with the repository environment and simplifies ongoing disposal operations. Our current knowledge and expected outcome of research can be assimilated with lessons learned to put forward designs and operational concepts for the next generation of salt repositories. Mined salt repositories have the potential to isolate permanently vast inventories of radioactive and hazardous wastes.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Advances in the glass formulations for the Hanford Tank Waste Treatment and Immobilization Plant - 15053

The Department of Energy-Office of River Protection (DOE-ORP) is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to treat radioactive waste currently stored in underground tanks at the Hanford site in Washington. The WTP that is being designed and constructed by a team led by Bechtel National, Inc. (BNI) will separate the tank waste into High Level Waste (HLW) and Low Activity Waste (LAW) fractions with the majority of the mass (∼90%) directed to LAW and most of the activity (>95%) directed to HLW. The pretreatment process, envisioned in the baseline, involves the dissolution of aluminum-bearing solids so as to allow the aluminum salts to be processed through the cesium ion exchange and report to the LAW Facility. There is an oxidative leaching process to affect a similar outcome for chromium-bearing wastes. Both of these unit operations were advanced to accommodate shortcomings in glass formulation for HLW inventories. A by-product of this are a series of technical challenges placed upon materials selected for the processing vessels. The advances in glass formulation play a role in revisiting the flow sheet for the WTP and hence, the unit operations that were being imposed by minimal waste loading requirements set forth in the contract for the design and construction of the plant. Another significant consideration to the most recent revision of the glass models are the impacts on resolution of technical questions associated with current efforts for design completion. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Sulfidation of magnetite with incorporated uranium

Uranium (U) is a radionuclide of key environmental interest due its abundance by mass within radioactive waste and presence in contaminated land scenarios. Ubiquitously present iron (oxyhydr)oxide mineral phases, such as (nano)magnetite, have been identified as candidates for immobilisation of U via incorporation into the mineral structure. Studies of how biogeochemicalprocesses, such as sulfidation from the presence of sulfate-reducing bacteria, may affect iron (oxyhydr)oxides and impact radionuclide mobility are important in order to underpin geological disposal of radioactive waste and manage radioactively contaminated land. Here, this study utilised a highly controlled abiotic method for sulfidation of U(V) incorporated into nanomagnetite to determine the fate and speciation of U. Upon sulfidation, transient release of U into solution occurred (~8.6 % total U) for up to 3 days, despite the highly reducing conditions. As the system evolved, lepidocrocite was observed to form over a period of days to weeks. After 10 months, XAS and geochemical data showed all U was partitioned to the solid phase, as both nanoparticulate uraninite (U(IV)O2) and a percentage of retained U(V). Further EXAFS analysis showed incorporation of the residual U(V) fraction into an iron (oxyhydr)oxide mineral phase, likely nanomagnetite or lepidocrocite. Overall, these results provide new insights into the stability of U(V) incorporated iron (oxyhydr)oxides during sulfidation, confirming the longer term retention of U in the solid phase under complex, environmentally relevant conditions.

Townsend, Luke↗